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All sourcesNASA NTRSDOE OSTI
DOE OSTI2018

DOSECALCX, A Bremsstrahlung Radiation Dose Code for Electron Beam Targets

The Dual-Axis Radiographic Hydrodynamic Test (DARHT) Facility is a two-axis, linear induction electron accelerator designed to produce Bremsstrahlung radiation for imaging hydrodynamic physics within rapidly compressed targets. The radiation dose code DOSECALC, used to obtain approximate results for experiments, was modified to include new materials and renamed DOSECALCX. Calculations were compared with results from the MCNP code. Photon spectra was found to agree visibly and quantitatively while angular dose data was only comparable within an order of magnitude.

43 PARTICLE ACCELERATORS↗
DOE OSTI2018

A Transient Reactor Physics Experiment with High-Fidelity, 3-D Flux Measurements for Validation and Verification

This project led to the design and construction of a system of novel micro-pocket fission detectors (MPFDs) and their deployment as part of a series of steady-state and transient experiments at the University of Wisconsin Nuclear Reactor (UWNR). The data acquired during these experiments is intended for use in validation of reactor-physics models, including those developed as part of the NEAMS Toolkit (e.g., PROTEUS) and others. Part of the work led to a completely new computational model of UWNR based on Python-automated input generation for both MCNP and Serpent, and preliminary efforts have been made to develop a mesh for models in PROTEUS and other deterministic tools.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2019

User Guide for the New Tally ACODE Capabilities

In the newest development version of MCNP (post-6.2.1), the α-eigenvalue solver has undergone some changes. A new approach for computing was added alongside the current ACODE algorithm. This document will briefly discuss the changes made by this modification, both in the current ACODE and in the new algorithm, as well as the new capabilities this method provides.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2014

GenSpec: A Genetic Algorithm for Neutron Energy Spectrum Adjustment

Presented in this report is the description of a new method for neutron energy spectrum adjustment which uses a genetic algorithm to minimize the difference between calculated and measured reaction probabilities. The measured reaction probabilities are found using neutron activation analysis. The method adjusts a trial spectrum provided by the user which is typically calculated using a neutron transport code such as MCNP. Observed benefits of this method over currently existing methods include the reduction in unrealistic artefacts in the spectral shape as well as a reduced sensitivity to increases in the energy resolution of the derived spectrum. This report presents the adjustment results for various spectrum altering bucket environments in the central cavity of the Annular Core Research Reactor, as well as the adjustment results for the spectrum in the Sandia Pulse Reactor III. In each case, the results are compared to those generated using LSL-M2, which is a code commonly used for the purpose of spectrum adjustment. The genetic algorithm produces spectrum-averaged reaction probabilities with agreement to measured values, and comparable to those resulting from LSL-M2. The true benefit to this method, the reduction of shape artefacts in the spectrum, is difficult to quantify but can be clearly seen in the comparison of the final adjustments. Beyond these preliminary results, this report also gives a thorough description of the genetic algorithm and presents instructions for running the code using the graphical user interface. In its present state, the code does not provide uncertainties or correlations for the adjusted spectrum. This capability is currently being added, and will be presented in future work.

42 ENGINEERING↗
DOE OSTI2019

Development of Critical Experiments to Provide Validation Data for Multiphysics Coupling Methods (Final Report)

Computational simulation has become one of the most important tools in the modern toolbox for nuclear reactor design. Reactor simulations, like the reactors themselves, involve multiple physical processes including neutronics, fluid flow and heat transfer (thermal-hydraulics), thermo-mechanics, fuel behavior, chemistry, and balance of plant. Each process can affect the behavior of the others, and advanced computational tools have been developed that couple separate physics models together to simulate the feedback effects. The objective of this work is to generate experimental data to support the validation of multi-physics reactor codes. It specifically targets the SHARP toolset from the NEAMS Reactor Product Line, but it is intended to be useful in any set of coded that couples thermal-hydraulics with neutronics simulations. The project takes advantage of the inherent flexibility of the RPI Reactor Critical Facility (RCF), a low power open-pool research reactor. The standard configuration of the RCF is used to explore the effects of moderator temperature on neutronic behavior and reactivity. Because the low power of the reactor generates no appreciable temperature change in the moderator, the moderator temperature is artificially changed with two 18 kW electric immersion heaters. The temperature feedback effects can be observed by recording the change in the reactor power as the reactor goes from a supercritical state, to critical, to a subcritical state. To expand the temperature range and temperature rate of change that can be investigated, an experimental apparatus consisting of a heated water loop was designed and installed in the RCF. In this setup, water from a heated reservoir tank is pumped through a pipe placed through the center of the reactor core. The water in the reservoir can be heated to around 70 °C and pumped through the system. The pump has a variable flow rate and electronic control, so a number of scenarios can be developed. A slug of water can be pushed through the system to generate a reasonably fast change in reactivity. The pump speed can be increased or decreased over the course of a measurement, or even oscillated at different periods, to observe the response of the reactor. To complement the experimental data generated by the RCF, a set of computational models have been developed. They are intended to illustrate the usefulness of the experimental data generated in validating coupled multi-physics simulations. A model previously constructed in MCNP has been used in conjunction with safety analysis at the RCF for many years and was modified to support the design of the heated water loop experiment. This model was translated into Serpent 2, a similar Monte Carlo code, which provided the added advantage of being able to produce collapsed cross-section data files for use in the deterministic neutron transport codes that are part of the SHARP toolset. A complete reconstruction of the RCF model, both with and without the heated fluid loop experiment, were necessary for use with the SHARP codes, PROTEUS (neutronics) and Nek5000 (thermal hydraulics), as these require an unstructured mesh format to operate. The computational requirements for fully coupled three dimensional simulations are significant, and challenges to full implementation and comparison with RCF experiments remain. A wide variety of experiments have been conducted for eventual benchmarking and validation of multi-physics reactor codes. A complete database of detailed time- and temperature-dependent measurements have been generated and stored. Several examples of the experimental results are presented here, with refinement and analysis of experiments ongoing. While these experiments still only address a finite range of operating conditions, in particular being limited to water in the sub-boiling regime, they represent a unique (to date) look at the feedback between temperature and neutronic behavior suited for validation of the coupling routines in multi-physics codes. It is hoped that these will be the foundation of future benchmarks that will provide confidence in the tools that will be used to develop the next generation of nuclear reactor systems.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Integrated Approach to Fluoride High Temperature Reactor Technology and Licensing Challenges (FHR-IRP)

This document constitutes progress made on this project by the Georgia Tech led Integrated Approach to Fluoride High Temperature Reactor Technology and Licensing Challenges (FHR-IRP). This is the first draft of the project final report. Chapter 3 is focused on Tritium management. Tritium management in Fluoride-salt High-temperature Reactors (FHRs) presents a major operational concern and a significant challenge. Tritium is generated via multiple pathways including various interactions in fuel and structural materials, as well as coolant (primarily interaction with lithium and beryllium), leading to a complex evolution and spatial distribution of tritium generation rate. FHR generates orders of magnitude more tritium than in LWRs, and has potential for a significant release, via permeation through structural materials (due to high temperature), and through heat exchanger(s). Therefore, accurate estimate is necessary. For that purpose relevant tritium production pathways were identified. A methodology to obtain space-time dependent tritium generation rate was developed and implemented, employing SCALE and MCNP codes. Detailed models of the reference Advanced High Temperature Reactor (AHTR) design were developed. A multi-parametric metrics was introduced to capture relevant tritium production indicators, including the instantaneous tritium production rate, total tritium production (neglecting subsequent decay), and accumulated tritium production (accounting for the decay). The obtained results will inform tritium management, including the sizing of the components, and to gain insight on the impact of the core design parameters on tritium generation rates. Research effort on Liquid Salt Impurity Removal, Redox and Corrosion Control is discussed in Chapter 4. Section 4.1 is an introduction chapter. Based on the current available knowledge and operation experience on previous MSRs and other types of nuclear reactors, the report identifies the major impurities that are related to either material compatibility, salt properties or radioactive source terms, and the potential contamination sources of these impurities, their transport in the primary coolant loop and available cleanup methodologies for a particular group of impurity. In section 4.2, the available purification methods were discussed, including the liquid/liquid extraction using liquid Bi-Li which has been applied in MSBR and electrochemical purification which is developed based on Pyroprocessing. In this section, the new multi-stage and multi-salt-component was developed and discussed. Section 4.3 discussed the redox potential control for salt chemistry and corrosion mitigation, including the control theory and the soluble/soluble buffer control method. The section also developed method on how to analyze soluble/soluble redox reaction in a molten salt system and designed an electrochemical sensor for redox potential control. Corrosion models were also reported in this section. Section 4.4 focused on how to measure the fundamental properties in a molten salt. As an example, the section focused on La fluoride in the molten salt. The section developed methods on how analyze electrochemical measurement data to obtain the fundamental properties including the redox potential, diffusion coefficient and exchange current density. All these fundamental information is important for salt purification system development. Section 4.5 focused on measurements of corrosion products in the salts. Because of the different behaviors between Cr and La in the salt, new analysis method for Cr fluorides in the salt was developed. Some results for Fe fluoride in the salt and corrosion measurements were also reported. Section 4.6 demonstrated the separation of oxide from the salt. This section also measured the solubility of La 2 O 3 in the salt as function of temperature. Section 4.7 focused on the separation of La and Ce from the molten salt. These two elements were selected because La was found to be separated by using inert electrode, while for separation of Ce, reactive electrode must be selected. Section 4.8 demonstrated the separation iodine ions from the molten salt. Two kinds of working electrode materials (metal W and graphite) were investigated. Section 4.9 focused on the fundamental data of fission products and actinides in liquid Bi. The chapter reviewed the available data, then based a CALPHAD model, the available data were broadened to the concentration-dependent cases.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2019

Pixelated scintillator array response to relativistic electrons

In order to make time resolved measurements of the gamma spectra on DARHT Axis 2 using the Compton Spectrometer (CS), the electron detector in the focal plane must be a scintillator that is optically coupled to a gated or streak camera. There are several issues we would like to address with experiments at Idaho Accelerator Center (IAC) before committing to a full size array (7”x2”), which would cost at least $100k. Two candidates for this application are organic (Bicron 422 or equivalent) or inorganic (LYSO) scintillator. Monte Carlo calculations were performed by Rachel Morneau, P-21, using MCNP to compare the two materials each 1.5mm thick for an energy range of 0.1 to 20MeV. The energy loss per particle in LYSO is much larger (100x) than in the plastic scintillator. We confirmed, by performing experiments with the CS on the Microton, that sensitivity of the LYSO was much larger (10x) than the plastic scintillator. Therefore, we have selected LYSO as the scintillator of choice. However, interest in the response of other organic and plastic scintillators prompted an increase in scope of the project. The goals of this experiment were as follows. First, we need to understand the optical response of the scintillators to electrons as a function of thickness, energy, and deposited charge. We also need to verify level of saturation in the scintillation process. Second, compare the scintillators’ response to an image plate (the current CS detector). Third, we need to quantify the effect of signal bleed through to adjacent pixels.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2019

Models for HEU objects in the NDSE IV and III series

A Neutron-Diagnosed Subcritical Experiment (NDSE) proof-of-principle setup exists at area-11 in Nevada. An exquisite set of gamma-ray flux die-away data has been obtained at this facility for a number of highly-enriched uranium (HEU) objects, including objects referred to as IV, IV', IV'' (the object IV series), IIIa, and IIIb (the object III series). These objects are constructed via a set of hemi-spherical concentric HEU shells (known as the Rocky Flats shells), surrounded by a sphere of aluminum (Al), encased in an outer sphere of high-density polyethylene (HDP). The “as built” detailed MCNP-calculated k eff values obtained by Goorley, for the object IV and III series, are listed in Table II. These model calculations contain detailed information about the inner and outer radii and masses of each hemi-shell used to make each of the objects. Objects IV', and IV'' were obtained by replacing inner HEU shells with Al. The low k eff of object IIIa was obtained by substituting a single HEU shell, with an inner radius of ~5 cm, with an Al shell. For object IIIb the substitution was a little further out but only in one hemi-shell. Except for the inner most shell (ball), all HEU shells have a thickness close to 0.333 cm. Previously the k eff of objects IV', IV'', IIIa, and IIIb, were determined from measured gamma ray die-away data via an analysis using a simple modification of the density of the HEU in our simple three layered model (HEU, Al, and HDP) of object IV including the outer safety can. This was done to see the quality of the inference and investigate any limitations associated with not using detailed information about the known perturbations used to make these objects. In this work, we built models (priors) of objects IV', IV'', IIIa, and IIIb that make use of our knowledge about the nature of the perturbations. With these models, instead of adjusting the density of the HEU in our object IV model to match the data from objects IV', IV'', IIIa, and IIIb, as done previously, we will be able to (in future studies) adjust r AL , and r i to infer the k eff of the corresponding objects.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Enhancing TREAT’s Pulsing Capabilities Using TREAT Upgrade Fuel

The Transient Reactor Test Facility (TREAT) at Idaho National Laboratory is an air-cooled, thermal, heterogeneous facility used to test reactor materials in simulated accident conditions by inducing fission heating with intense neutron pulses. TREAT operated from 1959 until 1994 with the primary goal of testing fast reactor fuels and was brought back online in 2017 to re-establish DOE's nuclear fuels transient testing capabilities. Beginning in the late 1970s, Idaho National Laboratory worked in conjunction with other organizations to increase TREAT's capability for in-pile testing. While these upgrades had not been implemented by the time the reactor was put into standby mode, new assemblies and graphite-urania fuel blocks with increased uranium concentrations had already been designed and fabricated. This project, as part of Idaho National Laboratory's Department of Reactor Physics, modeled the TREAT Upgrade (TU) fuel assemblies in MCNP and implemented them in an existing TREAT model. Calculations were performed using INL's high performance computer to find a critical combination of TU and standard fuel and this geometry's excess reactivity. Future work will include calculating a power coupling factor in the experiment, determining power peaking factors throughout the core, and assessing how much reactivity would be added by replacing the Inconel-625 cladding with silicon carbide.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2019

Enhancing TREAT's pulsing abilities using TREAT upgrade fuel. SULI Intern Report

The Transient Reactor Test Facility (TREAT) at Idaho National Laboratory is an air-cooled, thermal, heterogeneous facility used to test reactor materials in simulated accident conditions by inducing fission heating with intense neutron pulses. TREAT operated from 1959 until 1994 with the primary goal of testing fast reactor fuels and was brought back online in 2017 to re-establish DOE’s nuclear fuels transient testing capabilities. Beginning in the late 1970s, Idaho National Laboratory worked in conjunction with other organizations to increase TREAT’s capability for in-pile testing. While these upgrades had not been implemented by the time the reactor was put into standby mode, new assemblies and graphite-urania fuel blocks with increased uranium concentrations had already been designed and fabricated. This project, as part of Idaho National Laboratory’s Department of Reactor Physics, modeled the TREAT Upgrade (TU) fuel assemblies in MCNP and implemented them in an existing TREAT model. Calculations were performed using INL’s high performance computer to find a critical combination of TU and standard fuel and this geometry’s excess reactivity. Future work will include calculating a power coupling factor in the experiment, determining power peaking factors throughout the core, and assessing how much reactivity would be added by replacing the Inconel-625 cladding with silicon carbide.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2016

Deterministic Calculation of Gamma-Ray Transport Spectra with InterScatter.

Gamma-ray transport describes the way in which high-energy photons interact with matter. There are many well established, general purpose gamma transport calculation frameworks (such as GEANT4, MCNP, PARTISN, and GADRAS) that allow for complex generalized calculations, but they are typically not designed to be integrated into other multi-process, multi-platform applications, not designed for rapid computation, or their complexity leads to user errors. In this report, we introduce InterScatter: a deterministic simulation designed to quickly and easily calculate gamma-ray transport spectra for a 1D shielding geometry.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2016

Deterministic Calculation of Gamma-Ray Transport Spectra with InterScatter

Gamma-ray transport describes the way in which high-energy photons (gamma rays and x-rays) interact with matter. There are many well established, general purpose gamma transport calculation frameworks (such as GEANT4, MCNP, PARTISN, and GADRAS [1] [2] [3] [4]) that allow for complex generalized calculations, but they are typically not designed to be integrated into other multi-process, multi-platform applications, not designed for rapid computation, or their complexity leads to user errors. In this report, we introduce InterScatter: a deterministic simulation designed to quickly and easily calculate gamma-ray transport spectra for a 1D shielding geometry.

72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS↗
DOE OSTI2019

Nuclear Data – Benchmarking 19 F(α,n) Yield Data for Nuclear Safeguards

Fluorine compounds of U and Pu are ubiquitous in the nuclear fuel cycle, so F(α,n) neutrons are an important signature and quantitative source term that needs to be understood for physics-based interpretation of nondestructive assay measurements. Historically there have been large differences in the basic nuclear data for this reaction reported by various groups. UF 6 is the most prominent material in the nuclear fuel cycle, with the potential for short-term production into weaponizable form. Verification of bulk quantities, natural feed, depleted tails, and especially low enriched product in cylinders is particularly important. The principal physical measurement is neutron counting. For enriched material, the 19 F(α,n) reaction, driven by 234U, is the dominant source of neutrons. Basic nuclear data, cross sections, needed to calculate the yield, as well as information on the source spectrum for sensitivity studies are sparse and highly discrepant. This limits defensible physics-based performance evaluations. The same applies to holdup and criticality studies in which hydrated uranyl fluoride is the material of interest which accumulates in enrichment facilities. In the future we anticipate that the physics community will have made improved accelerator-based measurements and undertaken a more thorough evaluation and adjustment of all relevant available data. But in the short term, this does not help the safeguards community face the pressing real-world nondestructive assay requirements. To address this need, we have performed quality neutron measurements on UF6 materials using well-known material and high-capability neutron counters. A robust determination of the n/s/g of 234 U in UF 6 was generated along with a scientifically defensible uncertainty analysis. The weighted average value of the neutron yield is 509 n/s/g 234 U with a random uncertainty of approximately 1%. Dominant sources of systematic uncertainty are in the efficiency determination where we incur approximately 1.1% relative standard deviation associated with the 252 Cf reference source and about 0.9% uncertainty coming from the uncertainty in the F(α,n) spectrum. The stated uncertainties are an order of magnitude better than current data based on accelerator data reported in the literature. Aside from being of immediate and direct use by the safeguards community, this result will serve as an enduring integral benchmark for subsequent data evaluations that will need to match it in order to be credible. In addition to yield normalization, safeguards users of F(α,n) data also need guidance of the neutron emission spectrum. We have generated a neutron emission using a modified version of the well-known SOURCES 4C code, updated with new alpha stopping power coefficients, and a blended microscopic cross section data set. The yield and spectra are available in a form that MCNP users can readily use.

36 MATERIALS SCIENCE↗
DOE OSTI2019

Critical Experiment Benchmark Results using MCNP6.2 Unstructured Mesh

The purpose of this report is to document the results of five critical experiment benchmarks modeled using the unstructured mesh (UM) geometry capability in MCNP®1 code, version 6.2. LANL currently has a vast critical experiment benchmark library with over 1100 cases, modeled using constructive solid geometry (CSG) and packaged with Whisper-1.1. Validation efforts for the purposes of nuclear criticality safety typically focus on benchmark libraries built using CSG models. Results of UM geometry models are compared with CSG results and the critical experiment k-effective from the ICSBEP handbook. All computed k-effective results in this study are well within 1% of the experimental result.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2019

Summary of LANL FY19 Technical Interface Coordination with SHINE

In FY19, Los Alamos National Laboratory (LANL) provided technical support for SHINE emergent needs and technical interface coordination including coordination of SHINE CFD and thermal hydraulics conference calls, Mo-99 meetings and reviews, national laboratory conference calls, and program management activities such as reporting and tracking requirements for current FY19 activities. LANL has also provided general technology consultation support for emerging issues. LANL has participated in the bi-weekly conference calls with SHINE and other national laboratories regarding LANL’s current activities. Activities discussed include FY18 carryover items including the ANSYS Fluent-Nek5000 comparison study (for assessment of Nek5000 for coupled calculations), Supo 3-D CFD model geometry validation, and releasing CFD-MCNP coupled codes for the generic solution vessel simulations. Also, additional mid-year funded activities discussed during these conference calls include the ANL bubble experiment #2 collaboration. SHINE had a general technology consultation request in FY19. SHINE inquired about LANL’s capability to measure enrichment percentage of uranium samples, however, additional information is needed to continue. The meeting minutes for these activities are shown in Appendix A. LANL has also been providing monthly updates on the laboratory SharePoint site (managed by Pacific Northwest National Laboratory) regarding current FY19 activities. Monthly updates include current activity milestone and deliverable status, completion date updates and variances if applicable, and actual cost at the task dependent activity level. LANL attended several Mo-99 meetings and reviews in FY19. LANL attended the FY19 Mo-99 Laboratory Mid-Year Program and Financial Review at Argonne National Laboratory (ANL) on March 20-21, 2019, as well as the FY19 Mo-99 Program Meeting and Mid-Year Financial Review in Chicago, IL on April 29-30, 2019. LANL also attended the FY20 Laboratory Work Package Planning Meeting at ANL on June 4-6, 2019.

43 PARTICLE ACCELERATORS↗
DOE OSTI2019

Extension and Demonstration of NEAMS Multiphysics Tools to Lead-Cooled, Sodium-Cooled, and Molten Salt Fast Reactor Applications

The SHARP toolkit is a high-fidelity reactor simulation tool developed under the U.S. Department of Energy, Office of Nuclear Energy Advanced Modeling and Simulation (NEAMS) Campaign. SHARP toolkit is comprised of the neutronics module PROTEUS thermal hydraulics module Nek5000, and structural mechanics module Diablo. During FY17 and FY18, the PROTEUS and Nek5000 components of SHARP were applied to solve challenging sodium-cooled fast reactor (SFR) problems. In particular, selected hot channel factors (HCF) for a prototype metal-fueled SFR design (the AFR-100) were analyzed in high fidelity, and the “SHARP zooming capability” for SFRs was developed and demonstrated to reduce computational expense for full core problems in cases where detailed data is needed in selected fuel assemblies. After the previous success applying SHARP to challenging SFR problems, the focus in FY19 expanded to additional fast reactor applications: lead cooled fast reactors (LFR), sodium cooled fast reactors (SFR), and fast molten salt reactors (MSR). The specific technical tasks were (1) assessment of hot channel factors for LFR, for which no data currently exists, (2) implementation of a gamma transport capability in the PROTEUS solvers to extend the zooming capability to non-fueled assemblies, (3) demonstration of zooming capability in assemblies of the Versatile Test Reactor (VTR), and (4) development of a new coupled capability for simulating fast MSRs, specifically modeling precursor transport inside and outside the core. The progress on tasks (1) and (3) are discussed primarily in this document. Tasks (2) and (4) are briefly mentioned in this document, and the interested reader is referred to separate reports for detailed information. First-of-a-kind hot channel factor (HCF) estimation for LFR with high fidelity codes (PROTEUS/Nek5000) was successfully demonstrated in this study. Selected HCF were computed and compared with SFR data (AFR-100, EBR-II). The findings confirm that different reactor types, design parameters and uncertainties lead to different HCFs. Careful estimation of HCF for a specific design is necessary to obtain appropriate HCFs. In addition to improvement in HCF accuracy, high fidelity tools generate data to help the designer better understand the mechanism of the impact from these uncertainties. For example, the impact of cladding thickness manufacturing tolerance resulted in non-intuitive effects in the corner pins of the LFR assembly. This procedure of computing HCF using high fidelity models shows promise and flexibility for being repeated for any arbitrary reactor of choice, since these tools are targeted to solve many reactor types and geometries. Progress was made towards extending the previously demonstrated SHARP zooming capability to non-fueled SFR assemblies. In particular, a gamma transport capability was implemented in both high fidelity PROTEUS solvers in order to accurately account for heat deposition caused by gamma particles, which accounts for ~10% of total core power. Neutronics verification cases were carried out for a candidate Versatile Test Reactor (VTR) design using the new gamma transport capability in PROTEUS. Comparisons were made with continuous energy MCNP calculations and shown to agree well. The models for the full core design with heterogeneous control and fuel assemblies is in progress and will be completed in FY20. In the fast molten salt reactor (MSR) domain, a new capability was developed to model the flow of neutron precursors both inside and outside of the core, which impacts the effective delayed neutron fraction of the system. The PROTEUS-SN and PROTEUS-NODAL solvers were coupled to Nek5000 in order to simulate precursor flow. The workflows were demonstrated on test problems. The fast MSR modeling capability was supported under two other work packages, and comprehensive details are included in a separate report.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

Development of a Fast Spectrum Test Loop at the Advanced Test Reactor

The United States currently has no infrastructure to provide a fast neutron flux for research purposes. With the increased interest in Sodium Fast Reactor (SFR) technology, this type of infrastructure is necessary for further development of key components such as Versatile Test Reactor. The purpose of this research is to characterize the thermal hydraulics of a test loop capable of providing SFR like operating conditions. This includes showcasing feasibility in mechanical design, reaching a high fast to thermal neutron flux ratio ideally near 15, an inlet temperature of 600 K, a temperature increase of 50 K from inlet to outlet, and test chamber radial temperature profiles similar to SFR fuel rods. Existing infrastructure must be used to provide the neutron flux for the test loop, and it was decided that the Advanced Test Reactor (ATR) was the optimal choice due to its availability and previous work conducted similar to this project. To achieve the goals of this project, several mechanical designs were put forward with the most feasible design being chosen. The neutronics and thermal hydraulics analysis were conducted in tandem, with an iterative process used to maximize efficiency of both systems. Analysis of the neutronics system was done using Attila and MCNP while the thermal hydraulic analysis was done using RELAP5. This document outlines the thermal hydraulic analysis, though it includes the results from the mechanical design and neutronics analysis. The loop design includes a primary system with molten sodium, a set of “booster fuel plates” to increase the fast flux within the test section, a water system to cool the booster fuel plates, a helium system to separate the sodium and water systems, and a thermal neutron shield to decrease the thermal flux. Each of these systems is modeled in the neutronics and thermal hydraulic analyses. The RELAP5 model showcased the feasibility of the design proposed by this report. For all designs tested, the desired 50 K temperature increase from inlet to outlet was reached. The fuel rod radial temperature profile was lower than was expected, likely due to a lower rod thermal power than expected. A parametric study was also conducted to characterize the system with three fuel pins. This study showed the effect of thermal power, fluid pressure, fluid velocity, and inlet temperature on the system. ©Copyright by Anton Higgins December 11, 2019 All Rights Reserved

99 GENERAL AND MISCELLANEOUS↗