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At least 1,261 records · Page 70
All sourcesNASA NTRSDOE OSTI
DOE OSTI2015

The prospect of uranium nitride (UN) and mixed nitride fuel (UN-PuN) for pressurized water reactor

Design study of small Pressurized Water Reactors (PWRs) core loaded with uranium nitride fuel (UN) and mixed nitride fuel (UN-PuN), Pa-231 as burnable poison, and Americium has been performed. Pa-231 known as actinide material, have large capture cross section and can be converted into fissile material that can be utilized to reduce excess reactivity. Americium is one of minor actinides with long half life. The objective of adding americium is to decrease nuclear spent fuel in the world. The neutronic analysis results show that mixed nitride fuel have k-inf greater than uranium nitride fuel. It is caused by the addition of Pu-239 in mixed nitride fuel. In fuel fraction analysis, for uranium nitride fuel, the optimum volume fractions are 45% fuel fraction, 10% cladding and 45% moderator. In case of UN-PuN fuel, the optimum volume fractions are 30% fuel fraction, 10% cladding and 60% coolant/ moderator. The addition of Pa-231 as burnable poison for UN fuel, enrichment U-235 5%, with Pa-231 1.6% has k-inf more than one and excess reactivity of 14.45%. And for mixed nitride fuel, the lowest value of reactivity swing is when enrichment (U-235+Pu) 8% with Pa-231 0.4%, the excess reactivity value 13,76%. The fuel pin analyze for the addition of Americium, the excess reactivity value is lower than before, because Americium absorb the neutron. For UN fuel, enrichment U-235 8%, Pa-231 1.6% and Am 0.5%, the excess reactivity is 4.86%. And for mixed nitride fuel, when enrichment (U-235+Pu) 13%, Pa-231 0.4% and Am 0.1%, the excess reactivity is 11.94%. For core configuration, it is better to use heterogeneous than homogeneous core configuration, because the radial power distribution is better.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

Field experimental evidence shows that self-interest attracts more sunlight

This study examines how messaging approaches in a prosocial intervention can influence not only the effectiveness of the intervention but also, contagion afterward. Our investigation focuses on leveraging two motivations for solar adoption: self-interest and prosocial. Using data from a natural field experiment in 29 municipalities containing 684,000 people, we find that self-interest messaging is twice as effective in inducing solar adoption both during and after the intervention. Adoptions under self-interest messaging have 10% higher net present value, but prosocial messaging increases the likelihood that adopters recommend solar to their friends and neighbors. Income moderates the effectiveness of self-interest messaging, performing much better in high-income communities than low- and moderate-income communities. There was no significant difference across income groups for prosocial messaging. These results provide guidance to policy makers aiming to encourage prosocial behavior across all income groups.

14 SOLAR ENERGY↗
DOE OSTI2017

Two-Dimensional Neutronic and Fuel Cycle Analysis of the Transatomic Power Molten Salt Reactor

This status report presents the results from the first phase of the collaboration between Transatomic Power Corporation (TAP) and Oak Ridge National Laboratory (ORNL) to provide neutronic and fuel cycle analysis of the TAP core design through the Department of Energy Gateway for Accelerated Innovation in Nuclear, Nuclear Energy Voucher program. The TAP design is a molten salt reactor using movable moderator rods to shift the neutron spectrum in the core from mostly epithermal at beginning of life to thermal at end of life. Additional developments in the ChemTriton modeling and simulation tool provide the critical moderator-to-fuel ratio searches and time-dependent parameters necessary to simulate the continuously changing physics in this complex system. Results from simulations with these tools show agreement with TAP-calculated performance metrics for core lifetime, discharge burnup, and salt volume fraction, verifying the viability of reducing actinide waste production with this design. Additional analyses of time step sizes, mass feed rates and enrichments, and isotopic removals provide additional information to make informed design decisions. This work further demonstrates capabilities of ORNL modeling and simulation tools for analysis of molten salt reactor designs and strongly positions this effort for the upcoming three-dimensional core analysis.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

Neutronics analyses for the conceptual design of the SNS Second Target Station

The Spallation Neutron Source, in operation at the Oak Ridge National Laboratory since 2006, was designed to allow the addition of a second target station and an upgrade of the accelerator proton power. Both upgrades are now underway. Here we describe the evolution of the design of the target of the second target station with the emphasis on the effects of the proton beam footprint on the energy deposition in the target, stresses induced by the pulsed operation, and the importance of the residual heat. The moderator configurations and their optimization are discussed. With the utilization of pure parahydrogen moderators, small neutron beam cross-sections, and the specific optimization, neutron beams of the second target station will achieve exceptionally high peak brightness and time-averaged brightness.

72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS↗
DOE OSTI2018

Rooftop Energy Potential of Low Income Communities in America REPLICA

The Rooftop Energy Potential of Low Income Communities in America REPLICA data set provides estimates of residential rooftop solar technical potential at the tract-level with emphasis on estimates for Low and Moderate Income LMI populations. In addition to technical potential REPLICA is comprised of 10 additional datasets at the tract-level to provide socio-demographic and market context. The model year vintage of REPLICA is 2015. The LMI solar potential estimates are made at the tract level grouped by Area Median Income AMI income tenure and building type. These estimates are based off of LiDAR data of 128 metropolitan areas statistical modeling and ACS 2011-2015 demographic data. The remaining datasets are supplemental datasets that can be used in conjunction with the technical potential data for general LMI solar analysis planning and policy making. The core dataset is a wide-format CSV file seeds_ii_replica.csv that can be tagged to a tract geometry using the GEOID or GISJOIN fields. In addition users can download geographic shapefiles for the main or supplemental datasets. This dataset was generated as part of the larger NREL-led SEEDSII Solar Energy Evolution and Diffusion Studies project and specifically for the NREL technical report titled Rooftop Solar Technical Potential for Low-to-Moderate Income Households in the United States by Sigrin and Mooney 2018. This dataset is intended to give researchers planners advocates and policy-makers access to credible data to analyze low-income solar issues and potentially perform cost-benefit analysis for program design. To explore the data in an interactive web mapping environment use the NREL SolarForAll app.

14 SOLAR ENERGY↗
DOE OSTI2012

MCNP LWR Core Generator

The reactor core input generator allows for MCNP input files to be tailored to design specifications and generated in seconds. Full reactor models can now easily be created by specifying a small set of parameters and generating an MCNP input for a full reactor core. Axial zoning of the core will allow for density variation in the fuel and moderator, with pin-by-pin fidelity, so that BWR cores can more accurately be modeled. LWR core work in progress: (1) Reflectivity option for specifying 1/4, 1/2, or full core simulation; (2) Axial zoning for moderator densities that vary with height; (3) Generating multiple types of assemblies for different fuel enrichments; and (4) Parameters for specifying BWR box walls. Fuel pin work in progress: (1) Radial and azimuthal zoning for generating further unique materials in fuel rods; (2) Options for specifying different types of fuel for MOX or multiple burn assemblies; (3) Additional options for replacing fuel rods with burnable poison rods; and (4) Control rod/blade modeling.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

Iso-geometric analysis for neutron diffusion problems

Iso-geometric analysis can be viewed as a generalisation of the finite element method. It permits the exact representation of a wider range of geometries including conic sections. This is possible due to the use of concepts employed in computer-aided design. The underlying mathematical representations from computer-aided design are used to capture both the geometry and approximate the solution. In this paper the neutron diffusion equation is solved using iso-geometric analysis. The practical advantages are highlighted by looking at the problem of a circular fuel pin in a square moderator. For this problem the finite element method requires the geometry to be approximated. This leads to errors in the shape and size of the interface between the fuel and the moderator. In contrast to this iso-geometric analysis allows the interface to be represented exactly. It is found that, due to a cancellation of errors, the finite element method converges more quickly than iso-geometric analysis for this problem. A fuel pin in a vacuum was then considered as this problem is highly sensitive to the leakage across the interface. In this case iso-geometric analysis greatly outperforms the finite element method. Due to the improvement in the representation of the geometry iso-geometric analysis can outperform traditional finite element methods. It is proposed that the use of iso-geometric analysis on neutron transport problems will allow deterministic solutions to be obtained for exact geometries. Something that is only currently possible with Monte Carlo techniques. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2012

Design of a proteus lattice representative of a burnt and fresh fuel interface at power conditions in light water reactors

The research program LIFE (Large-scale Irradiated Fuel Experiment) between PSI and Swissnuclear has been started in 2006 to study the interaction between large sets of burnt and fresh fuel pins in conditions representative of power light water reactors. Reactor physics parameters such as flux ratios and reaction rate distributions ({sup 235}U and {sup 238}U fissions and {sup 238}U capture) are calculated to estimate an appropriate arrangement of burnt and fresh fuel pins within the central element of the test zone of the zero-power research reactor PROTEUS. The arrangement should minimize the number of burnt fuel pins to ease fuel handling and reduce costs, whilst guaranteeing that the neutron spectrum in both burnt and fresh fuel regions and at their interface is representative of a large uniform array of burnt and fresh pins in the same moderation conditions. First results are encouraging, showing that the burnt/fresh fuel interface is well represented with a 6 x 6 bundle of burnt pins. The second part of the project involves the use of TSUNAMI, CASMO-4E and DAKOTA to perform parametric and optimization studies on the PROTEUS lattice by varying its pitch (P) and fraction of D{sub 2}O in moderator (F{sub D2O}) to be as representative as possible of a power light water reactor core at hot full power conditions at beginning of cycle (BOC). The parameters P and F{sub D2O} that best represent a PWR at BOC are 1.36 cm and 5% respectively. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2012

Feasibility study on the thorium fueled boiling water breeder reactor

The feasibility of (Th,U)O 2 fueled, boiling water breeder reactor based on conventional BWR technology has been studied. In order to determine the potential use of water cooled thorium reactor as a competitive breeder, this study evaluated criticality, breeding and void reactivity coefficient in response to changes made in MFR and fissile enrichments. The result of the study shows that while using light water as moderator, low moderator to fuel volume ratio (MFR=0.5), it was possible to breed fissile fuel in negative void reactivity condition. However the burnup value was lower than the value of the current LWR. On the other hand, heavy water cooled reactor shows relatively wider feasible breeding region, which lead into possibility of designing a core having better neutronic and economic performance than light water with negative void reactivity coefficient. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

Effect of spray parameter on containment depressurization during LOCA in KAPP 3 and 4, 700 MWE IPHWR

KAPP 3 and 4 is an Indian Pressurized Heavy Water Reactor (IPHWR) of 700 MWe capacities. It is a pressure tube type reactor with heavy water as moderator and coolant and natural Uranium Dioxide as fuel. It consists of 392 horizontal fuel channel assemblies and surrounded by three separate water systems i.e. primary coolant, moderator and calandria vault water system. Containment of Indian PHWR is an ultimate barrier, which is designed to envelope whole reactor systems, to prevent the spread of active air-borne fission products in accident condition. Containment Spray System has been provided for energy as well as activity removal from the Containment system. This paper discusses about the studies done to assess the effect of spray parameters such as spray flow rate, droplets diameter and height of fall on containment peak pressure and temperature, long term containment depressurization and energy removal from the containment during Loss of Coolant Accident (LOCA). The spray flow rate and droplets diameter play an important role in removing residual energy from containment atmosphere, which influences depressurization of containment. It is obvious that faster depressurization of containment during postulated LOCA helps in limiting radiological consequences. From radiological considerations, droplets diameter is required to be kept to the lowest practically possible value and flow rate of spray should be high. Spray water droplets fall height governs the exposure time of droplets, which is the direct indication of energy removal rate. However, it is observed from the sensitivity studies that for a height of spray droplet fall more than 16.5 m, for the range of spray water flow rate and droplets sizes considered in the analyses, there is no significant change in heat removal. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

Treatment of Irradiated Graphite to meet Acceptance Criteria for Waste Disposal: A New IAEA Collaborative Research Program - 12443

World-wide, more than 250,000 tonnes of irradiated graphite have arisen through commercial nuclear-power operations and from military production reactors. Whilst most nations responsible for the generation of this material have in mind repository disposal alongside other radwaste, the lack of progress in this regard has led in some cases to difficulties where, for example, the site of an existing graphite-moderated reactor is required for re-utilisation. In any case, graphite as a radwaste stream has unique chemical and physical properties which may lend itself to more radical and innovative treatment and disposal options, including the recovery of useful isotopes and also recycling within the nuclear industry. Such aspects are important in making the case for future graphite-moderated reactor options (for example, High-Temperature Reactors planned for simultaneous power production and high-grade heat sources for such applications as hydrogen production for road fuel). A number of initiatives have taken place since the mid 1990s aimed at exploring such alternative strategies and, more recently, improving technology offers new options at all stages of the dismantling and disposal process. A new IAEA Collaborative Research Program aims to build upon the work already done and the knowledge achieved, in order to identify the risks and uncertainties associated with alternative options for graphite disposal, along with cost comparisons, thus enabling individual Member States to have the best-available information at their disposal to configure their own programs. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

Optimization of Neutron Spectrum in Northwest Beam Tube of Tehran Research Reactor for BNCT, by MCNP Code

In order to gain the neutron spectrum with proper components specification for BNCT, it is necessary to design a Beam Shape Assembling (BSA), include of moderator, collimator, reflector, gamma filter and thermal neutrons filter, in front of the initial radiation beam from the source. According to the result of MCNP4C simulation, the Northwest beam tube has the most optimized neuron flux between three north beam tubes of Tehran Research Reactor (TRR). So, it has been chosen for this purpose. Simulation of the BSA has been done in four above mentioned phases. In each stage, ten best configurations of materials with different length and width were selected as the candidates for the next stage. The last BSA configuration includes of: 78 centimeters of air as an empty space, 40 centimeters of Iron plus 52 centimeters of heavy-water as moderator, 30 centimeters of water or 90 centimeters of Aluminum-Oxide as a reflector, 1 millimeters of lithium (Li) as thermal neutrons filter and finally 3 millimeters of Bismuth (Bi) as a filter of gamma radiation. The result of Calculations shows that if we use this BSA configuration for TRR Northwest beam tube, then the best neutron flux and spectrum will be achieved for BNCT. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Nuclear Fuel Traces Definition in Storage Ponds of Research VVR-2 and OR Reactors in NRC 'Kurchatov Institute'

The Gas Plant complex is the experimental base of the Institute of Nuclear Reactors, which is part of the Kurchatov Institute. In 1954 the commissioning of the first Soviet water-cooled water-moderated research reactor VVR-2 on enriched uranium, and until 1983 the complex operated two research water-cooled water-moderated reactors 3 MW (VVR-2) and 300 kW (OR) capacity, which were dismantled in connection with the overall upgrades of the complex. The complex has three storage ponds in the reactor building. They are sub-surface vessels filled with water (the volume of water in each is about 6 m{sup 3}). In 2007-2013 the spent nuclear fuel from storages was removed for processing to 'Mayk'. Survey of Storage Ponds by Underwater Collimated Spectrometric System shows a considerable layer of slime on the bottom of ponds and traces of spent nuclear fuel in one of the storage. For determination qualitative and the quantitative composition of radionuclide we made complex α-, β-, γ- spectrometric research of water and bottom slimes from Gas Plant complex storage ponds. We found the spent nuclear fuel in water and bottom slime in all storage ponds. Specific activity of radionuclides in the bottom slime exceeded specific activity of radionuclides in the ponds water and was closed to levels of high radioactive waste. Analysis of the obtained data and data from earlier investigation of reactor MR storage ponds showed distinctions of specific activity of uranium and plutonium radionuclides. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Multipurpose epithermal neutron beam on new research station at MARIA research reactor in Swierk-Poland

MARIA reactor is an open-pool research reactor what gives the chance to install uranium fission converter on the periphery of the core. It could be installed far enough not to induce reactivity of the core but close enough to produce high flux of fast neutrons. Special design of the converter is now under construction. It is planned to set the research stand based on such uranium converter in the near future: in 2015 MARIA reactor infrastructure should be ready (preparation started in 2013), in 2016 the neutron beam starts and in 2017 opening the stand for material and biological research or for medical training concerning BNCT. Unused for many years, horizontal channel number H2 at MARIA research rector in Poland, is going to be prepared as a part of unique stand. The characteristics of the neutron beam will be significant advantage of the facility. High flux of neutrons at the level of 2x10{sup 9} cm{sup -2}s{sup -1} will be obtainable by uranium neutron converter located 90 cm far from the reactor core fuel elements (still inside reactor core basket between so called core reflectors). Due to reaction of core neutrons with converter U{sub 3}Si{sub 2} material it will produce high flux of fast neutrons. After conversion neutrons will be collimated and moderated in the channel by special set of filters and moderators. At the end of H2 channel i.e. at the entrance to the research room neutron energy will be in the epithermal energy range with neutron intensity at least at the level required for BNCT (2x10{sup 9} cm{sup -2}s{sup -1}). For other purposes density of the neutron flux could be smaller. The possibility to change type and amount of installed filters/moderators which enables getting different properties of the beam (neutron energy spectrum, neutron-gamma ratio and beam profile and shape) is taken into account. H2 channel is located in separate room which is adjacent to two other empty rooms under the preparation for research laboratories (200 m2). It is planned to create fully equipped complex facility possible to perform various experiments on the intensive neutron beam. Epithermal neutron beam enables development across the full spectrum of materials research for example shielding concrete tests or electronic devices construction improvement. Due to recent reports on the construction of the accelerator for the Boron Neutron Capture Therapy (BNCT) it has the opportunity to become useful and successful method in the fight against brain and other types of cancers not treated with well known medical methods. In Europe there is no such epithermal neutron source which could be used throughout the year for training and research for scientist working on BNCT what makes the stand unique in Europe. Also our research group which specializes in mixed radiation dosimetry around nuclear and medical facilities would be able to carry out research on new detectors and methods of measurements for radiological protection and in-beam (therapeutic) dosimetry. Another group of scientists from National Centre for Nuclear Research, where MARIA research reactor is located, is involved in research of gamma detector systems. There is an idea to develop Prompt-gamma Single Photon Emission Computed Tomography (Pg- SPECT). This method could be used as imaging system for compounds emitting gamma rays after nuclear reaction with thermal neutrons e.g. for boron concentration in BNCT. Inside the room, where H2 channel is located, there is another horizontal channel - H1 which is also unused. Simultaneously with the construction of the H2 stand it will be possible to create special pneumatic horizontal mail inside the H1 channel for irradiation material samples in the vicinity of the core i.e. in the distal part of the H1 channel. It might expand the scope of research at the planned neutron station. Secondly it is planned to equip both stands with moveable positioning system, video system and facilities to perform animal experiments (anaesthesia, vital signs control, imaging devices, positioning). These all above make constructed station unique in the world (uranium fission converter-based beam) and the only one of such intense neutron beam in the Europe. Moreover implementation of the station would allow the development of research on a number of issues for researchers from all over the Europe. One of very important advantages of the station is undisturbed exploitation of the reactor and other vertical and horizontal channels. MARIA reactor operates 6000 hours per year and that amount of time will be achievable for research on the neutron station. It have to be underlined that new neutron station will work parallel to all another ventures. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Property analysis and advanced manufacturing technique development for light water reactor annular composite fuel

The University of Florida (UF) has been developing a composite UO{sub 2}-Diamond Fuel design for water-cooled reactors that offers greatly increased thermal conductivity as measured at BOL. Currently, Diamond composite fuel pellets are being irradiated in the ATR to determine whether the thermal conductivity increase is maintained as a function of burnup. UF has further increased the benefits of this increased thermal conductivity of the additive UO{sub 2} fuel by implementing an annular pellet design. The central core material for our annular fuel design must have a high thermal conductivity, a high melting point, a low absorption cross section, a low coefficient of thermal expansion and a neutron thermal moderating capability, which will enhanced the reactivity and performance of the cored pellet. The chosen ZrC core material has a very high melting temperature of 3,700 Celsius degrees, a high thermal conductivity, a good moderating effect, and a low value of thermal expansion. UF has successfully manufactured UO{sub 2} annular pellets. The University of Florida (UF) team believes such an annular pellet can eliminate pellet cracking, reduce fission gas release by over 50% and reduce the maximum pellet temperature by between 300 to 600 Celsius degrees. It is believed that such a pellet design would tolerate a loss of coolant accident (LOCA). (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Nuclear Fuel Traces Definition in Storage Ponds of Research VVR-2 and OR Reactors in NRC 'Kurchatov Institute' - 16022

The Gas Plant complex is the experimental base of the Institute of Nuclear Reactors, which is part of the Kurchatov Institute. In 1954 the commissioning of the first Soviet water-cooled water-moderated research reactor VVR-2 on enriched uranium, and until 1983 the complex operated two research water-cooled water-moderated reactors 3 MW (VVR-2) and 300 kW (OR) capacity, which were dismantled in connection with the overall upgrades of the complex. The complex has three storage ponds in the reactor building. They are sub-surface vessels filled with water (the volume of water in each is about 6 m{sup 3}). In 2007-2013 the spent nuclear fuel from storages was removed for processing to 'Mayak'. Survey of Storage Ponds by Underwater Collimated Spectrometric System shows a considerable layer of slime on the bottom of ponds and traces of spent nuclear fuel in one of the storage pond. For determination qualitative and the quantitative composition of radionuclide we made complex α-, β-, γ - spectrometric research of water and bottom slimes from Gas Plant complex storage ponds. We found the spent nuclear fuel in water and bottom slime in all storage ponds. Specific activity of radionuclides in the bottom slime exceeded specific activity of radionuclides in the ponds water and was closed to levels of high radioactive waste. Analysis of the obtained data and data from earlier investigation of reactor MR storage ponds showed distinctions of specific activity of uranium and plutonium radionuclides. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

An Investigation of MCNP6.1 Beryllium Oxide S(α,β) Cross Sections

In MCNP6.1, materials are constructed by identifying the constituent isotopes (or elements in a few cases) individually. This list selects the corresponding microscopic cross sections calculated from the free-gas model to create the material macroscopic cross sections. The free-gas model and the corresponding material macroscopic cross sections assume that the interactions of atoms do not affect the nuclear cross sections. With certain compounds, the free-gas model cross sections are not sufficient because molecular effects or crystalline structure effects do affect the neutron scattering cross sections at low energies. These effects are included in the MCNP6.1 thermal scattering kernels, or S(α,β) thermal neutron scattering data. To treat a particular isotope (or isotopes) as a molecular compound in the thermal regime, one or more of the available S(α,β) data sets is selected (with the MT card). As MCNP6.1 transports particles, the free-gas model cross sections are used down to the energy where S(α,β) data are available. At that point, the S(α,β) cross sections automatically overrides the free-gas cross sections. Most of the S(α,β) files have a maximum energy of 10 eV. In MCNP6.1, a binary compound may have one or two S(α,β) libraries. Beryllium oxide (BeO) is one of the materials that has two S(α,β) libraries one for the beryllium S(α,β) cross sections (be-o) and one for the oxygen S(α,β) cross sections (o-be). From its moderating properties, it is apparent that the S(α,β) library for the beryllium must be included to accurately model beryllium oxide. The necessity of the S(α,β) library for oxygen is less clear because oxygen is a relatively poor moderator. This paper investigates the effect of using only the be-o S(α,β) library or using both be-o and o-be S(α,β) libraries. The objective is to provide guidance on the use of the beryllium oxide S(α,β) libraries in criticality safety analyses. (authors)

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2016

Evaluation of B and W UO{sub 2}/ThO{sub 2} VIII Experimental Core: Criticality and Thermal Disadvantage Factor Analysis

In the framework of the OECD/NEA International Reactor Physics Experiment (IRPHE) Project, an evaluation of core VIII of the Babcock and Wilcox (B and W) Spectral Shift Control Reactor (SSCR), critical experiment program was performed. The SSCR concept, moderated and cooled by a variable mixture of heavy and light water, envisaged changing of the thermal neutron spectrum during the operation to encourage breeding and to sustain the core criticality. Core VIII contained 2188 fuel rods with 93% enriched UO{sub 2}-ThO{sub 2} fuel in a moderator mixture of heavy and light water. The criticality experiment and measurements of the thermal disadvantage factor were evaluated. Benchmark criticality results using different cross-section libraries are reported. ENDF/B-VI.6 calculation was run using the S(α, β) model based on ENDF/B-VI.3 data while ENDF/B-VII.0 and ENDF/BVII.1 used an S(α, β) model based on ENDF/B-VII.0 data. JEFF calculations were run using the S(α, β) model based on JEFF 3.1.1 data except for the aluminum isotope, for which the S(α, β) model was missing. The maximum absolute deviation from the benchmark model was obtained using the JEFF-3.1.1 cross sections while closer results were obtained using ENDF/ B-VII.0 and ENDF/B-VII.1. Only ENDF/B-VII.1 and ENDF/B-VII.0 results are inside the 3σ uncertainty band (+2.8σ and 2.4σ, respectively), having deviations on the order of 300 pcm. The maximum deviation is -827 pcm (-6.1σ), using JEFF.3.1.1. Sample calculations of the TDF benchmark were also performed, studying the effects of the main nuclear data libraries. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗