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At least 163 records · Page 9
All sourcesNASA NTRSDOE OSTI
DOE OSTI2019

Architecture for the Performance of Nuclear Fuel Depletion Calculations

Computational analysis of the depletion of nuclear fuel is crucial to the accurate modeling of the time-dependent behavior of nuclear reactors. The reactor physics application MAMMOTH is modified to compute radioactive decay and neutron transmutation in order to model physical processes which occur within nuclear reactors as reactor fuel is burned. The ability to read decay and cross-section data files of other codes are also added and the addition of a standardized decay and transmutation XML file format for MAMMOTH, known as ISOXML. Comparisons between MAMMOTH and the Oak Ridge Isotope Generation (ORIGEN) code are performed to demonstrate the improved precision possible with MAMMOTH. In addition, several tests are run to demonstrate the ability of MAMMOTH to model decay and transmutation processes involving hundreds of isotopes. Future work will focus on solving depletion problems with multiple neutron energy groups and the implementation of predictor-corrector methods, several of which are described in this work.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2022

Chain Information Table Format (Version 1.1)

Materials present in a radiation environment typically undergo transmutation. The process whereby the material is transmuted is known by many terms including production-depletion, radiochemistry, activation and burn-up. In nature, this process is not limited and may encompass the entire table of isotopes. In the finite world of computer simulation, this process is bounded by the limitations of the data whereby isotopes are transmuted and the limitations of the physics, i.e. particle transport, used by the simulation.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Second Target Station High-Fidelity Target Activation Comparison

The development of the Second Target Station (STS) target system at the Spallation Neutron Source (SNS) at Oak Ridge National Laboratory (ORNL) is well underway. The target system at STS consists of a rotating target disk that contains 21 segments of tungsten clad in tantalum clad in steel. A key aspect of the design of the target system is to account for the delayed heating and material damage caused by the delayed dose from decaying radionuclides. These radionuclides are a product of either spallation reactions or transmutation of the nuclei in the target system. These radionuclides build up in the target system components over the lifetime of the facility, and the radiation that is emitted can deposit energy in the components causing significant component heating and material damage. Monte Carlo N-Particle (MCNP) Version 6.2 transports the various particle species and calculates the spallation products and neutron fluxes throughout the target system. These spallation products and neutron fluxes along with the material definition of each component are relayed to the CINDER2008 transmutation code to calculate the radionuclide inventories and the corresponding decay gamma emission spectra. MCNP6.2 coupled with CINDER2008 is the computational method-of-choice for the analysis discussed in the following sections of this report. The analysis focuses on validating major assumptions in calculating the radionuclide inventory in the STS target system: all of the target segments are fresh, unirradiated material when the protons are incident on the segment, the average of the 21 segments of the target is sufficient to represent a single segment, and that averaging the proton pulse structure over time does not significantly affect the radionuclide inventory. The position-averaged, high-fidelity, and single-tally computational methods are used to validate the assumptions and provide a point of comparison to evaluate how the assumptions impact the radionuclide inventories. A more detailed explanation of the three computational methods is provided in Section 2. The position-averaged and single-tally methods are less computationally expensive when compared with the high-fidelity method where 54,000 individual calculations are needed to calculate 1 hr of STS operation. Section 3 details the comparison of the three methods to show that the assumptions made in the position-averaged method do not significantly impact the radionuclide inventory after 1 hr of operation. The discussions and results in this report are for 1 hr of operation. Due to the computational cost associated with calculating the transmutation and activation using the high-fidelity method, only 1 hr of operation has been calculated. The discrepancies observed after 1 hr of operation are not extrapolated out to longer operational times, and this report does not address how the discrepancies between the computational methods may manifest for longer operational periods.

43 PARTICLE ACCELERATORS↗
DOE OSTI2025

Impact of Time Dependent Reactor and Sensor Physics on Core Power Synthesis

Online synthesis of the power distribution is critical in the operation and control of nuclear power reactors to ensure that the core is operating within safety margins, and to provide essential knowledge associated with the burnup of the fuel. In light water reactors (LWRs), power synthesis is achieved by using some a priori knowledge of the state of the reactor core and updating based on the signals coming from in-core sensors—namely, self-powered neutron detectors (SPNDs). This report aims to study the effects of fuel burnup and sensor degradation on the ability to accurately synthesize the power distribution in a LWR. Several modeling tools were used to simulate power synthesis based on the responses of SPNDs, with emitters made out of Rh or V. A representative pressurized water reactor low-enriched uranium (LEU) core was modeled using the Polaris/Purdue Advanced Reactor Core Simulator (PARCS) approach. The Monte Carlo N-Particle Transport 6 (MCNP6) code was used, as well, to calculate response functions between different segments of fuel to individual SPNDs; this is a crucial parameter for power synthesis. The Oak Ridge Isotope GENeration (ORIGEN) package in the Standardized Computer Analyses for Licensing Evaluation (SCALE) code was used to model the time-dependent isotopic transmutation in the SPND emitters. All these data were fed into a custom code that enacted the point-based iterative (PBI) method to simulate power synthesis. Developmental work was also performed on high-fidelity SPND models in the GEometry ANd Tracking 4 (Geant4) code, which enables higher-accuracy modeling of the current responses from SPNDs. In this work, five sets of time-dependent power synthesis test cases were conducted. In these test cases, systematic changes in the input conditions enabled an analysis of the effect of (1) slightly inaccurate a priori power distribution assumptions with respect to fuel burnup, (2) highly inaccurate a priori assumptions with respect to fuel burnup (such that burnup is not included in the a priori assumed distribution), and (3) differences between Rh and V SPNDs in terms of downstream consequences of the transmutation in the emitters. The authors discovered that one may permissibly have slightly inaccurate a priori assumptions of the fuel burnup (such that the level of burnup may be slightly under- or over-approximated by the accumulated burnup in approximately 9.3 full power days), but to not account for burnup at all in the a priori assumption leads to severe levels of error, approaching 25% at maximum. The authors also discovered that V SPNDs are extraordinarily robust in the low-enriched uranium fuel cycle considered in this modeling work, whereas Rh SPNDs undergo significant transmutation that can result in large errors in the synthesized power distribution.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2026

Impact of Time-Dependent Reactor and Sensor Physics on Core Power Synthesis (Rev.1)

Online synthesis of power distribution is critical in the operation and control of nuclear power reactors to ensure that the core is operating within safety margins and to provide essential knowledge associated with the burnup of the fuel. In light-water reactors, power synthesis is achieved by using some a priori knowledge of the state of the reactor core and updating based on the signals coming from in-core sensors—namely, self-powered neutron detectors (SPNDs). This report examines the effects of fuel burnup and sensor degradation on the ability to accurately synthesize the power distribution in a pressurized water reactor (PWR), considering the typical low-enriched uranium (LEU, 3%-5% enrichment) fuel cycle as well as the higher enrichment LEU+ (5%-8% enrichment) fuel cycle. Several modeling tools were used to simulate power synthesis based on the responses of SPNDs, with emitters made out of Rh or V. A representative PWR LEU core was modeled using the Polaris/Purdue Advanced Reactor Core Simulator (PARCS) approach. The Monte Carlo N-Particle Transport 6 (MCNP6) code was used as well to calculate response functions between different segments of fuel to individual SPNDs; this is a crucial parameter for power synthesis. The Oak Ridge Isotope GENeration (ORIGEN) package in the Standardized Computer Analyses for Licensing Evaluation (SCALE) code was used to model the time-dependent isotopic transmutation in the SPND emitters. All these data were fed into a custom code that enacted the point-based iterative method to simulate power synthesis. Developmental work was also performed on high-fidelity SPND models in the GEometry ANd Tracking 4 (Geant4) code, which enables higher-accuracy modeling of the current responses from SPNDs. In this work, five sets of time-dependent power synthesis test cases were conducted. In these test cases, systematic changes in the input conditions enabled an analysis of the effect of (1) slightly inaccurate a priori power distribution assumptions with respect to fuel burnup, (2) highly inaccurate a priori power distribution assumptions with respect to fuel burnup (such that burnup is not included in the a priori assumed distribution), and (3) differences between Rh and V SPNDs in terms of downstream consequences of the transmutation in the emitters and the extended nature of the LEU+ fuel cycle in comparison with LEU. The authors discovered that one may permissibly have slightly inaccurate a priori assumptions of the fuel burnup (such that the level of burnup may be slightly underapproximated or overapproximated by the accumulated burnup in approximately 9.3 full power days), but to not account for burnup at all in the a priori assumptions leads to severe levels of error, approaching 25% at maximum (for LEU). The authors also discovered that V SPNDs are extraordinarily robust in both the LEU and LEU+ fuel cycles considered in this modeling work, whereas Rh SPNDs undergo significant transmutation that can result in large errors in the synthesized power distribution.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2015

MAGNESIUM PRECIPITATION AND DIFUSSION IN Mg+ ION IMPLANTED SILICON CARBIDE

As a candidate material for fusion reactor applications, silicon carbide (SiC) undergoes transmutation reactions under high-energy neutron irradiation with magnesium as the major metallic transmutant; the others include aluminum, beryllium and phosphorus in addition to helium and hydrogen gaseous species. Calculations by Sawan et al. predict that at a dose of ~100 dpa (displacements per atom), there is ~0.5 at.% Mg generated in SiC. The impact of these transmutants on SiC structural stability is currently unknown. This study uses ion implantation to introduce Mg into SiC. Multiaxial ion-channeling analysis of the as-produced damage state indicates a lower dechanneling yield observed along the <100> axis. The microstructure of the annealed sample was examined using high-resolution scanning transmission electron microscopy. The results show a high concentration of likely non-faulted tetrahedral voids and possible stacking fault tetrahedra near the damage peak. In addition to lattice distortion, dislocations and intrinsic and extrinsic stacking faults are also observed. Magnesium in 3C–SiC prefers to substitute for Si and it forms precipitates of cubic Mg2Si and tetragonal MgC2. The diffusion coefficient of Mg in 3C–SiC single crystal at 1573 K has been determined to be 3.8 ± 0.4E-19 m2/s.

diffusion↗
DOE OSTI2016

Accelerator Reactor Coupling for Energy Production in Advanced Nuclear Fuel Cycles

This article is a review of several accelerator–reactor interface issues and nuclear fuel cycle applications of acceleratordriven subcritical systems. The systems considered here have the primary goal of energy production, but that goal is accomplished via a specific application in various proposed nuclear fuel cycles, such as breed-and-burn of fertile material or burning of transuranic material. Several basic principles are reviewed, starting from the proton beam window including the target, blanket, reactor core, and up to the fuel cycle. We focus on issues of interest, such as the impact of the energy required to run the accelerator and associated systems on the potential electricity delivered to the grid. Accelerator-driven systems feature many of the constraints and issues associated with critical reactors, with the added challenges of subcritical operation and coupling to an accelerator. Reliable accelerator operation and avoidance of beam trips are critically important. One interesting challenge is measurement of blanket subcriticality level during operation. We also review the potential benefits of accelerator-driven systems in various nuclear fuel cycle applications. Ultimately, accelerator-driven subcritical systems with the goal of transmutation of transuranic material have lower 100,000-year radioactivity than a critical fast reactor with recycling of uranium and plutonium.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Microstructural stability of tantalum-alloyed ferritic-martensitic steel with neutron irradiation to 7.4 dpa at ~490 °C

Specimens of a reduced-activation ferritic-martensitic steel, called CNA1, were irradiated to 7.4 displacements per atom at ~490 °C in the High Flux Isotope Reactor. Vickers hardness and tensile properties revealed consistent softening that is attributed to some recovery during irradiation, with reduced laths and dislocation density. Detailed microstructural characterization by transmission electron microscopy and energy dispersive x-ray spectroscopy showed radiation-induced dislocation loops, few cavities, and slight dissolution of M 23 C 6 and MX precipitates. Partial amorphization was only observed in M 23 C 6 precipitated in matrix. Transmutation somewhat destabilized M 23 C 6 and MX, with a stronger effect on the later one by noticeable depletion of Ta with enriched W, which is consistent with the transmutation calculations. The microstructural evolution and associated strength changes were discussed with the assistance of computational thermodynamics.

36 MATERIALS SCIENCE↗
DOE OSTI2012

Diagnostics for hybrid reactors

The Hybrid Reactor(HR) can be considered an attractive actinide-burner or a fusion assisted transmutation for destruction of transuranic(TRU) nuclear waste. The hybrid reactor has two important subsystems: the tokamak neutron source and the blanket which includes a fuel zone where the TRU are placed and a tritium breeding zone. The diagnostic system for a HR must be as simple and robust as possible to monitor and control the plasma scenario, guarantee the protection of the machine and monitor the transmutation.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2014

Design of radiation resistant metallic multilayers for advanced nuclear systems

Helium implantation from transmutation reactions is a major cause of embrittlement and dimensional instability of structural components in nuclear energy systems. Development of novel materials with improved radiation resistance, which is of the utmost importance for progress in nuclear energy, requires guidelines to arrive at favorable parameters more efficiently. Here, we present a methodology that can be used for the design of radiation tolerant materials. We used synchrotron X-ray reflectivity to nondestructively study radiation effects at buried interfaces and measure swelling induced by He implantation in Cu/Nb multilayers. The results, supported by transmission electron microscopy, show a direct correlation between reduced swelling in nanoscale multilayers and increased interface area per unit volume, consistent with helium storage in Cu/Nb interfaces in forms that minimize dimensional changes. In addition, for Cu/Nb layers, a linear relationship is demonstrated between the measured depth-dependent swelling and implanted He density from simulations, making the reflectivity technique a powerful tool for heuristic material design.

36 MATERIALS SCIENCE↗
DOE OSTI2014

A brief history of design studies on innovative nuclear reactors

In a short period after the success of CP1, many types of nuclear reactors were proposed and investigated. However, soon only a small number of reactors were selected for practical use. Around 1970, only LWRs with small number of CANDUs were operated in the western world, and FBRs were under development. It was about the time when Apollo moon landing was accomplished. However, at the same time, the future of human being was widely considered pessimistic and Limits to Growth was published. In the end of 1970’s the TMI accident occurred and many nuclear reactor contracts were cancelled in USA and any more contracts had not been concluded until recent years. From the reflection of this accident, many Inherent Safe Reactors (ISRs) were proposed, though none of them were constructed. A common idea of ISRs is smallness of their size. Tokyo Institute of Technology (TokyoTech) held a symposium on small reactors, SR/TIT, in 1991, where many types of small ISRs were presented. Recently small reactors attract interest again. The most ideas employed in these reactors were the same discussed in SR/TIT. In 1980’s the radioactive wastes from fuel cycle became a severe problem around the world. In TokyoTech, this issue was discussed mainly from the viewpoint of nuclear transmutations. The neutron economy became inevitable for these innovative nuclear reactors especially small long-life reactors and transmutation reactors.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2019

Elemental Characterization of Neutron-Irradiated Tungsten Using the GD-OES Technique

In reactor-relevant fusion divertor conditions, tungsten (W) will be used as an armor material due to its excellent thermal properties. It will be exposed to impurities from numerous sources, including ion implantation and mixing, neutron transmutation, low-Z plasma-facing-component (PFC) redeposition and codeposition of deuterium and tritium fuel, and trapped helium bubbles. The impurity plasma material–interaction effects are a concern because they can cause gradual degradation of the material and of plasma performance due to dust formation, fuel retention, and even changes to the thermal and mechanical properties of the W armor. It is crucial to measure the amount of impurities in W, and the glow discharge–optical emission spectroscopy (GD-OES) technique is exceptionally well suited for analysis of irradiated samples. GD-OES can measure a sample’s elemental composition by sputtering the surface of the sample, ionizing the eroded material, and measuring the optical emission of the excited atoms. In order for the GD-OES technique to be applied to neutron-irradiated tungsten samples, a mounting system for miniature samples was designed. The sample mounting and centering procedure was successful in measuring the depth distribution of control W and W alloy sample elemental concentrations. These control depth spectra will be used as elemental references for postirradiated samples. The residence time of surface layers was measured, a comparison of signals from different anodes was completed, and the influence of initial surface roughness or nonuniformity was understood. The depth distribution of an arc-welded W-0.4% rhenium (Re) alloy was measured to have a stable Re signal that was distributed evenly in the W matrix. Here, the methods developed here will allow for quantification of impurities and transmutation amounts in neutron-irradiated W. GD-OES is a powerful tool but requires calibration and careful optimization of the parameters to obtain meaningful results.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2015

Self-Sustaining Thorium Boiling Water Reactors

The primary objectives of this project are to: Perform a pre-conceptual design of a core for an alternative to the Hitachi proposed fuel-self- sustaining RBWR-AC, to be referred to as a RBWR-Th. The use of thorium fuel is expected to assure negative void coefficient of reactivity (versus positive of the RBWR-AC) and improve reactor safety; Perform a pre-conceptual design of an alternative core to the Hitachi proposed LWR TRU transmuting RBWR-TB2, to be referred to as the RBWR-TR. In addition to improved safety, use of thorium for the fertile fuel is expected to improve the TRU transmutation effectiveness; Compare the RBWR-Th and RBWR-TR performance against that of the Hitachi RBWR core designs and sodium cooled fast reactor counterparts - the ARR and ABR; and, Perform a viability assessment of the thorium-based RBWR design concepts to be identified along with their associated fuel cycle, a technology gap analysis, and a technology development roadmap. A description of the work performed and of the results obtained is provided in this Overview Report and, in more detail, in the Attachments. The major findings of the study are summarized.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

Radiation Damage Calculation Methodology

This report outlines the consensus radiation damage calculation methodology used by the Neutronics Analysis Group at Idaho National Laboratory. This methodology includes calculation of radiation damage rates via the Monte Carlo N-Particle (MCNP) code, and calculation of radiation damage in mixed materials, such as alloys. A consistent method for calculating the average displacement threshold energy of a mixed material has been established by the Neutronics Analysis Group. No standard method currently exists to account for radiation damage caused by neutron-induced transmutations. Only one documented case addresses transmutation radiation damage in materials containing nickel, and that method was incorporated by the Neutronics Analysis Group into their radiation damage calculation methodology.

36 MATERIALS SCIENCE↗
DOE OSTI2012

Plutonium and minor actinide utilisation in a pebble-bed high temperature reactor

This paper contains results of the analysis of the pebble-bed high temperature gas-cooled PUMA reactor loaded with plutonium and minor actinide (Pu/MA) fuel. Starting from knowledge and experience gained in the Euratom FP5 projects HTR-N and HTR-N1, this study aims at demonstrating the potential of high temperature reactors to utilize or transmute Pu/MA fuel. The work has been performed within the Euratom FP6 project PUMA. A number of different fuel types and fuel configurations have been analyzed and compared with respect to incineration performance and safety-related reactor parameters. The results show the excellent plutonium and minor actinide burning capabilities of the high temperature reactor. The largest degree of incineration is attained in the case of an HTR fuelled by pure plutonium fuel as it remains critical at very deep burnup of the discharged pebbles. Addition of minor actinides to the fuel leads to decrease of the achievable discharge burnup and therefore smaller fraction of actinides incinerated during reactor operation. The inert-matrix fuel design improves the transmutation performance of the reactor, while the 'wallpaper' fuel does not have advantage over the standard fuel design in this respect. After 100 years of decay following the fuel discharge, the total amount of actinides remains almost unchanged for all of the fuel types considered. Among the plutonium isotopes, only the amount of Pu-241 is reduced significantly due to its relatively short half-life. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

On the use of moderating material to enhance the feedback coefficients in SFR cores with high minor actinide content

The use of fine distributed moderating material to enhance the feedback effects and to reduce the sodium void effecting sodium cooled fast reactor cores is described. The influence of the moderating material on the neutron spectrum, the power distribution, and the burnup distribution is shown. The consequences of the use of fine distributed moderating material into fuel assemblies with fuel configurations foreseen for minor actinide transmutation is analyzed and the transmutation efficiency is compared. The degradation of the feedback effects due to the insertion of minor actinides and the compensation by the use of moderating materials is discussed. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2013

Grouped actinide separation in advanced nuclear fuel cycles

Aiming at cleaner waste streams (containing only the short-lived fission products) a partitioning and transmutation (P-T) scheme can significantly reduce the quantities of long-lived radionuclides consigned to waste. Many issues and options are being discussed and studied at present in view of selecting the optimal route. The choice is between individual treatment of the relevant elements and a grouped treatment of all actinides together. In the European Collaborative Project ACSEPT (Actinide recycling by Separation and Transmutation), grouped separation options derived from an aqueous extraction or from a dry pyroprocessing route were extensively investigated. Successful demonstration tests for both systems have been carried out in the frame of this project. The aqueous process called GANEX (Grouped Actinide Extraction) is composed of 2 cycles, a first one to recover the major part of U followed by a co-extraction of Np, Pu, Am, and Cm altogether. The pyro-reprocessing primarily applicable to metallic fuels such as the U-Pu-Zr alloy originally developed by the Argonne National Laboratory (US) in the mid 1980s, has also been applied to the METAPHIX fuels containing up to 5% of minor actinides and 5% of lanthanides (e.g. U{sub 60}Pu{sub 20}-Zr{sub 10}Am{sub 2}Nd{sub 3.5}Y{sub 0.5}Ce{sub 0.5}Gd{sub 0.5}). A grouped actinide separation has been successfully carried out by electrorefining on solid Al cathodes. At present the recovery of the actinides from the alloy formed with Al upon electrodeposition is under investigation, because an efficient P-T cycle requires multiple re-fabrication and re-irradiation. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗