Progress/Status of an Upgraded Reactor Safety Analysis Code for Sandia's Annular Core Research Reactor.
Abstract not provided.
Talley, Darren G.↗
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Cavity temperature and flow characteristics in a gas-core test reactor
A test reactor concept for conducting basic studies on a fissioning uranium plasma and for testing various gas-core reactor concepts is analyzed. The test reactor consists of a conventional fuel-element region surrounding a 61-cm-(2-ft-) diameter cavity region which contains the plasma experiment. The fuel elements provide the neutron flux for the cavity region. The design operating conditions include 60-MW reactor power, 2.7-MW cavity power, 200-atm cavity pressure, and an average uranium plasma temperature of 15,000 K. The analytical results are given for cavity radiant heat transfer, hydrogen transpiration cooling, and uranium wire or powder injection.
Putre, H. A.↗
Evaluation of critical mass for open-cycle gas- core rocket reactors
Critical mass calculation for open cycle gas core rocket reactors, considering cavity size, fuel radius, reflector thickness and hydrogen bypass flow
Hyland, R. E.↗
HFIR Core Modeling and Analysis for the PROSPECT Reactor Antineutrino Experiment
Nuclear reactors have been employed as a copious source of antineutrinos for over six decades. Research initially directed at the study of antineutrino physics has led to applications of these highly penetrating particles to reactor safeguards because the antineutrino flux and spectrum can be related to the state of the reactor core, including burnup evolution. One of the main challenges of antineutrino-based safeguards is operation of the detectors near surface (within the reactor containment or in the close proximity). Lack of overburden results in increased contribution to the detector noise from cosmic radiation. Similarly to safeguards, experiments to definitively confirm or exclude the existence of sterile neutrino oscillations require detector technologies capable of operating at short baselines (tens of meters) at the surface. The Precision Reactor Oscillation and Spectrum Experiment (PROSPECT) at the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL) is an experiment with a main objective to make a precision measurement of the flux and energy spectrum of antineutrinos emitted from nuclear reactors in order to search for the oscillation signature of sterile neutrinos. Fissions of highly-enriched uranium of HFIR will serve as a source of antineutrinos. (authors)
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Some Nuclear Calculations of U-235-D2O Gaseous-Core Cavity Reactors
The results of a multigroup, diffusion theory study of spherical gaseous-core cavity reactors are presented in this report. The reactor cavity of gaseous U235 is enclosed by a region of hydrogen gas and is separated from an external D2O moderator-reflector by a zirconium structural shell. Some cylindrical reactors are also investigated. A parametric study of spherical reactors indicates that, for the range of variables studied, critical mass increases as: (1) Fuel region is compressed within the reactor cavity, (2) moderator thickness is decreased, (3) structural shell thickness is increased, and (4) moderator temperature is increased. A buckling analogy is used to estimate the critical mass of fully reflected cylindrical reactors from spherical results without fuel compression. For a reactor cavity of a 120-centimeter radius uniformly filled with fuel, no structural shell, a moderator temperature of 70 F, and a moderator thickness of 100 centimeters, the critical mass of a spherical reactor is 3.1 kilograms while that of a cylinder with a length-to-diameter ratio of 1.0 (L/D = 1) is approximately 3.8 kilograms and, with L/D = 2, 5.9 kilograms. For the range of variables considered for U235-D2O gaseous-core cavity reactors, the systems are characterized by 95 to 99 percent thermal absorptions, with the flux reaching a maximum in the moderator about 10 to 15 centimeters from the reactor cavity.
Ragsdale, Robert G.↗
Technical basis in support of the conversion of the University of Missouri Research Reactor (MURR) core from highly-enriched to low-enriched uranium - core neutron physics
This report contains the results of reactor design and performance for conversion of the University of Missouri Research Reactor (MURR) from the use of highly-enriched uranium (HEU) fuel to the use of low-enriched uranium (LEU) fuel. The analyses were performed by staff members of the Global Threat Reduction Initiative (GTRI) Reactor Conversion Program at the Argonne National Laboratory (ANL) and the MURR Facility. The core conversion to LEU is being performed with financial support of the U. S. government.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
Annular Core Research Reactor (ACRR) Pulse Curve Characterization
Reactor pulse characterization at the Annular Core Research Reactor (ACRR) at Sandia Technical Area V (TA-V) is commonly done through photo conductive detection (PCD) and calorimeter detectors. Each of these offer a mode of analyzing a digital signal with different advantages/methods for determination of integrated dose or temporal based metrology. This report outlines a method and code that takes the millions of data points from such detectors and delivers a characteristic pulse trendline through two main methods: digital signal filtration and machine learning, in particular, Support Vector Machines (SVMs). Each method's endpoint is to deliver a characteristic curve for the many bucket environments of ACRR while considering other points of interest including delayed gamma removal for prompt dose metrics. This work draws and adds on previous work detailing the delayed gamma fraction contributions from CINDER simulations of the ACRR. Results from this project show a method to determine characteristic curves in a way that has previously been limited by data set size.
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Shift Validation Plan for CASL
This document outlines the benchmarks that will be used for Shift validation for CASL. This plan is divided into five main types of problems, with each building upon the previous: small criticality benchmark problems, including small reactor configurations and critical assemblies; full-core reactor benchmarks for light water reactors; fixed source coupled neutron-photon dosimetry benchmarks; depletion/burnup benchmarks ranging from small pincell problems to full-core reactor depletion; and full-core reactor performance benchmarks. Some of these benchmarks have measured data for comparison and some only have simulated Monte Carlo radiation transport code results for comparison. Upon completion of this validation plan, we will have confidence in Shift to provide reference results for CASL benchmarking.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
Annular Core Research Reactor High Bay Neutron Energy Spectrum Measurements
The Annular Core Research Reactor (ACRR) recently underwent a multi-department effort to characterize the neutron energy leakage spectra in the reactor High Bay for multiple shielding configurations. The ACRR Operations Department identified and obtained a Rotating Neutron Spectrometer (ROSPEC) from Los Alamos National Laboratories (LANL) for obtaining neutron energy spectra. Spectra were obtained in the following facility configurations; Fuel Ringed External Cavity, Version II (FREC-II) decoupled with the standard FREC shield plug installed with the following operational permutations: the 4-inch central cavity (CC) shield plug removed in both a free-field (FF) and 44(3y (Blead-boron (LB-44) bucket neutron modifier configuration, fully shielded CC plug FF condition, neutron radiography collimator aperture settings 250:1, 125:1, and 65:1, FREC-II coupled with both the lightweight (2T1) and standard shield plug configurations. The ROSPEC spectrum results were unfolded, evaluated and weighted using the International Commission on Radiation Protection (ICRP) 103 neutron radiation weighting factors to obtain a ratio amongst the different shielding configurations at the ACRR. Results of the collected spectra confirm that the predominant neutron energy range in the High Bay area is less than or equal to 250 keV.
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Evaluation of critical mass for open-cycle, gas-core rocket reactors
Evaluation of critical mass for open cycle, gas core rocket reactors
Hyland, R. E.↗
Characteristics of Axial Power Shape and Flow Field at Low Flow Condition
The representative event which experiences the low flow condition is SLB (Steam Line Break) event. During the SLB event with LOOP (Loss of Offsite Power), the reactor core is tripped and all RCPs (Reactor Coolant Pump) are stopped. The biggest worth control rod also stuck out. As a result, the flow rate and pressure in reactor core decrease and the peaking factor is relatively high compared to a nominal operating condition. In this paper, we study the tendency of the core axial power distribution and flow field with respect to the core average power at the SLB accident condition. Generally, the condition has been analyzed using the neutronics code with one dimensional flow result of the assemblies. However, if the core average power increases, the coolant density in the hot channel decreases and mass flux increases. The assumption that the mass flux is constant at the one dimensional analysis is incorrect. Hence, it is needed to couple the neutronics code and the subchannel analysis code in order to give feedback to neutronics code with the 3-D flow information. In this study, we coupled the nuclear design code and the subchannel analysis code and showed the characteristics of the axial power distribution and the flow field at the low flow condition. Since there is no experimental data at the condition, the characteristics of the power distribution and flow field were analyzed to understand the physical phenomena. (authors)
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
Transformational Challenge Reactor preconceptual core design studies
In the nuclear industry, a manufacturing-informed design approach has the potential to yield the most benefit from advanced manufacturing. By leveraging advanced materials, data science, and rapid testing and deployment, manufacturing-informed design can drive down costs and development times, ultimately improving future commercial viability. This approach is being demonstrated in the US Department of Energy Office of Nuclear Energy (DOE-NE) Transformational Challenge Reactor (TCR) program. Preconceptual design activities for TCR have been focused on analyzing and maturing four reactor core design concepts: two fast-spectrum and two thermal-spectrum systems. Furthermore, the designs were iteratively modified and analyzed, and subcomponents were manufactured in parallel over weeks instead of months or years. To meet key program initiatives (e.g., timeline and material use), several constraints—including fissile material availability, component availability, materials compatibility, and additive manufacturing capabilities—were factored into the design effort, yielding small cores less than one cubic meter in volume with near-term viability. Additionally, the TCR program has made significant progress on development of advanced moderator materials such as yttrium hydride, advancing the feasibility of gas-cooled thermal spectrum systems using less than 250 kg of high-assay low enriched uranium (HALEU) and occupying less than 1 m3. Each of the two resulting thermal designs uses a different fuel form: traditional UO2 ceramic fuel and tristructural isotropic (advanced TRISO) fuel particles embedded inside a SiC matrix. Core neutronics and thermal performance for these systems were assessed and summarized. Evaluation of the performance metrics for these two moderated designs has yielded the downselected TCR design: a TRISO-fueled and yttrium hydride moderated gas-cooled reactor.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
Fuel elements of research reactor CM
In 1961 the CM research reactor was commissioned at the Research Institute of Atomic Reactors (Dimitrovgrad, Russia), it was intended to carry on investigations and the production of transuranium nuclides. The reactor is of a tank type. Original fuel assembly contained plate fuels that were spaced with vanes and corrugated bands. Nickel was used as a cladding material, fuel meat was produced from UO{sub 2} + electrolytic nickel composition. Fuel plates have been replaced by self-spacing cross-shaped dispersion fuels clad in stainless steel. In 2005 the reactor was updated. The purpose of this updating was to increase the quantity of irradiation channels in the reactor core and to improve the neutron balance. The updating was implemented at the expense of 20 % reduction in the quantity of fuel elements in the core which released a space for extra channels and decreased the mass of structural materials in the core. The updated reactor is loaded with modified standard fuel elements with 20 % higher uranium masses. At the same time stainless steel in fuel assembly shrouds was substituted by zirconium alloy. Today in progress are investigations and work to promote the second stage of reactor updating that involve developments of cross-shaped fuel elements having low neutron absorption matrix materials. This article gives an historical account of the design and main technical changes that occurred for the CM reactor since its commissioning.
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
The SAS4A/SASSYS-1 Version 5.8 Safety Analysis Code System
SAS4A/SASSYS-1 is a software simulation tool used to perform deterministic analysis of anticipated events as well as design basis and beyond design basis accidents for advanced nuclear reactors. Detailed, mechanistic models of steady-state and transient thermal, hydraulic, kinetic, and mechanical phenomena are employed to describe the response of the reactor core, the reactor primary and secondary coolant loops, the reactor control and protection systems, and the balance-of-plant to accidents caused by changes in coolant flow, loss of heat rejection, or reactivity insertion. The consequences of single and double-fault accidents can be modeled, including fuel and coolant heating, fuel and cladding mechanical behavior, core reactivity feedbacks, coolant loop performance including natural circulation, and decay heat removal. Analyses are typically terminated upon demonstration of reactor and plant shutdown to permanently coolable conditions, or upon violation of design basis margins. The objective of the analysis is to quantify accident consequences as measured by the transient behavior of system performance parameters, such as fuel and cladding temperatures, reactivity, and cladding strain. Originally developed for analysis of sodium cooled reactors with oxide fuel clad by stainless steel, the models were subsequently extended and specialized to metallic fuel clad with advanced alloys and to several other coolant options, including lead, LBE, and water.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
Efficiency and accuracy of the perturbation response coefficient generation method for whole core comet calculations in BWR and CANDU configurations
The coarse mesh transport method (COMET) is a highly accurate and efficient computational tool which predicts whole-core neutronics behaviors for heterogeneous reactor cores via a pre-computed eigenvalue-dependent response coefficient (function) library. Recently, a high order perturbation method was developed to significantly improve the efficiency of the library generation method. In that work, the method's accuracy and efficiency was tested in a small PWR benchmark problem. This paper extends the application of the perturbation method to include problems typical of the other water reactor cores such as BWR and CANDU bundles. It is found that the response coefficients predicted by the perturbation method for typical BWR bundles agree very well with those directly computed by the Monte Carlo method. The average and maximum relative errors in the surface-to-surface response coefficients are 0.02%-0.05% and 0.06%-0.25%, respectively. For CANDU bundles, the corresponding quantities are 0.01%-0.05% and 0.04% -0.15%. It is concluded that the perturbation method is highly accurate and efficient with a wide range of applicability. (authors)
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Fast reactor 3D core and burnup analysis using VESTA
Burnup analyses using the VESTA code have been performed on a MOX-fuelled fast reactor model as specified by an IAEA computational benchmark. VESTA is a relatively new code that has been used for burnup credit calculations and thermal reactor models, but not typically for fast reactor applications. The detailed input and results of the IAEA benchmark provides an opportunity to gauge the use of VESTA in a fast reactor application. VESTA employs an ultra-fine multi-group binning approach that accelerates Monte Carlo burnup calculations. Using VESTA to compute the end of cycle (EOC) power fractions by enrichment zone showed agreement with the published values within 5%. When comparing the ultra-fine multi-group binning approach to the tally-based approach, EOC isotopic masses also agree within 5%. Using the ultra-fine multi-group binning approach, we obtain a wall-time speedup factor of 35 when compared to the tally-based approach for computing a k{sub eff} eigenvalue with burnup problem. The authors conclude the use of VESTA's ultra-fine multi-group binning approach with Monte Carlo transport performs accurate depletion calculations for this fast reactor benchmark. (authors)
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Sizing an external-fueled in-core thermionic reactor.
Parametric studies on sizing of external-fueled in-core thermionic reactors are presented. Reactor physics results obtained for a variety of fuel element designs are used as a basis for nuclear criticality, power distribution, and control worth design. Thermionic performance results for a single fuel element for several sets of operating conditions are presented. An algorithm combining the electrical and reactor physics results in a form amenable to preliminary systems analysis is presented.
Nakashima, A. M.↗
Gas-core rocket reactors - A new look
Open cycle gas core nuclear rocket engine, considering reactor critical experiments and cold and hot flow tests
Ragsdale, R. G.↗