Heterogeneous sodium fast reactor designed for transmuting minor actinide waste isotopes into plutonium fuel

NASA Astrophysics Data System (ADS)

Bays, Samuel Eugene

2008-10-01

In the past several years there has been a renewed interest in sodium fast reactor (SFR) technology for the purpose of destroying transuranic waste (TRU) produced by light water reactors (LWR). The utility of SFRs as waste burners is due to the fact that higher neutron energies allow all of the actinides, including the minor actinides (MA), to contribute to fission. It is well understood that many of the design issues of LWR spent nuclear fuel (SNF) disposal in a geologic repository are linked to MAs. Because the probability of fission for essentially all the "non-fissile" MAs is nearly zero at low neutron energies, these isotopes act as a neutron capture sink in most thermal reactor systems. Furthermore, because most of the isotopes produced by these capture reactions are also non-fissile, they too are neutron sinks in most thermal reactor systems. Conversely, with high neutron energies, the MAs can produce neutrons by fast fission. Additionally, capture reactions transmute the MAs into mostly plutonium isotopes, which can fission more readily at any energy. The transmutation of non-fissile into fissile atoms is the premise of the plutonium breeder reactor. In a breeder reactor, not only does the non-fissile "fertile" U-238 atom contribute fast fission neutrons, but also transmutes into fissile Pu-239. The fissile value of the plutonium produced by MA transmutation can only be realized in fast neutron spectra. This is due to the fact that the predominate isotope produced by MA transmutation, Pu-238, is itself not fissile. However, the Pu-238 fission cross section is significantly larger than the original transmutation parent, predominately: Np-237 and Am-241, in the fast energy range. Also, Pu-238's fission cross section and fission-to-capture ratio is almost as high as that of fissile Pu-239 in the fast neutron spectrum. It is also important to note that a neutron absorption in Pu-238, that does not cause fission, will instead produce fissile Pu-239. Given this fast fissile quality and also the fact that Pu-238 is transmuted from Np-237 and Am-241, these MAs are regarded as fertile material in the SFR design proposed by this dissertation. This dissertation demonstrates a SFR design which is dedicated to plutonium breeding by targeting Am-241 transmutation. This SFR design uses a moderated axial transmutation target that functions primarily as a pseudo-blanket fuel, which is reprocessed with the active driver fuel in an integrated recycling strategy. This work demonstrates the cost and feasibility advantages of plutonium breeding via MA transmutation by adopting reactor, reprocessing and fuel technologies previously demonstrated for traditional breeder reactors. The fuel cycle proposed seeks to find a harmony between the waste management advantages of transuranic burning SFRs and the resource sustainability of traditional plutonium breeder SFRs. As a result, the enhanced plutonium conversion from MAs decreases the burner SFR's fuel costs, by extracting more fissile value from the initial TRU purchased through SNF reprocessing.

Actinide management with commercial fast reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ohki, Shigeo

The capability of plutonium-breeding and minor-actinide (MA) transmutation in the Japanese commercial sodium-cooled fast reactor offers one of practical solutions for obtaining sustainable energy resources as well as reducing radioactive toxicity and inventory. The reference core design meets the requirement of flexible breeding ratio from 1.03 to 1.2. The MA transmutation amount has been evaluated as 50-100 kg/GW{sub e}y if the MA content in fresh fuel is 3-5 wt%, where about 30-40% of initial MA can be transmuted in the discharged fuel.

Comparative analysis of thorium and uranium fuel for transuranic recycle in a sodium cooled Fast Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

C. Fiorina; N. E. Stauff; F. Franceschini

2013-12-01

The present paper compares the reactor physics and transmutation performance of sodium-cooled Fast Reactors (FRs) for TRansUranic (TRU) burning with thorium (Th) or uranium (U) as fertile materials. The 1000 MWt Toshiba-Westinghouse Advanced Recycling Reactor (ARR) conceptual core has been used as benchmark for the comparison. Both burner and breakeven configurations sustained or started with a TRU supply, and assuming full actinide homogeneous recycle strategy, have been developed. State-of-the-art core physics tools have been employed to establish fuel inventory and reactor physics performances for equilibrium and transition cycles. Results show that Th fosters large improvements in the reactivity coefficients associatedmore » with coolant expansion and voiding, which enhances safety margins and, for a burner design, can be traded for maximizing the TRU burning rate. A trade-off of Th compared to U is the significantly larger fuel inventory required to achieve a breakeven design, which entails additional blankets at the detriment of core compactness as well as fuel manufacturing and separation requirements. The gamma field generated by the progeny of U-232 in the U bred from Th challenges fuel handling and manufacturing, but in case of full recycle, the high contents of Am and Cm in the transmutation fuel impose remote fuel operations regardless of the presence of U-232.« less

Georgia Tech Studies of Sub-Critical Advanced Burner Reactors with a D-T Fusion Tokamak Neutron Source for the Transmutation of Spent Nuclear Fuel

NASA Astrophysics Data System (ADS)

Stacey, W. M.

2009-09-01

The possibility that a tokamak D-T fusion neutron source, based on ITER physics and technology, could be used to drive sub-critical, fast-spectrum nuclear reactors fueled with the transuranics (TRU) in spent nuclear fuel discharged from conventional nuclear reactors has been investigated at Georgia Tech in a series of studies which are summarized in this paper. It is found that sub-critical operation of such fast transmutation reactors is advantageous in allowing longer fuel residence time, hence greater TRU burnup between fuel reprocessing stages, and in allowing higher TRU loading without compromising safety, relative to what could be achieved in a similar critical transmutation reactor. The required plasma and fusion technology operating parameter range of the fusion neutron source is generally within the anticipated operational range of ITER. The implications of these results for fusion development policy, if they hold up under more extensive and detailed analysis, is that a D-T fusion tokamak neutron source for a sub-critical transmutation reactor, built on the basis of the ITER operating experience, could possibly be a logical next step after ITER on the path to fusion electrical power reactors. At the same time, such an application would allow fusion to contribute to meeting the nation's energy needs at an earlier stage by helping to close the fission reactor nuclear fuel cycle.

Georgia Institute of Technology research on the Gas Core Actinide Transmutation Reactor (GCATR)

NASA Technical Reports Server (NTRS)

Clement, J. D.; Rust, J. H.; Schneider, A.; Hohl, F.

1976-01-01

The program reviewed is a study of the feasibility, design, and optimization of the GCATR. The program is designed to take advantage of initial results and to continue work carried out on the Gas Core Breeder Reactor. The program complements NASA's program of developing UF6 fueled cavity reactors for power, nuclear pumped lasers, and other advanced technology applications. The program comprises: (1) General Studies--Parametric survey calculations performed to examine the effects of reactor spectrum and flux level on the actinide transmutation for GCATR conditions. The sensitivity of the results to neutron cross sections are to be assessed. Specifically, the parametric calculations of the actinide transmutation are to include the mass, isotope composition, fission and capture rates, reactivity effects, and neutron activity of recycled actinides. (2) GCATR Design Studies--This task is a major thrust of the proposed research program. Several subtasks are considered: optimization criteria studies of the blanket and fuel reprocessing, the actinide insertion and recirculation system, and the system integration. A brief review of the background of the GCATR and ongoing research is presented.

Analysis and optimization of minor actinides transmutation blankets with regards to neutron and gamma sources

NASA Astrophysics Data System (ADS)

Kooymana, Timothée; Buiron, Laurent; Rimpault, Gérald

2017-09-01

Heterogeneous loading of minor actinides in radial blankets is a potential solution to implement minor actinides transmutation in fast reactors. However, to compensate for the lower flux level experienced by the blankets, the fraction of minor actinides to be loaded in the blankets must be increased to maintain acceptable performances. This severely increases the decay heat and neutron source of the blanket assemblies, both before and after irradiation, by more than an order of magnitude in the case of neutron source for instance. We propose here to implement an optimization methodology of the blankets design with regards to various parameters such as the local spectrum or the mass to be loaded, with the objective of minimizing the final neutron source of the spent assembly while maximizing the transmutation performances of the blankets. In a first stage, an analysis of the various contributors to long and short term neutron and gamma source is carried out while in a second stage, relevant estimators are designed for use in the effective optimization process, which is done in the last step. A comparison with core calculations is finally done for completeness and validation purposes. It is found that the use of a moderated spectrum in the blankets can be beneficial in terms of final neutron and gamma source without impacting minor actinides transmutation performances compared to more energetic spectrum that could be achieved using metallic fuel for instance. It is also confirmed that, if possible, the use of hydrides as moderating material in the blankets is a promising option to limit the total minor actinides inventory in the fuel cycle. If not, it appears that focus should be put upon an increased residence time for the blankets rather than an increase in the acceptable neutron source for handling and reprocessing.

Nuclear Methods for Transmutation of Nuclear Waste: Problems, Perspextives, Cooperative Research - Proceedings of the International Workshop

NASA Astrophysics Data System (ADS)

Khankhasayev, Zhanat B.; Kurmanov, Hans; Plendl, Mikhail Kh.

1996-12-01

The Table of Contents for the full book PDF is as follows: * Preface * I. Review of Current Status of Nuclear Transmutation Projects * Accelerator-Driven Systems — Survey of the Research Programs in the World * The Los Alamos Accelerator-Driven Transmutation of Nuclear Waste Concept * Nuclear Waste Transmutation Program in the Czech Republic * Tentative Results of the ISTC Supported Study of the ADTT Plutonium Disposition * Recent Neutron Physics Investigations for the Back End of the Nuclear Fuel Cycle * Optimisation of Accelerator Systems for Transmutation of Nuclear Waste * Proton Linac of the Moscow Meson Factory for the ADTT Experiments * II. Computer Modeling of Nuclear Waste Transmutation Methods and Systems * Transmutation of Minor Actinides in Different Nuclear Facilities * Monte Carlo Modeling of Electro-nuclear Processes with Nonlinear Effects * Simulation of Hybrid Systems with a GEANT Based Program * Computer Study of 90Sr and 137Cs Transmutation by Proton Beam * Methods and Computer Codes for Burn-Up and Fast Transients Calculations in Subcritical Systems with External Sources * New Model of Calculation of Fission Product Yields for the ADTT Problem * Monte Carlo Simulation of Accelerator-Reactor Systems * III. Data Basis for Transmutation of Actinides and Fission Products * Nuclear Data in the Accelerator Driven Transmutation Problem * Nuclear Data to Study Radiation Damage, Activation, and Transmutation of Materials Irradiated by Particles of Intermediate and High Energies * Radium Institute Investigations on the Intermediate Energy Nuclear Data on Hybrid Nuclear Technologies * Nuclear Data Requirements in Intermediate Energy Range for Improvement of Calculations of ADTT Target Processes * IV. Experimental Studies and Projects * ADTT Experiments at the Los Alamos Neutron Science Center * Neutron Multiplicity Distributions for GeV Proton Induced Spallation Reactions on Thin and Thick Targets of Pb and U * Solid State Nuclear Track Detector and Radiochemical Studies on the Transmutation of Nuclei Using Relativistic Heavy Ions * Experimental and Theoretical Study of Radionuclide Production on the Electronuclear Plant Target and Construction Materials Irradiated by 1.5 GeV and 130 MeV Protons * Neutronics and Power Deposition Parameters of the Targets Proposed in the ISTC Project 17 * Multicycle Irradiation of Plutonium in Solid Fuel Heavy-Water Blanket of ADS * Compound Neutron Valve of Accelerator-Driven System Sectioned Blanket * Subcritical Channel-Type Reactor for Weapon Plutonium Utilization * Accelerator Driven Molten-Fluoride Reactor with Modular Heat Exchangers on PB-BI Eutectic * A New Conception of High Power Ion Linac for ADTT * Pions and Accelerator-Driven Transmutation of Nuclear Waste? * V. Problems and Perspectives * Accelerator-Driven Transmutation Technologies for Resolution of Long-Term Nuclear Waste Concerns * Closing the Nuclear Fuel-Cycle and Moving Toward a Sustainable Energy Development * Workshop Summary * List of Participants

Application of gaseous core reactors for transmutation of nuclear waste

NASA Technical Reports Server (NTRS)

Schnitzler, B. G.; Paternoster, R. R.; Schneider, R. T.

1976-01-01

An acceptable management scheme for high-level radioactive waste is vital to the nuclear industry. The hazard potential of the trans-uranic actinides and of key fission products is high due to their nuclear activity and/or chemical toxicity. Of particular concern are the very long-lived nuclides whose hazard potential remains high for hundreds of thousands of years. Neutron induced transmutation offers a promising technique for the treatment of problem wastes. Transmutation is unique as a waste management scheme in that it offers the potential for "destruction" of the hazardous nuclides by conversion to non-hazardous or more manageable nuclides. The transmutation potential of a thermal spectrum uranium hexafluoride fueled cavity reactor was examined. Initial studies focused on a heavy water moderated cavity reactor fueled with 5% enriched U-235-F6 and operating with an average thermal flux of 6 times 10 to the 14th power neutrons/sq cm-sec. The isotopes considered for transmutation were I-129, Am-241, Am-242m, Am-243, Cm-243, Cm-244, Cm-245, and Cm-246.

Development of Metallic Fuels for Actinide Transmutation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hayes, Steven Lowe; Fielding, Randall Sidney; Benson, Michael Timothy

Research and development activities on metallic fuels are focused on their potential use for actinide transmutation in future sodium fast reactors. As part of this application, there is also a need for a near zero-loss fabrication process and a desire to demonstrate a multifold increase in burnup potential. The incorporation of Am and Np into the traditional U-20Pu-10Zr metallic fuel alloy was demonstrated in the US during the Integral Fast Reactor Program of the 1980’s and early 1990’s. However, the conventional counter gravity injection casting method performed under vacuum, previously used to fabricate these metallic fuel alloys, was not optimizedmore » for mitigating loss of the volatile Am constituent in the casting charge; as a result, approximately 40% of the Am casting charge failed to be incorporated into the as-cast fuel alloys. Fabrication development efforts of the past few years have pursued an optimized bottom-pour casting method to increase utilization of the melted charge to near 100%, and a differential pressure casting approach, performed under an argon overpressure, has been demonstrated to result in essentially no loss of Am due to volatilization during fabrication. In short, a path toward zero-loss fabrication of metallic fuels including minor actinides has been shown to be feasible. Irradiation testing of advanced metallic fuel alloys in the Advanced Test Reactor (ATR) has been underway since 2003. Testing in the ATR is performed inside of cadmium-shrouded positions to remove >99% of the thermal flux incident on the test fuels, resulting in an epi-thermal driven fuel test that is free from gross flux depression and producing an essentially prototypic radial temperature profile inside the fuel rodlets. To date, three irradiation test series (AFC-1,2,3) have been completed. Over 20 different metallic fuel alloys have been tested to burnups as high as 30% with constituent compositions of Pu up to 30%, Am up to 12%, Np up to 10%, and Zr between 10 and 60%. In general, the performance of all of these substantially disparate metallic fuel alloys has been observed to be excellent, and their irradiation behaviors are generally consistent with historic norms for metallic fuels without minor actinide additions and having lower Pu or Zr contents. Future work is being undertaken with a view toward increasing the burnup potential of metallic fuels even more. Design innovations under investigation include: 1) lowering the fuel smear density in order to accommodate more swelling, 2) annular fuel geometry to eliminate the need for a sodium bond, 3) minor alloy additions to stabilize lanthanide fission products inside the fuel and prevent their transport to the cladding where they can participate in fuel-cladding chemical interaction (FCCI), and 4) coatings/liners on the cladding inner surface to mitigate FCCI and enable higher temperature operation. This paper will present the current state of development of metallic fuels for actinide transmutation in the US. Highlights will include recent results from metallic fuel casting experiments, experiments to identify alloy additions to immobilize lanthanide fission products, and postirradiation examinations of annular metallic fuels at low burnup.« less

Roadmap to an Engineering-Scale Nuclear Fuel Performance & Safety Code

DOE Office of Scientific and Technical Information (OSTI.GOV)

Turner, John A; Clarno, Kevin T; Hansen, Glen A

2009-09-01

Developing new fuels and qualifying them for large-scale deployment in power reactors is a lengthy and expensive process, typically spanning a period of two decades from concept to licensing. Nuclear fuel designers serve an indispensable role in the process, at the initial exploratory phase as well as in analysis of the testing results. In recent years fuel performance capabilities based on first principles have been playing more of a role in what has traditionally been an empirically dominated process. Nonetheless, nuclear fuel behavior is based on the interaction of multiple complex phenomena, and recent evolutionary approaches are being applied moremore » on a phenomenon-by-phenomenon basis, targeting localized problems, as opposed to a systematic approach based on a fundamental understanding of all interacting parameters. Advanced nuclear fuels are generally more complex, and less understood, than the traditional fuels used in existing reactors (ceramic UO{sub 2} with burnable poisons and other minor additives). The added challenges are primarily caused by a less complete empirical database and, in the case of recycled fuel, the inherent variability in fuel compositions. It is clear that using the traditional approach to develop and qualify fuels over the entire range of variables pertinent to the U.S. Department of Energy (DOE) Office of Nuclear Energy on a timely basis with available funds would be very challenging, if not impossible. As a result the DOE Office of Nuclear Energy has launched the Nuclear Energy Advanced Modeling and Simulation (NEAMS) approach to revolutionize fuel development. This new approach is predicated upon transferring the recent advances in computational sciences and computer technologies into the fuel development program. The effort will couple computational science with recent advances in the fundamental understanding of physical phenomena through ab initio modeling and targeted phenomenological testing to leapfrog many fuel-development activities. Realizing the full benefits of this approach will likely take some time. However, it is important that the developmental activities for modeling and simulation be tightly coupled with the experimental activities to maximize feedback effects and accelerate both the experimental and analytical elements of the program toward a common objective. The close integration of modeling and simulation and experimental activities is key to developing a useful fuel performance simulation capability, providing a validated design and analysis tool, and understanding the uncertainties within the models and design process. The efforts of this project are integrally connected to the Transmutation Fuels Campaign (TFC), which maintains as a primary objective to formulate, fabricate, and qualify a transuranic-based fuel with added minor actinides for use in future fast reactors. Additional details of the TFC scope can be found in the Transmutation Fuels Campaign Execution Plan. This project is an integral component of the TFC modeling and simulation effort, and this multiyear plan borrowed liberally from the Transmutation Fuels Campaign Modeling and Simulation Roadmap. This document provides the multiyear staged development plan to develop a continuum-level Integrated Performance and Safety Code (IPSC) to predict the behavior of the fuel and cladding during normal reactor operations and anticipated transients up to the point of clad breach.« less

Promises and Challenges of Thorium Implementation for Transuranic Transmutation - 13550

DOE Office of Scientific and Technical Information (OSTI.GOV)

Franceschini, F.; Lahoda, E.; Wenner, M.

2013-07-01

This paper focuses on the challenges of implementing a thorium fuel cycle for recycle and transmutation of long-lived actinide components from used nuclear fuel. A multi-stage reactor system is proposed; the first stage consists of current UO{sub 2} once-through LWRs supplying transuranic isotopes that are continuously recycled and burned in second stage reactors in either a uranium (U) or thorium (Th) carrier. The second stage reactors considered for the analysis are Reduced Moderation Pressurized Water Reactors (RMPWRs), reconfigured from current PWR core designs, and Fast Reactors (FRs) with a burner core design. While both RMPWRs and FRs can in principlemore » be employed, each reactor and associated technology has pros and cons. FRs have unmatched flexibility and transmutation efficiency. RMPWRs have higher fuel manufacturing and reprocessing requirements, but may represent a cheaper solution and the opportunity for a shorter time to licensing and deployment. All options require substantial developments in manufacturing, due to the high radiation field, and reprocessing, due to the very high actinide recovery ratio to elicit the claimed radiotoxicity reduction. Th reduces the number of transmutation reactors, and is required to enable a viable RMPWR design, but presents additional challenges on manufacturing and reprocessing. The tradeoff between the various options does not make the choice obvious. Moreover, without an overarching supporting policy in place, the costly and challenging technologies required inherently discourage industrialization of any transmutation scheme, regardless of the adoption of U or Th. (authors)« less

Hardening neutron spectrum for advanced actinide transmutation experiments in the ATR.

PubMed

Chang, G S; Ambrosek, R G

2005-01-01

The most effective method for transmuting long-lived isotopes contained in spent nuclear fuel into shorter-lived fission products is in a fast neutron spectrum reactor. In the absence of a fast test reactor in the United States, initial irradiation testing of candidate fuels can be performed in a thermal test reactor that has been modified to produce a test region with a hardened neutron spectrum. Such a test facility, with a spectrum similar but somewhat softer than that of the liquid-metal fast breeder reactor (LMFBR), has been constructed in the INEEL's Advanced Test Reactor (ATR). The radial fission power distribution of the actinide fuel pin, which is an important parameter in fission gas release modelling, needs to be accurately predicted and the hardened neutron spectrum in the ATR and the LMFBR fast neutron spectrum is compared. The comparison analyses in this study are performed using MCWO, a well-developed tool that couples the Monte Carlo transport code MCNP with the isotope depletion and build-up code ORIGEN-2. MCWO analysis yields time-dependent and neutron-spectrum-dependent minor actinide and Pu concentrations and detailed radial fission power profile calculations for a typical fast reactor (LMFBR) neutron spectrum and the hardened neutron spectrum test region in the ATR. The MCWO-calculated results indicate that the cadmium basket used in the advanced fuel test assembly in the ATR can effectively depress the linear heat generation rate in the experimental fuels and harden the neutron spectrum in the test region.

Next generation fuel irradiation capability in the High Flux Reactor Petten

NASA Astrophysics Data System (ADS)

Fütterer, Michael A.; D'Agata, Elio; Laurie, Mathias; Marmier, Alain; Scaffidi-Argentina, Francesco; Raison, Philippe; Bakker, Klaas; de Groot, Sander; Klaassen, Frodo

2009-07-01

This paper describes selected equipment and expertise on fuel irradiation testing at the High Flux Reactor (HFR) in Petten, The Netherlands. The reactor went critical in 1961 and holds an operating license up to at least 2015. While HFR has initially focused on Light Water Reactor fuel and materials, it also played a decisive role since the 1970s in the German High Temperature Reactor (HTR) development program. A variety of tests related to fast reactor development in Europe were carried out for next generation fuel and materials, in particular for Very High Temperature Reactor (V/HTR) fuel, fuel for closed fuel cycles (U-Pu and Th-U fuel cycle) and transmutation, as well as for other innovative fuel types. The HFR constitutes a significant European infrastructure tool for the development of next generation reactors. Experimental facilities addressed include V/HTR fuel tests, a coated particle irradiation rig, and tests on fast reactor, transmutation and thorium fuel. The rationales for these tests are given, results are provided and further work is outlined.

Vaporisation of candidate nuclear fuels and targets for transmutation of minor actinides

NASA Astrophysics Data System (ADS)

Gotcu-Freis, P.; Hiernaut, J.-P.; Colle, J.-Y.; Nästrén, C.; Carretero, A. Fernandez; Konings, R. J. M.

2011-04-01

The thermal stability and high temperature behaviour of candidate fuels and targets for transmutation of minor actinides has been investigated. Zirconia-based solid solution, MgO-based CERCER and molybdenum-based CERMET fuels containing Am and/or Pu in various concentrations were heated up to 2700 K in a Knudsen cell coupled with a quadrupole mass spectrometer, to measure their vapour pressure and vapour composition. The results reveal that the vaporisation of the actinides from the samples is not only determined by the thermodynamics of the system but is also related to the dynamic evolution of multi-component mixtures with complex composition or microstructure.

Impacts of Heterogeneous Recycle in Fast Reactors on Overall Fuel Cycle

DOE Office of Scientific and Technical Information (OSTI.GOV)

Temitope A. Taiwo; Samuel E. Bays; Abdullatif M. Yacout

2011-03-01

A study in the United States has evaluated the attributes of the heterogeneous recycle approach for plutonium and minor actinide transmutation in fast reactor fuel cycles, with comparison to the homogeneous recycle approach, where pertinent. The work investigated the characteristics, advantages, and disadvantages of the approach in the overall fuel cycle, including reactor transmutation, systems and safety impacts, fuel separation and fabrication issues, and proliferation risk and transportation impacts. For this evaluation, data from previous and ongoing national studies on heterogeneous recycle were reviewed and synthesized. Where useful, information from international sources was included in the findings. The intent ofmore » the work was to provide a comprehensive assessment of the heterogeneous recycle approach at the current time.« less

Investigation of the feasibility of a small scale transmutation device

NASA Astrophysics Data System (ADS)

Sit, Roger Carson

This dissertation presents the design and feasibility of a small-scale, fusion-based transmutation device incorporating a commercially available neutron generator. It also presents the design features necessary to optimize the device and render it practical for the transmutation of selected long-lived fission products and actinides. Four conceptual designs of a transmutation device were used to study the transformation of seven radionuclides: long-lived fission products (Tc-99 and I-129), short-lived fission products (Cs-137 and Sr-90), and selective actinides (Am-241, Pu-238, and Pu-239). These radionuclides were chosen because they are major components of spent nuclear fuel and also because they exist as legacy sources that are being stored pending a decision regarding their ultimate disposition. The four designs include the use of two different devices; a Deuterium-Deuterium (D-D) neutron generator (for one design) and a Deuterium-Tritium (D-T) neutron generator (for three designs) in configurations which provide different neutron energy spectra for targeting the radionuclide for transmutation. Key parameters analyzed include total fluence and flux requirements; transmutation effectiveness measured as irradiation effective half-life; and activation products generated along with their characteristics: activity, dose rate, decay, and ingestion and inhalation radiotoxicity. From this investigation, conclusions were drawn about the feasibility of the device, the design and technology enhancements that would be required to make transmutation practical, the most beneficial design for each radionuclide, the consequence of the transmutation, and radiation protection issues that are important for the conceptual design of the transmutation device. Key conclusions from this investigation include: (1) the transmutation of long-lived fission products and select actinides can be practical using a small-scale, fusion driven transmutation device; (2) the transmutation of long-lived fission products could result in an irradiation effective half-life of a few years with a three order magnitude increase in the on-target neutron flux accomplishable through a combination of technological enhancements to the source and system design optimization; (3) the transmutation of long-lived fission products requires a thermal-slow energy spectrum to prevent the generation of activation products with half-lives even longer than the original radionuclide; (4) there is no benefit in trying to transmute short-lived fission products due to the ineffectiveness of the transmutation process and the generation of a multiplicity of counterproductive activation products; (5) for actinides, irradiation effective half-lives of < 1 year can be achieved with a four orders magnitude increase in the on-target flux; (6) the ideal neutron energy spectra for transmuting actinides is highly dependent on the particular radionuclide and its fission-to-capture ratio as they determine the generationrate of other actinides; and (7) the methodology developed in this dissertation provides a mechanism that can be used for studying the feasibility of transmuting other radionuclides, and its application can be extended to studying the production of radionuclides of interest in a transmutation process. Although large-scale transmutation technology is presently being researched world-wide for spent fuel management applications, such technology will not be viable for a couple of decades. This dissertation investigated the concept of a small-scale transmutation device using present technology. The results of this research show that with reasonable enhancements, transmutation of specific radionuclides can be practical in the near term.

Importance of the (n,gamma) Cm-247 Evaluation on Neutron Emission in Fast Reactor Fuel Cycle Analysis

DOE Office of Scientific and Technical Information (OSTI.GOV)

Benoit Forget; Mehdi Asgari; Rodolfo M. Ferrer

2007-11-01

As part of the GNEP program, it is envisioned to build a fast reactor for the transmutation of minor actinides. The spent nuclear fuel from the current fleet of light water reactors would be recycled, the current baseline is the UREX+1a process, and would act as a feed for the fast reactor. As the fuel is irradiated in a fast reactor a certain quantity of minor actinides would thus build up in the fuel stream creating possible concerns with the neutron emission of these minor actinides for fuel transportation, handling and fabrication. Past neutronic analyses had not tracked minor actinidesmore » above Cm-246 in the transmutation chain, because of the small influence on the overall reactor performance and cycle parameters. However, when trying to quantify the neutron emission from the recycled fuel with high minor actinide content, these higher isotopes play an essential role and should be included in the analysis. In this paper, the influence of tracking these minor actinides on the calculated neutron emission is presented. Also presented is the particular influence of choosing a different evaluated cross section data set to represent the minor actinides above Cm-246. The first representation uses the cross-sections provided by MC2-2 for all isotopes, while the second representation uses infinitely diluted ENDF/BVII.0 cross-sections for Cm-247 to Cf-252 and MC2-2 for all other isotopes.« less

Status of the Neutron Capture Measurement on 237Np with the DANCE Array at LANSCE

DOE Office of Scientific and Technical Information (OSTI.GOV)

Esch, E.-I.; Bond, E.M.; Bredeweg, T. A.

2005-05-24

Neptunium-237 is a major constituent of spent nuclear fuel. Estimates place the amount of 237Np bound for the Yucca Mountain high-level waste repository at 40 metric tons. The Department of Energy's Advanced Fuel Cycle Initiative program is evaluating methods for transmuting the actinide waste that will be generated by future operation of commercial nuclear power plants. The critical parameter that defines the transmutation efficiency of actinide isotopes is the neutron fission-to-capture ratio for the particular isotope in a given neutron spectrum. The calculation of transmutation efficiency therefore requires accurate fission and capture cross sections. Current 237Np evaluations available for transmutermore » system studies show significant discrepancies in both the fission and capture cross sections in the energy regions of interest. Herein we report on 237Np (n,{gamma}) measurements using the recently commissioned DANCE array.« less

The benefits of an advanced fast reactor fuel cycle for plutonium management

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hannum, W.H.; McFarlane, H.F.; Wade, D.C.

1996-12-31

The United States has no program to investigate advanced nuclear fuel cycles for the large-scale consumption of plutonium from military and civilian sources. The official U.S. position has been to focus on means to bury spent nuclear fuel from civilian reactors and to achieve the spent fuel standard for excess separated plutonium, which is considered by policy makers to be an urgent international priority. Recently, the National Research Council published a long awaited report on its study of potential separation and transmutation technologies (STATS), which concluded that in the nuclear energy phase-out scenario that they evaluated, transmutation of plutonium andmore » long-lived radioisotopes would not be worth the cost. However, at the American Nuclear Society Annual Meeting in June, 1996, the STATS panelists endorsed further study of partitioning to achieve superior waste forms for burial, and suggested that any further consideration of transmutation should be in the context of energy production, not of waste management. 2048 The U.S. Department of Energy (DOE) has an active program for the short-term disposition of excess fissile material and a `focus area` for safe, secure stabilization, storage and disposition of plutonium, but has no current programs for fast reactor development. Nevertheless, sufficient data exist to identify the potential advantages of an advanced fast reactor metallic fuel cycle for the long-term management of plutonium. Advantages are discussed.« less

Impact of the deployment schedule of fast breeding reactors in the frame of French act for nuclear materials and radioactive waste management

DOE Office of Scientific and Technical Information (OSTI.GOV)

Le Mer, J.; Garzenne, C.; Lemasson, D.

In the frame of the French Act of June 28, 2006 on 'a sustainable management of nuclear materials and radioactive waste' EDF R and D assesses various research scenarios of transition between the actual French fleet and a Generation IV fleet with a closed fuel cycle where plutonium is multi-recycled. The basic scenarios simulate a deployment of 60 GWe of Sodium-cooled Fast Reactors (SFRs) in two steps: one third from 2040 to 2050 and the rest from 2080 to 2100 (scenarios 2040). These research scenarios assume that SFR technology will be ready for industrial deployment in 2040. One of themore » many sensitivity analyses that EDF, as a nuclear power plant operator, must evaluate is the impact of a delay of SFR technology in terms of uranium consumptions, plutonium needs and fuel cycle utilities gauging. The sensitivity scenarios use the same assumptions as scenarios 2040 but they simulate a different transition phase: SFRs are deployed in one step between 2080 and 2110 (scenarios 2080). As the French Act states to conduct research on minor actinides (MA) management, we studied different options for 2040 and 2080 scenarios: no MA transmutation, americium transmutation in heterogeneous mode based on americium Bearing Blankets (AmBB) in SFRs and all MA transmutation in heterogeneous mode based on MA Bearing Blankets (MABB). Moreover, we studied multiple parameters that could impact the deployment of these reactors (SFR load factor, increase of the use of MOX in Light Water Reactors, increase of the cooling time in spent nuclear fuel storage...). Each scenario has been computed with the EDF R and D fuel cycle simulation code TIRELIRE-STRATEGIE and optimized to meet various fuel cycle constraints such as using the reprocessing facility with long period of constant capacity, keeping the temporary stored mass of plutonium and MA under imposed limits, recycling older assemblies first... These research scenarios show that the transition from the current PWR fleet to an equivalent fleet of Generation IV SFR can follow different courses. The design of SFR-V2B that we used in our studies needs a high inventory of plutonium resulting in tension on this resource. Several options can be used in order to loosen this tension: our results lead to favour the use of axial breeding blanket in SFR. Load factor of upcoming reactors has to be regarded with attention as it has a high impact on plutonium resource for a given production of electricity. The deployment of SFRs beginning in 2080 instead of 2040 following the scenarios we described creates higher tensions on reprocessing capacity, separated plutonium storage and spent fuel storage. In the frame of the French Act, we studied minor actinides transmutation. The flux of MA in all fuel cycle plants is really high, which will lead to decay heat, a and neutron emission related problems. In terms of reduction of MA inventories, the deployment of MA transmutation cycle must not delay the installation of SFRs. The plutonium production in MABB and AmBB does not allow reducing the use of axial breeding blankets. The impact of MA or Am transmutation over the high level waste disposal is more important if the SFRs are deployed later. Transmutation option (americium or all MA) does not have a significant impact on the number of canister produced nor on its long-term thermal properties. (authors)« less

AN EVALUATION OF POTENTIAL LINER MATERIALS FOR ELIMINATING FCCI IN IRRADIATED METALLIC NUCLEAR FUEL ELEMENTS

DOE Office of Scientific and Technical Information (OSTI.GOV)

D. D. Keiser; J. I. Cole

2007-09-01

Metallic nuclear fuels are being looked at as part of the Global Nuclear Energy Program for transmuting longlive transuranic actinide isotopes contained in spent nuclear fuel into shorter-lived fission products. In order to optimize the performance of these fuels, the concept of using liners to eliminate the fuel/cladding chemical interactions that can occur during irradiation of a fuel element has been investigated. The potential liner materials Zr and V have been tested using solid-solid diffusion couples, consisting of liner materials butted against fuel alloys and against cladding materials. The couples were annealed at the relatively high temperature of 700°C. Thismore » temperature would be the absolute maximum temperature present at the fuel/cladding interface for a fuel element in-reactor. Analysis was performed using a scanning electron microscope equipped with energy-dispersive and wavelengthdispersive spectrometers (SEM/EDS/WDS) to evaluate any developed diffusion structures. At 700°C, minimal interaction was observed between the metallic fuels and either Zr or V. Similarly, limited interaction was observed between the Zr and V and the cladding materials. The best performing liner material appeared to be the V, based on amounts of interaction.« less

Potential benefits of waste transmutation to the U.S. high-level waste respository

DOE Office of Scientific and Technical Information (OSTI.GOV)

Michaels, G.E.

1995-10-01

This paper reexamines the potential benefits of waste transmutation to the proposed U.S. geologic repository at the Yucca Mountain site based on recent progress in the performance assessment for the Yucca Mountain base case of spent fuel emplacement. It is observed that actinides are assumed to have higher solubility than in previous studies and that Np and other actinides now dominate the projected aqueous releases from a Yucca Mountain repository. Actinides are also indentified as the dominant source of decay heat in the repository, and the effect of decay heat in perturbing the hydrology, geochemistry, and thermal characteristics of Yuccamore » Mountain are reviewed. It is concluded that the potential for thermally-driven, buoyant, gas-phase flow at Yucca Mountain introduces data and modeling requirements that will increase the costs of licensing the site and may cause the site to be unattractive for geologic disposal of wastes. A transmutation-enabled cold repository is proposed that might allow licensing of a repository to be based upon currently observable characteristics of the Yucca Mountain site.« less

Thorium Fuel Options for Sustained Transuranic Burning in Pressurized Water Reactors - 12381

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rahman, Fariz Abdul; Lee, John C.; Franceschini, Fausto

2012-07-01

As described in companion papers, Westinghouse is proposing the adoption of a thorium-based fuel cycle to burn the transuranics (TRU) contained in the current Used Nuclear Fuel (UNF) and transition towards a less radio-toxic high level waste. A combination of both light water reactors (LWR) and fast reactors (FR) is envisaged for the task, with the emphasis initially posed on their TRU burning capability and eventually to their self-sufficiency. Given the many technical challenges and development times related to the deployment of TRU burners fast reactors, an interim solution making best use of the current resources to initiate burning themore » legacy TRU inventory while developing and testing some technologies of later use is desirable. In this perspective, a portion of the LWR fleet can be used to start burning the legacy TRUs using Th-based fuels compatible with the current plants and operational features. This analysis focuses on a typical 4-loop PWR, with 17x17 fuel assembly design and TRUs (or Pu) admixed with Th (similar to U-MOX fuel, but with Th instead of U). Global calculations of the core were represented with unit assembly simulations using the Linear Reactivity Model (LRM). Several assembly configurations have been developed to offer two options that can be attractive during the TRU transmutation campaign: maximization of the TRU transmutation rate and capability for TRU multi-recycling, to extend the option of TRU recycling in LWR until the FR is available. Homogeneous as well as heterogeneous assembly configurations have been developed with various recycling schemes (Pu recycle, TRU recycle, TRU and in-bred U recycle etc.). Oxide as well as nitride fuels have been examined. This enabled an assessment of the potential for burning and multi-recycling TRU in a Th-based fuel PWR to compare against other more typical alternatives (U-MOX and variations thereof). Results will be shown indicating that Th-based PWR fuel is a promising option to multi-recycle and burn TRU in a thermal spectrum, while satisfying top-level operational and safety constraints. Various assembly designs have been proposed to assess the TRU burning potential of Th-based fuel in PWRs. In addition to typical homogeneous loading patterns, heterogeneous configurations exploiting the breeding potential of thorium to enable multiple cycles of TRU irradiation and burning have been devised. The homogeneous assembly design, with all pins featuring TRU in Th, has the benefit of a simple loading pattern and the highest rate of TRU transmutation, but it can be used only for a few cycles due to the rapid rise in the TRU content of the recycled fuel, which challenges reactivity control, safety coefficients and fuel handling. Due to its simple loading pattern, such assembly design can be used as the first step of Th implementation, achieving up to 3 times larger TRU transmutation rate than conventional U-MOX, assuming same fraction of MOX assemblies in the core. As the next step in thorium implementation, heterogeneous assemblies featuring a mixed array of Th-U and Th-U-TRU pins, where the U is in-bred from Th, have been proposed. These designs have the potential to enable burning an external supply of TRU through multiple cycles of irradiation, recovery (via reprocessing) and recycling of the residual actinides at the end of each irradiation cycle. This is achieved thanks to a larger breeding of U from Th in the heterogeneous assemblies, which reduces the TRU supply and thus mitigates the increase in the TRU core inventory for the multi-recycled fuel. While on an individual cycle basis the amount of TRU burned in the heterogeneous assembly is reduced with respect to the homogeneous design, TRU burning rates higher than single-pass U-MOX fuel can still be achieved, with the additional benefits of a multi-cycle transmutation campaign recycling all TRU isotopes. Nitride fuel, due its higher density and U breeding potential, together with its better thermal properties, ideally suits the objectives and constraints of the heterogeneous assemblies. However, significant technological advancements must be made before nitride fuels can be employed in an LWR: its water resistance needs to be improved and a viable technology to enrich N in N-15 must be devised. Moreover, for the nitride heterogeneous configurations examined in this study, the enhancement in TRU burning performance is achieved not only by replacing oxide with nitride fuel, but also by increasing the fuel rod size. This latter modification, allowed by the high thermal conductivity of nitride fuel, leads however to a very tight lattice, which may challenge reactor coolant pumps and assembly hold-down mechanisms, the former through an increase in core pressure drop and the latter through an increase in assembly lift-off forces. To alleviate these issues, while still achieving the large fuel-to-moderator ratios resulting from using tight lattices, wire wraps could be used in place of grid spacers. For tight lattices, typical grid spacers are hard to manufacture and their replacement with wire wraps is known to allow for a pressure drop reduction by at least 2 times. The studies, while certainly very preliminary, provide a starting point to devise an optimum strategy for TRU transmutation in Th-based PWR fuel. The viability of the scheme proposed depends on the timely phasing in of the associated technologies, with proper lead time and to solve the many challenges. These challenges are certainly substantial, and make the current once-through U-based scheme pursued in the US by far a more practical (and cheaper) option. However, when compared to other transmutation schemes, the proposed one has arguably similar challenges and unknowns with potentially bigger rewards. (authors)« less

Transuranic Waste Burning Potential of Thorium Fuel in a Fast Reactor - 12423

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wenner, Michael; Franceschini, Fausto; Ferroni, Paolo

Westinghouse Electric Company (referred to as 'Westinghouse' in the rest of this paper) is proposing a 'back-to-front' approach to overcome the stalemate on nuclear waste management in the US. In this approach, requirements to further the societal acceptance of nuclear waste are such that the ultimate health hazard resulting from the waste package is 'as low as reasonably achievable'. Societal acceptability of nuclear waste can be enhanced by reducing the long-term radiotoxicity of the waste, which is currently driven primarily by the protracted radiotoxicity of the transuranic (TRU) isotopes. Therefore, a transition to a more benign radioactive waste can bemore » accomplished by a fuel cycle capable of consuming the stockpile of TRU 'legacy' waste contained in the LWR Used Nuclear Fuel (UNF) while generating waste which is significantly less radio-toxic than that produced by the current open U-based fuel cycle (once through and variations thereof). Investigation of a fast reactor (FR) operating on a thorium-based fuel cycle, as opposed to the traditional uranium-based is performed. Due to a combination between its neutronic properties and its low position in the actinide chain, thorium not only burns the legacy TRU waste, but it does so with a minimal production of 'new' TRUs. The effectiveness of a thorium-based fast reactor to burn legacy TRU and its flexibility to incorporate various fuels and recycle schemes according to the evolving needs of the transmutation scenario have been investigated. Specifically, the potential for a high TRU burning rate, high U-233 generation rate if so desired and low concurrent production of TRU have been used as metrics for the examined cycles. Core physics simulations of a fast reactor core running on thorium-based fuels and burning an external TRU feed supply have been carried out over multiple cycles of irradiation, separation and reprocessing. The TRU burning capability as well as the core isotopic content have been characterized. Results will be presented showing the potential for thorium to reach a high TRU transmutation rate over a wide variety of fuel types (oxide, metal, nitride and carbide) and transmutation schemes (recycle or partition of in-bred U-233). In addition, a sustainable scheme has been devised to burn the TRU accumulated in the core inventory once the legacy TRU supply has been exhausted, thereby achieving long-term virtually TRU-free. A comprehensive 'back-to-front' approach to the fuel cycle has recently been proposed by Westinghouse which emphasizes achieving 'acceptable', low-radiotoxicity, high-level waste, with the intent not only to satisfy all technical constraints but also to improve public acceptance of nuclear energy. Following this approach, the thorium fuel cycle, due to its low radiotoxicity and high potential for TRU transmutation has been selected as a promising solution. Additional studies not shown here have shown significant reduction of decay heat. The TRU burning potential of the Th-based fuel cycle has been illustrated with a variety of fuel types, using the Toshiba ARR to perform the analysis, including scenarios with continued LWR operation of either uranium fueled or thorium fueled LWRs. These scenarios will afford overall reduction in actinide radiotoxicity, however when the TRU supply is exhausted, a continued U- 235 LWR operation must be assumed to provide TRU makeup feed. This scenario will never reach the characteristically low TRU content of a closed thorium fuel cycle with its associated potential benefits on waste radiotoxicity, as exemplified by the transition scenario studied. At present, the cases studied indicate ThC as a potential fuel for maximizing TRU burning, while ThN with nitrogen enriched to 95% N-15 shows the highest breeding potential. As a result, a transition scenario with ThN was developed to show that a sustainable, closed Th-cycle can be achieved starting from burning the legacy TRU stock and completing the transmutation of the residual TRU remaining in the core inventory after the legacy TRU external supply has been exhausted. The radiotoxicity of the actinide waste during the various phases has been characterized, showing the beneficial effect of the decreasing content of TRU in the recycled fuel as the transition to a closed Th-based fuel cycle is undertaken. Due to the back-to-front nature of the proposed methodology, detailed designs are not the first step taken when assessing a fuel cycle scenario potential. As a result, design refinement is still required and should be expected in future studies. Moreover, significant safety assessment, including determination of associated reactivity coefficients, fuel and reprocessing feasibility studies and economic assessments will still be needed for a more comprehensive and meaningful comparison against other potential solutions. With the above considerations in mind, the potential advantages of thorium fuelled reactors on HLW management optimization lead us to believe that thorium fuelled reactor systems can play a significant role in the future and deserve further consideration. (authors)« less

Analysis of Advanced Fuel Assemblies and Core Designs for the Current and Next Generations of LWRs

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ragusa, Jean; Vierow, Karen

2011-09-01

The objective of the project is to design and analyze advanced fuel assemblies for use in current and future light water reactors and to assess their ability to reduce the inventory of transuranic elements, while preserving operational safety. The reprocessing of spent nuclear fuel can delay or avoid the need for a second geological repository in the US. Current light water reactor fuel assembly designs under investigation could reduce the plutonium inventory of reprocessed fuel. Nevertheless, these designs are not effective in stabilizing or reducing the inventory of minor actinides. In the course of this project, we developed and analyzedmore » advanced fuel assembly designs with improved thermal transmutation capability regarding transuranic elements and especially minor actinides. These designs will be intended for use in thermal spectrum (e.g., current and future fleet of light water reactors in the US). We investigated various fuel types, namely high burn-up advanced mixed oxides and inert matrix fuels, in various geometrical designs that are compliant with the core internals of current and future light water reactors. Neutronic/thermal hydraulic effects were included. Transmutation efficiency and safety parameters were used to rank and down-select the various designs.« less

Advanced Burner Reactor with Breed-and-Burn Thorium Blankets for Improved Economics and Resource Utilization

DOE Office of Scientific and Technical Information (OSTI.GOV)

Greenspan, Ehud

2015-11-04

This study assesses the feasibility of designing Seed and Blanket (S&B) Sodium-cooled Fast Reactor (SFR) to generate a significant fraction of the core power from radial thorium fueled blankets that operate on the Breed-and-Burn (B&B) mode without exceeding the radiation damage constraint of presently verified cladding materials. The S&B core is designed to maximize the fraction of neutrons that radially leak from the seed (or “driver”) into the subcritical blanket and reduce neutron loss via axial leakage. The blanket in the S&B core makes beneficial use of the leaking neutrons for improved economics and resource utilization. A specific objective ofmore » this study is to maximize the fraction of core power that can be generated by the blanket without violating the thermal hydraulic and material constraints. Since the blanket fuel requires no reprocessing along with remote fuel fabrication, a larger fraction of power from the blanket will result in a smaller fuel recycling capacity and lower fuel cycle cost per unit of electricity generated. A unique synergism is found between a low conversion ratio (CR) seed and a B&B blanket fueled by thorium. Among several benefits, this synergism enables the very low leakage S&B cores to have small positive coolant voiding reactivity coefficient and large enough negative Doppler coefficient even when using inert matrix fuel for the seed. The benefits of this synergism are maximized when using an annular seed surrounded by an inner and outer thorium blankets. Among the high-performance S&B cores designed to benefit from this unique synergism are: (1) the ultra-long cycle core that features a cycle length of ~7 years; (2) the high-transmutation rate core where the seed fuel features a TRU CR of 0.0. Its TRU transmutation rate is comparable to that of the reference Advanced Burner Reactor (ABR) with CR of 0.5 and the thorium blanket can generate close to 60% of the core power; but requires only one sixth of the reprocessing and fabrication capacity per unit of core power. Nevertheless, these high-performance cores were designed to set upper bounds on the S&B core performance by using larger height and pressure drop than those of typical SFR design. A study was subsequently undertaken to quantify the tradeoff between S&B core design variables and the core performance. This study concludes that a viable S&B core can be designed without significant deviation from SFR core design practices. For example, the S&B core with 120cm active height will be comparable in volume, HM mass and specific power with the S-PRISM core and could fit within the S-PRISM reactor vessel. 43% of this core power will be generated by the once-through thorium blanket; the required capacity for reprocessing and remote fuel fabrication per unit of electricity generated will be approximately one fifth of that for a comparable ABR. The sodium void worth of this 120cm tall S&B core is significantly less positive than that of the reference ABR and the Doppler coefficient is only slightly smaller even though the seed uses a fertile-free fuel. The seed in the high transmutation core requires inert matrix fuel (TRU-40Zr) that has been successfully irradiated by the Fuel Cycle Research & Development program. An additional sensitivity analysis was later conducted to remove the bias introduced by the discrepancy between radiation damage constraints -- 200 DPA applied for S&B cores and fast fluence of 4x1023 n(>0.1MeV)/cm2 applied for ABR core design. Although the performance characteristics of the S&B cores are sensitive to the radiation damage constraint applied, the S&B cores offer very significant performance improvements relative to the conventional ABR core design when using identical constraint.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hayes, S. L.; Harp, J. M.; Chichester, H. J. M.

Research and development activities on metallic fuels in the US are focused on their potential use for actinide transmutation in future sodium fast reactors. As part of this application, there is a desire to demonstrate a multifold increase in burnup potential. A number of metallic fuel design innovations are under investigation with a view toward significantly increasing the burnup potential of metallic fuels, since higher discharge burnups equate to lower potential actinide losses during recycle. Promising innovations under investigation include: 1) lowering the fuel smeared density in order to accommodate the additional swelling expected as burnups increase, 2) utilizing anmore » annular fuel geometry for better geometrical stability at low smeared densities, as well as the potential to eliminate the need for a sodium bond, and 3) minor alloy additions to immobilize lanthanide fission products inside the metallic fuel matrix and prevent their transport to the cladding resulting in fuel-cladding chemical interaction. This paper presents results from these efforts to advance metallic fuel technology in support of high burnup and actinide transmutation objectives. Highlights include examples of fabrication of low smeared density annular metallic fuels, experiments to identify alloy additions effective in immobilizing lanthanide fission products, and early postirradiation examinations of annular metallic fuels having low smeared densities and palladium additions for fission product immobilization.« less

Applications in Nuclear Energy Security

NASA Astrophysics Data System (ADS)

Sheffield, Richard

2009-05-01

A key roadblock to development of additional nuclear power capacity is a concern over management of nuclear waste. Nuclear waste is predominantly comprised of used fuel discharged from operating nuclear reactors. The roughly 100 operating US reactors currently produce about 20% of the US electricity and will create about 87,000 tons of such discharged or ``spent'' fuel over the course of their lifetimes. The long-term radioactivity of the spent fuel drives the need for deep geologic storage that remains stable for millions of years. Nearly all issues related to risks to future generations arising from long-term disposal of such spent nuclear fuel is attributable to approximately the 1% made up primarily of minor actinides. If we can reduce or eliminate this 1% of the spent fuel, then within a few hundred years the toxic nature of the spent fuel drops below that of the natural uranium ore that was originally mined for nuclear fuel. The minor actinides can be efficiently eliminated through nuclear transmutation using as a driver fast-neutrons produced by a spallation process initiated with a high-energy proton beam. This presentation will cover the system design considerations and issues of an accelerator driven transmutation system.

DOE Office of Scientific and Technical Information (OSTI.GOV)

J. Carmack; L. Braase; F. Goldner

The mission of the Advanced Fuels Campaign (AFC) is to perform Research, Development, and Demonstration (RD&D) activities for advanced fuel forms (including cladding) to enhance the performance and safety of the nation’s current and future reactors, enhance proliferation resistance of nuclear fuel, effectively utilize nuclear energy resources, and address the longer-term waste management challenges. This includes development of a state of the art Research and Development (R&D) infrastructure to support the use of a “goal oriented science based approach.” AFC uses a “goal oriented, science based approach” aimed at a fundamental understanding of fuel and cladding fabrication methods and performancemore » under irradiation, enabling the pursuit of multiple fuel forms for future fuel cycle options. This approach includes fundamental experiments, theory, and advanced modeling and simulation. One of the most challenging aspects of AFC is the management, integration, and coordination of major R&D activities across multiple organizations. AFC interfaces and collaborates with Fuel Cycle Technologies (FCT) campaigns, universities, industry, various DOE programs and laboratories, federal agencies (e.g., Nuclear Regulatory Commission [NRC]), and international organizations. Key challenges are the development of fuel technologies to enable major increases in fuel performance (safety, reliability, power and burnup) beyond current technologies, and development of characterization methods and predictive fuel performance models to enable more efficient development and licensing of advanced fuels. Challenged with the research and development of fuels for two different reactor technology platforms, AFC targeted transmutation fuel development and focused ceramic fuel development for Advanced LWR Fuels.« less

New developments and prospects on COSI, the simulation software for fuel cycle analysis

DOE Office of Scientific and Technical Information (OSTI.GOV)

Eschbach, R.; Meyer, M.; Coquelet-Pascal, C.

2013-07-01

COSI, software developed by the Nuclear Energy Direction of the CEA, is a code simulating a pool of nuclear power plants with its associated fuel cycle facilities. This code has been designed to study various short, medium and long term options for the introduction of various types of nuclear reactors and for the use of associated nuclear materials. In the frame of the French Act for waste management, scenario studies are carried out with COSI, to compare different options of evolution of the French reactor fleet and options of partitioning and transmutation of plutonium and minor actinides. Those studies aimmore » in particular at evaluating the sustainability of Sodium cooled Fast Reactors (SFR) deployment and the possibility to transmute minor actinides. The COSI6 version is a completely renewed software released in 2006. COSI6 is now coupled with the last version of CESAR (CESAR5.3 based on JEFF3.1.1 nuclear data) allowing the calculations on irradiated fuel with 200 fission products and 100 heavy nuclides. A new release is planned in 2013, including in particular the coupling with a recommended database of reactors. An exercise of validation of COSI6, carried out on the French PWR historic nuclear fleet, has been performed. During this exercise quantities like cumulative natural uranium consumption, or cumulative depleted uranium, or UOX/MOX spent fuel storage, or stocks of reprocessed uranium, or plutonium content in fresh MOX fuel, or the annual production of high level waste, have been computed by COSI6 and compared to industrial data. The results have allowed us to validate the essential phases of the fuel cycle computation, and reinforces the credibility of the results provided by the code.« less

Role of ADS in the back-end of the fuel cycle strategies and associated design activities: The case of Japan

NASA Astrophysics Data System (ADS)

Oigawa, Hiroyuki; Tsujimoto, Kazufumi; Nishihara, Kenji; Sugawara, Takanori; Kurata, Yuji; Takei, Hayanori; Saito, Shigeru; Sasa, Toshinobu; Obayashi, Hironari

2011-08-01

Reduction of burden caused by radioactive waste management is one of the most critical issues for the sustainable utilization of nuclear power. The Partitioning and Transmutation (P&T) technology provides the possibility to reduce the amount of the radiotoxic inventory of the high-level radioactive waste (HLW) dramatically and to extend the repository capacity. The accelerator-driven system (ADS) is regarded as a powerful tool to effectively transmute minor actinides (MAs) in the "double-strata" fuel cycle strategy. The ADS has a potential to flexibly manage MA in the transient phase from light water reactors (LWRs) to fast breeder reactors (FBRs), and can co-exist with FBR symbiotically and complementarily to enhance the reliability and the safety of the commercial FBR cycle. The concept of ADS in JAEA is a lead-bismuth eutectic (LBE) cooled, tank-type subcritical reactor with the power of 800 MWth driven by a 30 MW superconducting LINAC. By such an ADS, 250 kg of MA can be transmuted annually, which corresponds to the amount of MA produced in 10 units of LWR with 1 GWe. The design study was performed mainly for the subcritical reactor and the spallation target with a beam window. In Japan, Atomic Energy Commission (AEC) has implemented the check and review (C&R) on P&T technology from 2008 to 2009. In the C&R, the benefit of P&T technology, the current status of the R&D, and the way forward to promote it were discussed.

Post-irradiation examinations of THERMHET composite fuels for transmutation

NASA Astrophysics Data System (ADS)

Noirot, J.; Desgranges, L.; Chauvin, N.; Georgenthum, V.

2003-07-01

The thermal behaviour of composite targets dedicated to minor actinide transmutation was studied using THERMHET (thermal behaviour of heterogeneous fuel) irradiation in the SILOE reactor. Three inert matrix fuel designs were tested (macro-mass, jingle and microdispersion) all with a MgAl 2O 4 spinel inert matrix and around 40% weight of UO 2 to simulate minor actinide inclusions. The post-irradiation examinations led to a new interpretation of the temperature measurement by thermocouples located in the central hole of the pellets. A major change in the micro-dispersed structure was detected. The examinations enabled us to understand the behaviour of the spinel during the different stages of irradiation. They revealed an amorphisation at low temperature and then a nano re-crystallisation at high temperature of the spinel in the micro-dispersed case. These results, together with those obtained in the MATINA irradiation of an equivalent structure, show the importance of the irradiation temperature on spinel behaviour.

Industrial research for transmutation scenarios

NASA Astrophysics Data System (ADS)

Camarcat, Noel; Garzenne, Claude; Le Mer, Joël; Leroyer, Hadrien; Desroches, Estelle; Delbecq, Jean-Michel

2011-04-01

This article presents the results of research scenarios for americium transmutation in a 22nd century French nuclear fleet, using sodium fast breeder reactors. We benchmark the americium transmutation benefits and drawbacks with a reference case consisting of a hypothetical 60 GWe fleet of pure plutonium breeders. The fluxes in the various parts of the cycle (reactors, fabrication plants, reprocessing plants and underground disposals) are calculated using EDF's suite of codes, comparable in capabilities to those of other research facilities. We study underground thermal heat load reduction due to americium partitioning and repository area minimization. We endeavor to estimate the increased technical complexity of surface facilities to handle the americium fluxes in special fuel fabrication plants, americium fast burners, special reprocessing shops, handling equipments and transport casks between those facilities.

ADS Model in the TIRELIRE-STRATEGIE Fuel Cycle Simulation Code Application to Minor Actinides Transmutation Studies

DOE Office of Scientific and Technical Information (OSTI.GOV)

Garzenne, Claude; Massara, Simone; Tetart, Philippe

2006-07-01

Accelerator Driven Systems offer the advantage, thanks to the core sub-criticality, to burn highly radioactive elements such as americium and curium in a dedicated stratum, and then to avoid polluting with these elements the main part of the nuclear fleet, which is optimized for electricity production. This paper presents firstly the ADS model implemented in the fuel cycle simulation code TIRELIRE-STRATEGIE that we developed at EDF R and D Division for nuclear power scenario studies. Then we show and comment the results of TIRELIRE-STRATEGIE calculation of a transition scenario between the current French nuclear fleet, and a fast reactor fleetmore » entirely deployed towards the end of the 21. century, consistently with the EDF prospective view, with 3 options for the minor actinides management:1) vitrified with fission products to be sent to the final disposal; 2) extracted together with plutonium from the spent fuel to be transmuted in Generation IV fast reactors; 3) eventually extracted separately from plutonium to be incinerated in a ADSs double stratum. The comparison of nuclear fuel cycle material fluxes and inventories between these options shows that ADSs are not more efficient than critical fast reactors for reducing the high level waste radio-toxicity; that minor actinides inventory and fluxes in the fuel cycle are more than twice as high in case of a double ADSs stratum than in case of minor actinides transmutation in Generation IV FBRs; and that about fourteen 400 MWth ADS are necessary to incinerate minor actinides issued from a 60 GWe Generation IV fast reactor fleet, corresponding to the current French nuclear fleet installed power. (authors)« less

Transuranic inventory reduction in repository by partitioning and transmutation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kang, C.H.; Kazimi, M.S.

1992-01-01

The promise of a new reprocessing technology and the issuance of Environmental Protection Agency (EPA) and U.S. Nuclear Regulatory Commission regulations concerning a geologic repository rekindle the interest in partitioning and transmutation of transuranic (TRU) elements from discharged reactor fuel as a high level waste management option. This paper investigates the TRU repository inventory reduction capability of the proposed advanced liquid metal reactors (ALMRs) and integral fast reactors (IFRs) as well as the plutonium recycled light water reactors (LWRs).

Closed DTU fuel cycle with Np recycle and waste transmutation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Beller, D.E.; Sailor, W.C.; Venneri, F.

1999-09-01

A nuclear energy scenario for the 21st century that included a denatured thorium-uranium-oxide (DTU) fuel cycle and new light water reactors (LWRs) supported by accelerator-driven transmutation of waste (ATW) systems was previously described. This coupled system with the closed DTU fuel cycle provides several improvements beyond conventional LWR (CLWR) (once-through, UO{sub 2} fuel) nuclear technology: increased proliferation resistance, reduced waste, and efficient use of natural resources. However, like CLWR fuel cycles, the spent fuel in the first one-third core discharged after startup contains higher-quality Pu than the equilibrium fuel cycle. To eliminate this high-grade Pu, Np is separated and recycledmore » with Th and U--rather than with higher actinides [(HA) including Pu]. The presence of Np in the LWR feed greatly increases the production of {sup 238}Pu so that a few kilograms of Pu generated enough alpha-decay heat that the separated Pu is highly resistant to proliferation. This alternate process also simplifies the pyrochemical separation of fuel elements (Th and U) from HAs. To examine the advantages of this concept, the authors modeled a US deployment scenario for nuclear energy that includes DTU-LWRs plus ATW`s to burn the actinides produced by these LWRs and to close the back-end of the DTU fuel cycle.« less

MYRRHA: A multipurpose nuclear research facility

NASA Astrophysics Data System (ADS)

Baeten, P.; Schyns, M.; Fernandez, Rafaël; De Bruyn, Didier; Van den Eynde, Gert

2014-12-01

MYRRHA (Multi-purpose hYbrid Research Reactor for High-tech Applications) is a multipurpose research facility currently being developed at SCK•CEN. MYRRHA is based on the ADS (Accelerator Driven System) concept where a proton accelerator, a spallation target and a subcritical reactor are coupled. MYRRHA will demonstrate the ADS full concept by coupling these three components at a reasonable power level to allow operation feedback. As a flexible irradiation facility, the MYRRHA research facility will be able to work in both critical as subcritical modes. In this way, MYRRHA will allow fuel developments for innovative reactor systems, material developments for GEN IV and fusion reactors, and radioisotope production for medical and industrial applications. MYRRHA will be cooled by lead-bismuth eutectic and will play an important role in the development of the Pb-alloys technology needed for the LFR (Lead Fast Reactor) GEN IV concept. MYRRHA will also contribute to the study of partitioning and transmutation of high-level waste. Transmutation of minor actinides (MA) can be completed in an efficient way in fast neutron spectrum facilities, so both critical reactors and subcritical ADS are potential candidates as dedicated transmutation systems. However critical reactors heavily loaded with fuel containing large amounts of MA pose reactivity control problems, and thus safety problems. A subcritical ADS operates in a flexible and safe manner, even with a core loading containing a high amount of MA leading to a high transmutation rate. In this paper, the most recent developments in the design of the MYRRHA facility are presented.

FCRD Transmutation Fuels Handbook 2015

DOE Office of Scientific and Technical Information (OSTI.GOV)

Janney, Dawn Elizabeth; Papesch, Cynthia Ann

2015-09-01

Transmutation of minor actinides such as Np, Am, and Cm in spent nuclear fuel is of international interest because of its potential for reducing the long-term health and safety hazards caused by the radioactivity of the spent fuel. One important approach to transmutation (currently being pursued by the DOE Fuel Cycle Research & Development Advanced Fuels Campaign) involves incorporating the minor actinides into U-Pu-Zr alloys, which can be used as fuel in fast reactors. It is, therefore, important to understand the properties of U-Pu-Zr alloys, both with and without minor actinide additions. In addition to requiring extensive safety precautions, alloysmore » containing U and Pu are difficult to study for numerous reasons, including their complex phase transformations, characteristically sluggish phase-transformation kinetics, tendency to produce experimental results that vary depending on the histories of individual samples, and sensitivity to contaminants such as oxygen in concentrations below a hundred parts per million. Many of the experimental measurements were made before 1980, and the level of documentation for experimental methods and results varies widely. It is, therefore, not surprising that little is known with certainty about U-Pu-Zr alloys, and that general acceptance of results sometimes indicates that there is only a single measurement for a particular property. This handbook summarizes currently available information about U, Pu, Zr, and alloys of two or three of these elements. It contains information about phase diagrams and related information (including phases and phase transformations); heat capacity, entropy, and enthalpy; thermal expansion; and thermal conductivity and diffusivity. In addition to presenting information about materials properties, it attempts to provide information about how well the property is known and how much variation exists between measurements. Although the handbook includes some references to publications about modeling, its primary focus is experimental data. Most of the data has been published elsewhere (although scattered throughout numerous references, some quite obscure); however, some data is presented here for the first time.« less

Enhancing BWR proliferation resistance fuel with minor actinides

NASA Astrophysics Data System (ADS)

Chang, Gray S.

2009-03-01

To reduce spent fuel for storage and enhance the proliferation resistance for the intermediate-term, there are two major approaches (a) increase the discharged spent fuel burnup in the advanced light water reactor- LWR (Gen-III Plus), which not only can reduce the spent fuel for storage, but also increase the 238Pu isotopes ratio to enhance the proliferation resistance, and (b) use of transuranic nuclides ( 237Np and 241Am) in the high burnup fuel, which can drastically increase the proliferation resistance isotope ratio of 238Pu/Pu. For future advanced nuclear systems, minor actinides (MA) are viewed more as a resource to be recycled, and transmuted to less hazardous and possibly more useful forms, rather than simply disposed of as a waste stream in an expensive repository facility. As a result, MAs play a much larger part in the design of advanced systems and fuel cycles, not only as additional sources of useful energy, but also as direct contributors to the reactivity control of the systems into which they are incorporated. In the study, a typical boiling water reactor (BWR) fuel unit lattice cell model with UO 2 fuel pins will be used to investigate the effectiveness of minor actinide reduction approach (MARA) for enhancing proliferation resistance and improving the fuel cycle performance in the intermediate-term goal for future nuclear energy systems. To account for the water coolant density variation from the bottom (0.76 g/cm 3) to the top (0.35 g/cm 3) of the core, the axial coolant channel and fuel pin were divided to 24 nodes. The MA transmutation characteristics at different elevations were compared and their impact on neutronics criticality discussed. The concept of MARA, which involves the use of transuranic nuclides ( 237Np and/or 241Am), significantly increases the 238Pu/Pu ratio for proliferation resistance, as well as serves as a burnable absorber to hold-down the initial excess reactivity. It is believed that MARA can play an important role in atoms for peace and the intermediate-term of nuclear energy reconnaissance.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dunn, Darrell; Poinssot, Christophe; Begg, Bruce

Management of nuclear waste remains an important international topic that includes reprocessing of commercial nuclear fuel, waste-form design and development, storage and disposal packaging, the process of repository site selection, system design, and performance assessment. Requirements to manage and dispose of materials from the production of nuclear weapons, and the renewed interest in nuclear power, in particular through the Generation IV Forum and the Advanced Fuel Cycle Initiative, can be expected to increase the need for scientific advances in waste management. A broad range of scientific and engineering disciplines is necessary to provide safe and effective solutions and address complexmore » issues. This volume offers an interdisciplinary perspective on materials-related issues associated with nuclear waste management programs. Invited and contributed papers cover a wide range of topics including studies on: spent fuel; performance assessment and models; waste forms for low- and intermediate-level waste; ceramic and glass waste forms for plutonium and high-level waste; radionuclides; containers and engineered barriers; disposal environments and site characteristics; and partitioning and transmutation.« less

Accelerator-driven Transmutation of Waste

NASA Astrophysics Data System (ADS)

Venneri, Francesco

1998-04-01

Nuclear waste from commercial power plants contains large quantities of plutonium, other fissionable actinides, and long-lived fission products that are potential proliferation concerns and create challenges for the long-term storage. Different strategies for dealing with nuclear waste are being followed by various countries because of their geologic situations and their views on nuclear energy, reprocessing and non-proliferation. The current United States policy is to store unprocessed spent reactor fuel in a geologic repository. Other countries are opting for treatment of nuclear waste, including partial utilization of the fissile material contained in the spent fuel, prior to geologic storage. Long-term uncertainties are hampering the acceptability and eventual licensing of a geologic repository for nuclear spent fuel in the US, and driving up its cost. The greatest concerns are with the potential for radiation release and exposure from the spent fuel for tens of thousands of years and the possible diversion and use of the actinides contained in the waste for weapons construction. Taking advantage of the recent breakthroughs in accelerator technology and of the natural flexibility of subcritical systems, the Accelerator-driven Transmutation of Waste (ATW) concept offers the United States and other countries the possibility to greatly reduce plutonium, higher actinides and environmentally hazardous fission products from the waste stream destined for permanent storage. ATW does not eliminate the need for, but instead enhances the viability of permanent waste repositories. Far from being limited to waste destruction, the ATW concept also brings to the table new technologies that could be relevant for next-generation power producing reactors. In the ATW concept, spent fuel would be shipped to the ATW site where the plutonium, transuranics and selected long-lived fission products would be destroyed by fission or transmutation in their first and only pass through the facility, using an accelerator-driven subcritical burner cooled by liquid lead/bismuth and limited pyrochemical treatment of the spent fuel and residual waste. This approach contrasts with the present-day practices of aqueous reprocessing (Europe and Japan), in which high purity plutonium is produced and used in the fabrication of fresh mixed oxide fuel (MOX) that is shipped off-site for use in light water reactors.

Breeding of 233U in the thorium-uranium fuel cycle in VVER reactors using heavy water

NASA Astrophysics Data System (ADS)

Marshalkin, V. E.; Povyshev, V. M.

2015-12-01

A method is proposed for achieving optimal neutron kinetics and efficient isotope transmutation in the 233U-232Th oxide fuel of water-moderated reactors with variable water composition (D2O, H2O) that ensures breeding of the 233U and 235U isotopes. The method is comparatively simple to implement.

Thermal Stability of Acetohydroxamic Acid/Nitric Acid Solutions

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rudisill, T.S.

2002-03-13

The transmutation of transuranic actinides and long-lived fission products in spent commercial nuclear reactor fuel has been proposed as one element of the Advanced Accelerator Applications Program. Preparation of targets for irradiation in an accelerator-driven subcritical reactor would involve dissolution of the fuel and separation of uranium, technetium, and iodine from the transuranic actinides and other fission products. The UREX solvent extraction process is being developed to reject and isolate the transuranic actinides in the acid waste stream by scrubbing with acetohydroxamic acid (AHA). To ensure that a runaway reaction will not occur between nitric acid and AHA, an analoguemore » of hydroxyl amine, thermal stability tests were performed to identify if any processing conditions could lead to a runaway reaction.« less

Reliability assessment of MVP-BURN and JENDL-4.0 related to nuclear transmutation of light platinum group elements

NASA Astrophysics Data System (ADS)

Terashima, Atsunori; Nilsson, Mikael; Ozawa, Masaki; Chiba, Satoshi

2017-09-01

The Aprés ORIENT research program, as a concept of advanced nuclear fuel cycle, was initiated in FY2011 aiming at creating stable, highly-valuable elements by nuclear transmutation from ↓ssion products. In order to simulate creation of such elements by (n, γ) reaction succeeded by β- decay in reactors, a continuous-energy Monte Carlo burnup calculation code MVP-BURN was employed. Then, it is one of the most important tasks to con↓rm the reliability of MVP-BURN code and evaluated neutron cross section library. In this study, both an experiment of neutron activation analysis in TRIGA Mark I reactor at University of California, Irvine and the corresponding burnup calculation using MVP-BURN code were performed for validation of the simulation on transmutation of light platinum group elements. Especially, some neutron capture reactions such as 102Ru(n, γ)103Ru, 104Ru(n, γ)105Ru, and 108Pd(n, γ)109Pd were dealt with in this study. From a comparison between the calculation (C) and the experiment (E) about 102Ru(n, γ)103Ru, the deviation (C/E-1) was signi↓cantly large. Then, it is strongly suspected that not MVP-BURN code but the neutron capture cross section of 102Ru belonging to JENDL-4.0 used in this simulation have made the big di↑erence as (C/E-1) >20%.

Natural Transmutation of Actinides via the Fission Reaction in the Closed Thorium-Uranium-Plutonium Fuel Cycle

NASA Astrophysics Data System (ADS)

Marshalkin, V. Ye.; Povyshev, V. M.

2017-12-01

It is shown for a closed thorium-uranium-plutonium fuel cycle that, upon processing of one metric ton of irradiated fuel after each four-year campaign, the radioactive wastes contain 54 kg of fission products, 0.8 kg of thorium, 0.10 kg of uranium isotopes, 0.005 kg of plutonium isotopes, 0.002 kg of neptunium, and "trace" amounts of americium and curium isotopes. This qualitatively simplifies the handling of high-level wastes in nuclear power engineering.

Gas core reactors for actinide transmutation and breeder applications

NASA Technical Reports Server (NTRS)

Clement, J. D.; Rust, J. H.

1978-01-01

This work consists of design power plant studies for four types of reactor systems: uranium plasma core breeder, uranium plasma core actinide transmuter, UF6 breeder and UF6 actinide transmuter. The plasma core systems can be coupled to MHD generators to obtain high efficiency electrical power generation. A 1074 MWt UF6 breeder reactor was designed with a breeding ratio of 1.002 to guard against diversion of fuel. Using molten salt technology and a superheated steam cycle, an efficiency of 39.2% was obtained for the plant and the U233 inventory in the core and heat exchangers was limited to 105 Kg. It was found that the UF6 reactor can produce high fluxes (10 to the 14th power n/sq cm-sec) necessary for efficient burnup of actinide. However, the buildup of fissile isotopes posed severe heat transfer problems. Therefore, the flux in the actinide region must be decreased with time. Consequently, only beginning-of-life conditions were considered for the power plant design. A 577 MWt UF6 actinide transmutation reactor power plant was designed to operate with 39.3% efficiency and 102 Kg of U233 in the core and heat exchanger for beginning-of-life conditions.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cappiello, M.; Hobbins, R.; Penny, K.

As part of the Department of Energy Advanced Fuel Cycle program, a series of fuels development irradiation tests have been performed in the Advanced Test Reactor (ATR) at the Idaho National Laboratory. These tests are providing excellent data for advanced fuels development. The program is focused on the transmutation of higher actinides which best can be accomplished in a sodium-cooled fast reactor. Because a fast test reactor is no longer available in the US, a special test vehicle is used to achieve near-prototypic fast reactor conditions (neutron spectra and temperature) for use in ATR (a water-cooled thermal reactor). As partmore » of the testing program, there were many successful tests of advanced fuels including metals and ceramics. Recently however, there have been three experimental campaigns using metal fuels that experienced failure during irradiation. At the request of the program, an independent review committee was convened to review the post-test analyses performed by the fuels development team, to assess the conclusions of the team for the cause of the failures, to assess the adequacy and completeness of the analyses, to identify issues that were missed, and to make recommendations for improvements in the design and operation of future tests. Although there is some difference of opinion, the review committee largely agreed with the conclusions of the fuel development team regarding the cause of the failures. For the most part, the analyses that support the conclusions are sufficient.« less

Breeding of {sup 233}U in the thorium–uranium fuel cycle in VVER reactors using heavy water

DOE Office of Scientific and Technical Information (OSTI.GOV)

Marshalkin, V. E., E-mail: marshalkin@vniief.ru; Povyshev, V. M.

A method is proposed for achieving optimal neutron kinetics and efficient isotope transmutation in the {sup 233}U–{sup 232}Th oxide fuel of water-moderated reactors with variable water composition (D{sub 2}O, H{sub 2}O) that ensures breeding of the {sup 233}U and {sup 235}U isotopes. The method is comparatively simple to implement.

Transmutation Fuel Fabrication-Fiscal Year 2016

DOE Office of Scientific and Technical Information (OSTI.GOV)

Fielding, Randall Sidney; Grover, Blair Kenneth

ABSTRACT Nearly all of the metallic fuel that has been irradiated and characterized by the Advanced Fuel Campaign, and its earlier predecessors, has been arc cast. Arc casting is a very flexible method of casting lab scale quantities of materials. Although the method offers flexibility, it is an operator dependent process. Small changes in parameter space or alloy composition may affect how the material is cast. This report provides a historical insight in how the casting process has been modified over the history of the advanced fuels campaign as well as the physical parameters of the fuels cast in fiscalmore » year 2016.« less

FCRD Advanced Reactor (Transmutation) Fuels Handbook

DOE Office of Scientific and Technical Information (OSTI.GOV)

Janney, Dawn Elizabeth; Papesch, Cynthia Ann

2016-09-01

Transmutation of minor actinides such as Np, Am, and Cm in spent nuclear fuel is of international interest because of its potential for reducing the long-term health and safety hazards caused by the radioactivity of the spent fuel. One important approach to transmutation (currently being pursued by the DOE Fuel Cycle Research & Development Advanced Fuels Campaign) involves incorporating the minor actinides into U-Pu-Zr alloys, which can be used as fuel in fast reactors. U-Pu-Zr alloys are well suited for electrolytic refining, which leads to incorporation rare-earth fission products such as La, Ce, Pr, and Nd. It is, therefore, importantmore » to understand not only the properties of U-Pu-Zr alloys but also those of U-Pu-Zr alloys with concentrations of minor actinides (Np, Am) and rare-earth elements (La, Ce, Pr, and Nd) similar to those in reprocessed fuel. In addition to requiring extensive safety precautions, alloys containing U, Pu, and minor actinides (Np and Am) are difficult to study for numerous reasons, including their complex phase transformations, characteristically sluggish phasetransformation kinetics, tendency to produce experimental results that vary depending on the histories of individual samples, rapid oxidation, and sensitivity to contaminants such as oxygen in concentrations below a hundred parts per million. Although less toxic, rare-earth elements such as La, Ce, Pr, and Nd are also difficult to study for similar reasons. Many of the experimental measurements were made before 1980, and the level of documentation for experimental methods and results varies widely. It is, therefore, not surprising that little is known with certainty about U-Pu-Zr alloys, particularly those that also contain minor actinides and rare-earth elements. General acceptance of results commonly indicates that there is only a single measurement for a particular property. This handbook summarizes currently available information about U, Pu, Zr, Np, Am, La, Ce, Pr, and Nd and alloys of two or three of these elements. It contains information about phase diagrams and related information (including phases and phase transformations); heat capacity, entropy, and enthalpy; thermal expansion; and thermal conductivity and diffusivity. In addition to presenting information about materials properties, the handbook attempts to provide information about how well the property is known and how much variation exists between measurements. Although it includes some results from models, its primary focus is experimental data. The Handbook is organized in two sections: one with information about the U-Pu-Zr ternary and one with information about other elements and binary and vi ternary alloys in the U-Np-Pu-Am-La-Ce-Pr-Nd-Zr system. Within each section, information about elements is presented first, followed by information about binary alloys, then information about ternary alloys. The order in which the elements in each alloy are mentioned follows the order in the first sentence of this paragraph. Much of the information on the U-Pu-Zr system repeats information from the FCRD Transmutation Fuels Handbook 2015. Most of the other data has been published elsewhere (although scattered throughout numerous references, some quite obscure); however, some data from Idaho National Laboratory is presented here for the first time. As the FCRD programmatic mission evolves, future editions of this handbook will begin to include other advanced reactor fuel designs and compositions. Hence, the title of the handbook will transition to the Advanced Reactor Fuels Handbook.« less

Uranium oxide fuel cycle analysis in VVER-1000 with VISTA simulation code

NASA Astrophysics Data System (ADS)

Mirekhtiary, Seyedeh Fatemeh; Abbasi, Akbar

2018-02-01

The VVER-1000 Nuclear power plant generates about 20-25 tons of spent fuel per year. In this research, the fuel transmutation of Uranium Oxide (UOX) fuel was calculated by using of nuclear fuel cycle simulation system (VISTA) code. In this simulation, we evaluated the back end components fuel cycle. The back end component calculations are Spent Fuel (SF), Actinide Inventory (AI) and Fission Product (FP) radioisotopes. The SF, AI and FP values were obtained 23.792178 ton/y, 22.811139 ton/y, 0.981039 ton/y, respectively. The obtained value of spent fuel, major actinide, and minor actinide and fission products were 23.8 ton/year, 22.795 ton/year, 0.024 ton/year and 0.981 ton/year, respectively.

Status of the French Research on Partitioning and Transmutation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Warin, Dominique

2007-07-01

The global energy context pleads in favor of a sustainable development of nuclear energy since the demand for energy will likely increase, whereas resources will tend to get scarcer and the prospect of global warming will drive down the consumption of fossil fuel sources. How we deal with radioactive waste is crucial in this context. The production of nuclear energy in France has been associated, since its inception, with the optimization of radioactive waste management, including the partitioning and the recycling of recoverable energetic materials. The public's concern regarding the long-term waste management made the French Government prepare and passmore » the December 1991 Law, requesting in particular, the study for fifteen years of solutions for still minimizing the quantity and the hazardousness of final waste, via partitioning and transmutation. At the end of these fifteen years of research, it is considered that partitioning techniques, which have been validated on real solutions, are at disposal. Indeed, aqueous process for separation of minor actinides from the PUREX raffinate has been brought to a point where there is reasonable assurance that industrial deployment can be successful. A key experiment has been the successful kilogram scale trials in the CEA-Marcoule Atalante facility in 2005 and this result, together with the results obtained in the frame of the successive European projects, constitutes a considerable step forward. For transmutation, CEA has conducted programs proving the feasibility of the elimination of minor actinides and fission products: fabrication of specific targets and fuels for transmutation tests in the HFR and Phenix reactors, neutronics and technology studies for critical reactors and ADS developments. Scenario studies have also allowed assessing the feasibility, at the level of cycle and fuel facilities, and the efficiency of transmutation in terms of the quantitative reduction of the final waste inventory depending of the reactor fleet (PWR-FR-ADS). Important results are now available concerning the possibility of significantly reducing the quantity and the radiotoxicity of long-lived waste in association with a sustainable development of nuclear energy. As France has confirmed its long-term approach to nuclear energy, the most effective implementation of P and T of minor actinides relies on the fast neutron GEN IV systems, which are designed to recycle and manage their own actinides. The perspective to deploy a first series of such systems around 2040 supports the idea that progress is being made: the long-term waste would only be made up of fission products, with very low amounts of minor actinides. In this sense, the new waste management law passed by the French Parliament on June 28, 2006, demands that P and T research continues in strong connection to GEN IV systems and ADS development and allowing the assessment of the industrial perspectives of such systems in 2012 and to put into operation a transmutation demo facility in 2020. (author)« less

A new approach to nuclear fuel safeguard enhancement through radionuclide profiling

NASA Astrophysics Data System (ADS)

Peterson, Aaron Dawon

The United States has led the effort to promote peaceful use of nuclear power amongst states actively utilizing it as well as those looking to deploy the technology in the near future. With the attraction being demonstrated by various countries towards nuclear power comes the concern that a nation may have military aspirations for the use of nuclear energy. The International Atomic Energy Agency (IAEA) has established nuclear safeguard protocols and procedures to mitigate nuclear proliferation. The work herein proposed a strategy to further enhance existing safeguard protocols by considering safeguard in nuclear fuel design. The strategy involved the use of radionuclides to profile nuclear fuels. Six radionuclides were selected as identifier materials. The decay and transmutation of these radionuclides were analyzed in reactor operation environment. MCNPX was used to simulate a reactor core. The perturbation in reactivity of the core due to the loading of the radionuclides was insignificant. The maximum positive and negative reactivity change induced was at day 1900 with a value of 0.00185 +/- 0.00256 and at day 2000 with -0.00441 +/- 0.00249, respectively. The mass of the radionuclides were practically unaffected by transmutation in the core; the change in radionuclide inventory was dominated by natural decay. The maximum material lost due to transmutation was 1.17% in Eu154. Extraneous signals from fission products identical to the radionuclide compromised the identifier signals. Eu154 saw a maximum intensity change at EOC and 30 days post-irradiation of 1260% and 4545%, respectively. Cs137 saw a minimum change of 12% and 89%, respectively. Mitigation of the extraneous signals is cardinal to the success of the proposed strategy. The predictability of natural decay provides a basis for the characterization of the signals from the radionuclide.

Fast reactor core concepts to improve transmutation efficiency

DOE Office of Scientific and Technical Information (OSTI.GOV)

Fujimura, Koji; Kawashima, Katsuyuki; Itooka, Satoshi

Fast Reactor (FR) core concepts to improve transmutation efficiency were conducted. A heterogeneous MA loaded core was designed based on the 1000MWe-ABR breakeven core. The heterogeneous MA loaded core with Zr-H loaded moderated targets had a better transmutation performance than the MA homogeneous loaded core. The annular pellet rod design was proposed as one of the possible design options for the MA target. It was shown that using annular pellet MA rods mitigates the self-shielding effect in the moderated target so as to enhance the transmutation rate.

Reactor-based management of used nuclear fuel: assessment of major options.

PubMed

Finck, Phillip J; Wigeland, Roald A; Hill, Robert N

2011-01-01

This paper discusses the current status of the ongoing Advanced Fuel Cycle Initiative (AFCI) program in the U.S. Department of Energy that is investigating the potential for using the processing and recycling of used nuclear fuel to improve radioactive waste management, including used fuel. A key element of the strategies is to use nuclear reactors for further irradiation of recovered chemical elements to transmute certain long-lived highly-radioactive isotopes into less hazardous isotopes. Both thermal and fast neutron spectrum reactors are being studied as part of integrated nuclear energy systems where separations, transmutation, and disposal are considered. Radiotoxicity is being used as one of the metrics for estimating the hazard of used fuel and the processing of wastes resulting from separations and recycle-fuel fabrication. Decay heat from the used fuel and/or wastes destined for disposal is used as a metric for use of a geologic repository. Results to date indicate that the most promising options appear to be those using fast reactors in a repeated recycle mode to limit buildup of higher actinides, since the transuranic elements are a key contributor to the radiotoxicity and decay heat. Using such an approach, there could be much lower environmental impact from the high-level waste as compared to direct disposal of the used fuel, but there would likely be greater generation of low-level wastes that will also require disposal. An additional potential waste management benefit is having the ability to tailor waste forms and contents to one or more targeted disposal environments (i.e., to be able to put waste in environments best-suited for the waste contents and forms). Copyright © 2010 Health Physics Society

THE ECONOMICS OF REPROCESSING vs DIRECT DISPOSAL OF SPENT NUCLEAR FUEL

DOE Office of Scientific and Technical Information (OSTI.GOV)

Matthew Bunn; Steve Fetter; John P. Holdren

This report assesses the economics of reprocessing versus direct disposal of spent nuclear fuel. The breakeven uranium price at which reprocessing spent nuclear fuel from existing light-water reactors (LWRs) and recycling the resulting plutonium and uranium in LWRs would become economic is assessed, using central estimates of the costs of different elements of the nuclear fuel cycle (and other fuel cycle input parameters), for a wide range of range of potential reprocessing prices. Sensitivity analysis is performed, showing that the conclusions reached are robust across a wide range of input parameters. The contribution of direct disposal or reprocessing and recyclingmore » to electricity cost is also assessed. The choice of particular central estimates and ranges for the input parameters of the fuel cycle model is justified through a review of the relevant literature. The impact of different fuel cycle approaches on the volume needed for geologic repositories is briefly discussed, as are the issues surrounding the possibility of performing separations and transmutation on spent nuclear fuel to reduce the need for additional repositories. A similar analysis is then performed of the breakeven uranium price at which deploying fast neutron breeder reactors would become competitive compared with a once-through fuel cycle in LWRs, for a range of possible differences in capital cost between LWRs and fast neutron reactors. Sensitivity analysis is again provided, as are an analysis of the contribution to electricity cost, and a justification of the choices of central estimates and ranges for the input parameters. The equations used in the economic model are derived and explained in an appendix. Another appendix assesses the quantities of uranium likely to be recoverable worldwide in the future at a range of different possible future prices.« less

Transmutation of Nuclear Waste and the future MYRRHA Demonstrator

NASA Astrophysics Data System (ADS)

Mueller, Alex C.

2013-03-01

While a considerable and world-wide growth of the nuclear share in the global energy mix is desirable for many reasons, there are also, in particular in the "old world" major objections. These are both concerns about safety, in particular in the wake of the Fukushima nuclear accident and concerns about the long-term burden that is constituted by the radiotoxic waste from the spent fuel. With regard to the second topic, the present contribution will outline the concept of Partitioning & Transmutation (P&T), as scientific and technological answer. Deployment of P&T may use dedicated "Transmuter" or "Burner" reactors, using a fast neutron spectrum. For the transmutation of waste with a large content (up to 50%) of (very long-lived) Minor Actinides, a sub-critical reactor, using an external neutron source is a most attractive solution. It is constituted by coupling a proton accelerator, a spallation target and a subcritical core. This promising new technology is named ADS, for accelerator-driven system. The present paper aims at a short introduction into the field that has been characterized by a high collaborative activity during the last decade in Europe, in order to focus, in its later part, on the MYRRHA project as the European ADS technology demonstrator.

The physics design of accelerator-driven transmutation systems

DOE Office of Scientific and Technical Information (OSTI.GOV)

Venneri, F.

1995-10-01

Nuclear systems under study in the Los Alamos Accelerator-Driven Transmutation Technology program (ADTT) will allow the destruction of nuclear spent fuel and weapons-return plutonium, as well as the production of nuclear energy from the thorium cycle, without a long-lived radioactive waste stream. The subcritical systems proposed represent a radical departure from traditional nuclear concepts (reactors), yet the actual implementation of ADTT systems is based on modest extrapolations of existing technology. These systems strive to keep the best that the nuclear technology has developed over the years, within a sensible conservative design envelope and eventually manage to offer a safe, lessmore » expensive and more environmentally sound approach to nuclear power.« less

A feasibility study of reactor-based deep-burn concepts.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kim, T. K.; Taiwo, T. A.; Hill, R. N.

2005-09-16

A systematic assessment of the General Atomics (GA) proposed Deep-Burn concept based on the Modular Helium-Cooled Reactor design (DB-MHR) has been performed. Preliminary benchmarking of deterministic physics codes was done by comparing code results to those from MONTEBURNS (MCNP-ORIGEN) calculations. Detailed fuel cycle analyses were performed in order to provide an independent evaluation of the physics and transmutation performance of the one-pass and two-pass concepts. Key performance parameters such as transuranic consumption, reactor performance, and spent fuel characteristics were analyzed. This effort has been undertaken in close collaborations with the General Atomics design team and Brookhaven National Laboratory evaluation team.more » The study was performed primarily for a 600 MWt reference DB-MHR design having a power density of 4.7 MW/m{sup 3}. Based on parametric and sensitivity study, it was determined that the maximum burnup (TRU consumption) can be obtained using optimum values of 200 {micro}m and 20% for the fuel kernel diameter and fuel packing fraction, respectively. These values were retained for most of the one-pass and two-pass design calculations; variation to the packing fraction was necessary for the second stage of the two-pass concept. Using a four-batch fuel management scheme for the one-pass DB-MHR core, it was possible to obtain a TRU consumption of 58% and a cycle length of 286 EFPD. By increasing the core power to 800 MWt and the power density to 6.2 MW/m{sup 3}, it was possible to increase the TRU consumption to 60%, although the cycle length decreased by {approx}64 days. The higher TRU consumption (burnup) is due to the reduction of the in-core decay of fissile Pu-241 to Am-241 relative to fission, arising from the higher power density (specific power), which made the fuel more reactivity over time. It was also found that the TRU consumption can be improved by utilizing axial fuel shuffling or by operating with lower material temperatures (colder core). Results also showed that the transmutation performance of the one-pass deep-burn concept is sensitive to the initial TRU vector, primarily because longer cooling time reduces the fissile content (Pu-241 specifically.) With a cooling time of 5 years, the TRU consumption increases to 67%, while conversely, with 20-year cooling the TRU consumption is about 58%. For the two-pass DB-MHR (TRU recycling option), a fuel packing fraction of about 30% is required in the second pass (the recycled TRU). It was found that using a heterogeneous core (homogeneous fuel element) concept, the TRU consumption is dependent on the cooling interval before the 2nd pass, again due to Pu-241 decay during the time lag between the first pass fuel discharge and the second pass fuel charge. With a cooling interval of 7 years (5 and 2 years before and after reprocessing) a TRU consumption of 55% is obtained. With an assumed ''no cooling'' interval, the TRU consumption is 63%. By using a cylindrical core to reduce neutron leakage, TRU consumption of the case with 7-year cooling interval increases to 58%. For a two-pass concept using a heterogeneous fuel element (and homogeneous core) with first and second pass volume ratio of 2:1, the TRU consumption is 62.4%. Finally, the repository loading benefits arising from the deep-burn and Inert Matrix Fuel (IMF) concepts were estimated and compared, for the same initial TRU vector. The DB-MHR concept resulted in slightly higher TRU consumption and repository loading benefit compared to the IMF concept (58.1% versus 55.1% for TRU consumption and 2.0 versus 1.6 for estimated repository loading benefit).« less

Measurement and Analysis Plan for Investigation of Spent-Fuel Assay Using Lead Slowing-Down Spectroscopy

DOE Office of Scientific and Technical Information (OSTI.GOV)

Smith, Leon E.; Haas, Derek A.; Gavron, Victor A.

2009-09-25

Under funding from the Department of Energy Office of Nuclear Energy’s Materials, Protection, Accounting, and Control for Transmutation (MPACT) program (formerly the Advanced Fuel Cycle Initiative Safeguards Campaign), Pacific Northwest National Laboratory (PNNL) and Los Alamos National Laboratory (LANL) are collaborating to study the viability of lead slowing-down spectroscopy (LSDS) for spent-fuel assay. Based on the results of previous simulation studies conducted by PNNL and LANL to estimate potential LSDS performance, a more comprehensive study of LSDS viability has been defined. That study includes benchmarking measurements, development and testing of key enabling instrumentation, and continued study of time-spectra analysis methods.more » This report satisfies the requirements for a PNNL/LANL deliverable that describes the objectives, plans and contributing organizations for a comprehensive three-year study of LSDS for spent-fuel assay. This deliverable was generated largely during the LSDS workshop held on August 25-26, 2009 at Rensselaer Polytechnic Institute (RPI). The workshop itself was a prominent milestone in the FY09 MPACT project and is also described within this report.« less

Transmutation of planar media singularities in a conformal cloak.

PubMed

Liu, Yichao; Mukhtar, Musawwadah; Ma, Yungui; Ong, C K

2013-11-01

Invisibility cloaking based on optical transformation involves materials singularity at the branch cut points. Many interesting optical devices, such as the Eaton lens, also require planar media index singularities in their implementation. We show a method to transmute two singularities simultaneously into harmless topological defects formed by anisotropic permittivity and permeability tensors. Numerical simulation is performed to verify the functionality of the transmuted conformal cloak consisting of two kissing Maxwell fish eyes.

The procedure and results of calculations of the equilibrium isotopic composition of a demonstration subcritical molten salt reactor

NASA Astrophysics Data System (ADS)

Nevinitsa, V. A.; Dudnikov, A. A.; Blandinskiy, V. Yu.; Balanin, A. L.; Alekseev, P. N.; Titarenko, Yu. E.; Batyaev, V. F.; Pavlov, K. V.; Titarenko, A. Yu.

2015-12-01

A subcritical molten salt reactor with an external neutron source is studied computationally as a facility for incineration and transmutation of minor actinides from spent nuclear fuel of reactors of VVER-1000 type and for producing 233U from 232Th. The reactor configuration is chosen, the requirements to be imposed on the external neutron source are formulated, and the equilibrium isotopic composition of heavy nuclides and the key parameters of the fuel cycle are calculated.

Transmutation of Radioactive Nuclear Waste — Present Status and Requirement for the Problem-Oriented Nuclear Database: Approach to Scheduling the Experiments (Reactor, Target, Blanket)

NASA Astrophysics Data System (ADS)

Artisyuk, V.; Ignatyuk, A.; Korovin, Yu.; Lopatkin, A.; Matveenko, I.; Stankovskiy, A.; Titarenko, Yu.

2005-05-01

Transmutation of nuclear wastes (Minor Actinides and Long-Lived Fission Products) remains an important option to reduce the burden of high-level waste on final waste disposal in deep geological structures. Accelerator-Driven Systems (ADS) are considered as possible candidates to perform transmutation due to their subcritical operation mode that eliminates some of the serious safety penalties unavoidable in critical reactors. Specific requirements to nuclear data necessary for ADS transmutation analysis is the main subject of the ISTC Project ♯2578 which started in 2004 to identify the areas of research priorities in the future. The present paper gives a summary of ongoing project stressing the importance of nuclear data for blanket performance (reactivity behavior with associated safety characteristics) and uncertainties that affect characteristics of neutron producing target.

Molten salt considerations for accelerator-driven subcritical fission to close the nuclear fuel cycle

NASA Astrophysics Data System (ADS)

Sooby, Elizabeth; Adams, Marvin; Baty, Austin; Gerity, James; McIntyre, Peter; Melconian, Karie; Phongikaroon, Supathorn; Pogue, Nathaniel; Sattarov, Akhdiyor; Simpson, Michael; Tripathy, Prabhat; Tsevkov, Pavel

2013-04-01

The host salt selection, molecular modeling, physical chemistry, and processing chemistry are presented here for an accelerator-driven subcritical fission in a molten salt core (ADSMS). The core is fueled solely with the transuranics (TRU) and long-lived fission products (LFP) from used nuclear fuel. The neutronics and salt composition are optimized to destroy the transuranics by fission and the long-lived fission products by transmutation. The cores are driven by proton beams from a strong-focusing cyclotron stack. One such ADSMS system can destroy the transuranics in the used nuclear fuel produced by a 1GWe conventional reactor. It uniquely provides a method to close the nuclear fuel cycle for green nuclear energy.

Progress on inert matrix fuels for minor actinide transmutation in fast reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bonnerot, Jean-Marc; Ferroud-Plattet, Marie-Pierre; Lamontagne, Jerome

2007-07-01

An extensive irradiation program has been devoted by CEA to the assessment of transmutation using minor actinide bearing inert support targets. A first irradiation experiment was performed in the fast neutron reactor Phenix, in parallel to other experiments carried out in the HFR and Siloe reactors, in order to assess the behavior under fast neutron flux of various materials intended as inert support matrix for transmutation targets. This experiment, which included the two steps MATINA 1 and MATINA 1A, was completed in 2004 and underwent complete post irradiation examinations (PIE) , whose results are presented in this paper. All themore » pure inert materials showed a satisfactory behavior under fast neutrons except Al{sub 2}O{sub 3} - which exhibits a swelling close to 11 vol. % after irradiation. In presence of UO{sub 2} fissile particles, MgAl{sub 2}O{sub 4} proved to be more stable in term of swelling as inert support than MgO and Al{sub 2}O{sub 3} matrices, under the same irradiation conditions. A second experiment ECRIX H in Phenix involving composite pellets with an MgO matrix and AmO{sub 2-x} particles was completed in 2006. The very first PIE results on ECRIX H are described in this paper. At the light of these first experiments, a second phase dedicated to the design optimization of the target was initiated and three new irradiation experiments - MATINA 2-3, CAMIX COCHIX in Phenix and HELIOS in HFR - were started in 2006 and 2007. (authors)« less

Reducing Actinide Production Using Inert Matrix Fuels

DOE Office of Scientific and Technical Information (OSTI.GOV)

Deinert, Mark

2017-08-23

The environmental and geopolitical problems that surround nuclear power stem largely from the longlived transuranic isotopes of Am, Cm, Np and Pu that are contained in spent nuclear fuel. New methods for transmuting these elements into more benign forms are needed. Current research efforts focus largely on the development of fast burner reactors, because it has been shown that they could dramatically reduce the accumulation of transuranics. However, despite five decades of effort, fast reactors have yet to achieve industrial viability. A critical limitation to this, and other such strategies, is that they require a type of spent fuel reprocessingmore » that can efficiently separate all of the transuranics from the fission products with which they are mixed. Unfortunately, the technology for doing this on an industrial scale is still in development. In this project, we explore a strategy for transmutation that can be deployed using existing, current generation reactors and reprocessing systems. We show that use of an inert matrix fuel to recycle transuranics in a conventional pressurized water reactor could reduce overall production of these materials by an amount that is similar to what is achievable using proposed fast reactor cycles. Furthermore, we show that these transuranic reductions can be achieved even if the fission products are carried into the inert matrix fuel along with the transuranics, bypassing the critical separations hurdle described above. The implications of these findings are significant, because they imply that inert matrix fuel could be made directly from the material streams produced by the commercially available PUREX process. Zirconium dioxide would be an ideal choice of inert matrix in this context because it is known to form a stable solid solution with both fission products and transuranics.« less

Comparison of actinide production in traveling wave and pressurized water reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Osborne, A.G.; Smith, T.A.; Deinert, M.R.

The geopolitical problems associated with civilian nuclear energy production arise in part from the accumulation of transuranics in spent nuclear fuel. A traveling wave reactor is a type of breed-burn reactor that could, if feasible, reduce the overall production of transuranics. In one possible configuration, a cylinder of natural or depleted uranium would be subjected to a fast neutron flux at one end. The neutrons would transmute the uranium, producing plutonium and higher actinides. Under the right conditions, the reactor could become critical, at which point a self-stabilizing fission wave would form and propagate down the length of the reactormore » cylinder. The neutrons from the fission wave would burn the fissile nuclides and transmute uranium ahead of the wave to produce additional fuel. Fission waves in uranium are driven largely by the production and fission of {sup 239}Pu. Simulations have shown that the fuel burnup can reach values greater than 400 MWd/kgIHM, before fission products poison the reaction. In this work we compare the production of plutonium and minor actinides produced in a fission wave to that of a UOX fueled light water reactor, both on an energy normalized basis. The nuclide concentrations in the spent traveling wave reactor fuel are computed using a one-group diffusion model and are verified using Monte Carlo simulations. In the case of the pressurized water reactor, a multi-group collision probability model is used to generate the nuclide quantities. We find that the traveling wave reactor produces about 0.187 g/MWd/kgIHM of transuranics compared to 0.413 g/MWd/kgIHM for a pressurized water reactor running fuel enriched to 4.95 % and burned to 50 MWd/kgIHM. (authors)« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Youinou, Gilles Jean-Michel

Neutron cross-sections characterize the way neutrons interact with matter. They are essential to most nuclear engineering projects and, even though theoretical progress has been made as far as the predictability of neutron cross-section models, measurements are still indispensable to meet tight design requirements for reduced uncertainties. Within the field of fission reactor technology, one can identify the following specializations that rely on the availability of accurate neutron cross-sections: (1) fission reactor design, (2) nuclear fuel cycles, (3) nuclear safety, (4) nuclear safeguards, (5) reactor monitoring and neutron fluence determination and (6) waste disposal and transmutation. In particular, the assessment ofmore » advanced fuel cycles requires an extensive knowledge of transuranics cross sections. Plutonium isotopes, but also americium, curium and up to californium isotope data are required with a small uncertainty in order to optimize significant features of the fuel cycle that have an impact on feasibility studies (e.g. neutron doses at fuel fabrication, decay heat in a repository, etc.). Different techniques are available to determine neutron cross sections experimentally, with the common denominator that a source of neutrons is necessary. It can either come from an accelerator that produces neutrons as a result of interactions between charged particles and a target, or it can come from a nuclear reactor. When the measurements are performed with an accelerator, they are referred to as differential since the analysis of the data provides the cross-sections for different discrete energies, i.e. σ(E i), and for the diffusion cross sections for different discrete angles. Another approach is to irradiate a very pure sample in a test reactor such as the Advanced Test Reactor (ATR) at INL and, after a given time, determine the amount of the different transmutation products. The precise characterization of the nuclide densities before and after neutron irradiation allows to infer energy-integrated neutron cross sections, i.e. ∫₀ ∞σ(E)φ(E)dE, where φ(E) is the neutron flux “seen” by the sample. This approach, which is usually defined and led by reactor physicists, is referred to as integral and is the object of this report. These two sources of information, i.e. differential and integral, are complementary and are used by the nuclear physicists in charge of producing the evaluated nuclear data files used by the nuclear community (ENDF, JEFF…). The generation of accurate nuclear data files requires an iterative process involving reactor physicists and nuclear data evaluators. This experimental program has been funded by the ATR National Scientific User Facility (ATR-NSUF) and by the DOE Office of Science in the framework of the Recovery Act. It has been given the name MANTRA for Measurement of Actinides Neutron TRAnsmutation.« less

FIRST-PRINCIPLES CALCULATIONS OF INTRINSIC DEFECTS AND Mg TRANSMUTANTS IN 3C-SiC

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hu, Shenyang Y.; Setyawan, Wahyu; Van Ginhoven, Renee M.

2013-09-25

Silicon carbide (SiC) possesses many desirable attributes for applications in high-temperature and neutron radiation environments. These attributes include excellent dimensional and thermodynamic stability, low activation, high strength, and high thermal conductivity. Therefore, SiC based materials draw broad attention as structural materials for the first wall (FW) and blanket in fusion power plants. Under the severe high-energy neutron environment of D-T fusion systems, SiC suffers significant transmutation resulting in both gaseous and metallic transmutants. Recent calculations by Sawan, et al. [2] predict that at a fast neutron dose of ~100 dpa, there will be about 0.5 at% Mg generated in SiCmore » through nuclear transmutation. Other transmutation products, including 0.15 at% Al, 0.2 at% Be and 2.2 at% He, also emerge. Formation and migration energies of point defects in 3C-SiC have been widely investigated using density functional theory (DFT). However, the properties of defects associated with transmutants are currently not well understood. Fundamental understanding of where the transmutation products go and how they affect microstructure evolution of SiC composites will help to predict property evolution and performance of SiC-based materials in fusion reactors.« less

Specification-based Error Recovery: Theory, Algorithms, and Usability

DTIC Science & Technology

2013-02-01

transmuting the specification to an implementation at run-time and reducing the performance overhead. A suite of techniques and tools were designed...in the specification, thereby transmuting the specification to an implementation at run-time and reducing the perfor- mance overhead. A suite of

The procedure and results of calculations of the equilibrium isotopic composition of a demonstration subcritical molten salt reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Nevinitsa, V. A., E-mail: Neviniza-VA@nrcki.ru; Dudnikov, A. A.; Blandinskiy, V. Yu.

2015-12-15

A subcritical molten salt reactor with an external neutron source is studied computationally as a facility for incineration and transmutation of minor actinides from spent nuclear fuel of reactors of VVER-1000 type and for producing {sup 233}U from {sup 232}Th. The reactor configuration is chosen, the requirements to be imposed on the external neutron source are formulated, and the equilibrium isotopic composition of heavy nuclides and the key parameters of the fuel cycle are calculated.

IAEA activities in the area of partitioning and transmutation

NASA Astrophysics Data System (ADS)

Stanculescu, Alexander

2006-06-01

Four major challenges are facing the long-term development of nuclear energy: improvement of the economic competitiveness, meeting increasingly stringent safety requirements, adhering to the criteria of sustainable development, and public acceptance. Meeting the sustainability criteria is the driving force behind the topic of this paper. In this context, sustainability has two aspects: natural resources and waste management. IAEA's activities in the area of Partitioning and Transmutation (P&T) are mostly in response to the latter. While not involving the large quantities of gaseous products and toxic solid wastes associated with fossil fuels, radioactive waste disposal is today's dominant public acceptance issue. In fact, small waste quantities permit a rigorous confinement strategy, and mined geological disposal is the strategy followed by some countries. Nevertheless, political opposition arguing that this does not yet constitute a safe disposal technology has largely stalled these efforts. One of the primary reasons cited is the long life of many of the radioisotopes generated from fission. This concern has led to increased R&D efforts to develop a technology aimed at reducing the amount and radio-toxicity of long-lived radioactive waste through transmutation in fission reactors or sub-critical systems. In the frame of the Project on Technology Advances in Fast Reactors and Accelerator-Driven Systems (ADS), the IAEA initiated a number of activities on utilization of plutonium and transmutation of long-lived radioactive waste, ADS, and deuterium-tritium plasma-driven sub-critical systems. The paper presents past accomplishments, current status and planned activities of this IAEA project.

Modelling the behaviour of oxide fuels containing minor actinides with urania, thoria and zirconia matrices in an accelerator-driven system

NASA Astrophysics Data System (ADS)

Sobolev, V.; Lemehov, S.; Messaoudi, N.; Van Uffelen, P.; Aı̈t Abderrahim, H.

2003-06-01

The Belgian Nuclear Research Centre, SCK • CEN, is currently working on the pre-design of the multipurpose accelerator-driven system (ADS) MYRRHA. A demonstration of the possibility of transmutation of minor actinides and long-lived fission products with a realistic design of experimental fuel targets and prognosis of their behaviour under typical ADS conditions is an important task in the MYRRHA project. In the present article, the irradiation behaviour of three different oxide fuel mixtures, containing americium and plutonium - (Am,Pu,U)O 2- x with urania matrix, (Am,Pu,Th)O 2- x with thoria matrix and (Am,Y,Pu,Zr)O 2- x with inert zirconia matrix stabilised by yttria - were simulated with the new fuel performance code MACROS, which is under development and testing at the SCK • CEN. All the fuel rods were considered to be of the same design and sizes: annular fuel pellets, helium bounded with the stainless steel cladding, and a large gas plenum. The liquid lead-bismuth eutectic was used as coolant. Typical irradiation conditions of the hottest fuel assembly of the MYRRHA subcritical core were pre-calculated with the MCNPX code and used in the following calculations as the input data. The results of prediction of the thermo-mechanical behaviour of the designed rods with the considered fuels during three irradiation cycles of 90 EFPD are presented and discussed.

Hybrid systems for transuranic waste transmutation in nuclear power reactors: state of the art and future prospects

NASA Astrophysics Data System (ADS)

Yurov, D. V.; Prikhod'ko, V. V.

2014-11-01

The features of subcritical hybrid systems (HSs) are discussed in the context of burning up transuranic wastes from the U-Pu nuclear fuel cycle. The advantages of HSs over conventional atomic reactors are considered, and fuel cycle closure alternatives using HSs and fast neutron reactors are comparatively evaluated. The advantages and disadvantages of two HS types with neutron sources (NSs) of widely different natures -- nuclear spallation in a heavy target by protons and nuclear fusion in magnetically confined plasma -- are discussed in detail. The strengths and weaknesses of HSs are examined, and demand for them for closing the U-Pu nuclear fuel cycle is assessed.

Isolation of high purity americium metal via distillation

NASA Astrophysics Data System (ADS)

Squires, Leah N.; King, James A.; Fielding, Randall S.; Lessing, Paul

2018-03-01

Pure americium metal is a crucial component for the fabrication of transmutation fuels. Unfortunately, americium in pure metal form is not available; however, a number of mixed metals and mixed oxides that include americium are available. In this manuscript a method is described to obtain high purity americium metal from a mixture of americium and neptunium metals with lead impurity via distillation.

Isotopic Transmutations in Irradiated Beryllium and Their Implications on MARIA Reactor Operation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Andrzejewski, Krzysztof J.; Kulikowska, Teresa A

2004-04-15

Beryllium irradiated by neutrons with energies above 0.7 MeV undergoes (n,{alpha}) and (n,2n) reactions. The Be(n,{alpha}) reaction results in subsequent buildup of {sup 6}Li and {sup 3}He isotopes with large thermal neutron absorption cross sections causing poisoning of irradiated beryllium. The amount of the poison isotopes depends on the neutron flux level and spectrum. The high-flux MARIA reactor operated in Poland since 1975 consists of a beryllium matrix with fuel channels in cutouts of beryllium blocks. As the experimental determination of {sup 6}Li, {sup 3}H, and {sup 3}He content in the operational reactor is impossible, a systematic computational study ofmore » the effect of {sup 3}He and {sup 6}Li presence in beryllium blocks on MARIA reactor reactivity and power density distribution has been undertaken. The analysis of equations governing the transmutation has been done for neutron flux parameters typical for MARIA beryllium blocks. Study of the mutual influence of reactor operational parameters and the buildup of {sup 6}Li, {sup 3}H, and {sup 3}He in beryllium blocks has shown the necessity of a detailed spatial solution of transmutation equations in the reactor, taking into account the whole history of its operation. Therefore, fuel management calculations using the REBUS code with included chains for Be(n,{alpha})-initiated reactions have been done for the whole reactor lifetime. The calculated poisoning of beryllium blocks has been verified against the critical experiment of 1993. Finally, the current {sup 6}Li, {sup 3}H, and {sup 3}He contents, averaged for each beryllium block, have been calculated. The reactivity drop caused by this poisoning is {approx}7%.« less

Application of activation methods on the Dubna experimental transmutation set-ups.

PubMed

Stoulos, S; Fragopoulou, M; Adloff, J C; Debeauvais, M; Brandt, R; Westmeier, W; Krivopustov, M; Sosnin, A; Papastefanou, C; Zamani, M; Manolopoulou, M

2003-02-01

High spallation neutron fluxes were produced by irradiating massive heavy targets with proton beams in the GeV range. The experiments were performed at the Dubna High Energy Laboratory using the nuclotron accelerator. Two different experimental set-ups were used to produce neutron spectra convenient for transmutation of radioactive waste by (n,x) reactions. By a theoretical analysis neutron spectra can be reproduced from activation measurements. Thermal-epithermal and fast-super-fast neutron fluxes were estimated using the 197Au, 238U (n,gamma) and (n,2n) reactions, respectively. Depleted uranium transmutation rates were also studied in both experiments.

On the use of a molten salt fast reactor to apply an idealized transmutation scenario for the nuclear phase out.

PubMed

Merk, Bruno; Rohde, Ulrich; Glivici-Cotruţă, Varvara; Litskevich, Dzianis; Scholl, Susanne

2014-01-01

In the view of transmutation of transuranium (TRU) elements, molten salt fast reactors (MSFRs) offer certain advantages compared to solid fuelled reactor types like sodium cooled fast reactors (SFRs). In the first part these advantages are discussed in comparison with the SFR technology, and the research challenges are analyzed. In the second part cycle studies for the MSFR are given for different configurations--a core with U-238 fertile, a fertile free core, and a core with Th-232 as fertile material. For all cases, the transmutation potential is determined and efficient transmutation performance for the case with thorium as a fertile material as well as for the fertile free case is demonstrated and the individual advantages are discussed. The time evolution of different important isotopes is analyzed. In the third part a strategy for the optimization of the transmutation efficiency is developed. The final aim is dictated by the phase out decision of the German government, which requests to put the focus on the determination of the maximal transmutation efficiency and on an as much as possible reduced leftover of transuranium elements at the end of the reactor life. This minimal leftover is achieved by a two step procedure of a first transmuter operation phase followed by a second deep burning phase. There the U-233, which is bred in the blanket of the core consisting of thorium containing salt, is used as feed. It is demonstrated, that transmutation rates up to more than 90% can be achieved for all transuranium isotopes, while the production of undesired high elements like californium is very limited. Additionally, the adaptations needed for the simulation of a MSFR, and the used tool HELIOS 1.10 is described.

On the Use of a Molten Salt Fast Reactor to Apply an Idealized Transmutation Scenario for the Nuclear Phase Out

PubMed Central

Merk, Bruno; Rohde, Ulrich; Glivici-Cotruţă, Varvara; Litskevich, Dzianis; Scholl, Susanne

2014-01-01

In the view of transmutation of transuranium (TRU) elements, molten salt fast reactors (MSFRs) offer certain advantages compared to solid fuelled reactor types like sodium cooled fast reactors (SFRs). In the first part these advantages are discussed in comparison with the SFR technology, and the research challenges are analyzed. In the second part cycle studies for the MSFR are given for different configurations – a core with U-238 fertile, a fertile free core, and a core with Th-232 as fertile material. For all cases, the transmutation potential is determined and efficient transmutation performance for the case with thorium as a fertile material as well as for the fertile free case is demonstrated and the individual advantages are discussed. The time evolution of different important isotopes is analyzed. In the third part a strategy for the optimization of the transmutation efficiency is developed. The final aim is dictated by the phase out decision of the German government, which requests to put the focus on the determination of the maximal transmutation efficiency and on an as much as possible reduced leftover of transuranium elements at the end of the reactor life. This minimal leftover is achieved by a two step procedure of a first transmuter operation phase followed by a second deep burning phase. There the U-233, which is bred in the blanket of the core consisting of thorium containing salt, is used as feed. It is demonstrated, that transmutation rates up to more than 90% can be achieved for all transuranium isotopes, while the production of undesired high elements like californium is very limited. Additionally, the adaptations needed for the simulation of a MSFR, and the used tool HELIOS 1.10 is described. PMID:24690768

Fusion breeder

DOE Office of Scientific and Technical Information (OSTI.GOV)

Moir, R.W.

1982-02-22

The fusion breeder is a fusion reactor designed with special blankets to maximize the transmutation by 14 MeV neutrons of uranium-238 to plutonium or thorium to uranium-233 for use as a fuel for fission reactors. Breeding fissile fuels has not been a goal of the US fusion energy program. This paper suggests it is time for a policy change to make the fusion breeder a goal of the US fusion program and the US nuclear energy program. The purpose of this paper is to suggest this policy change be made and tell why it should be made, and to outlinemore » specific research and development goals so that the fusion breeder will be developed in time to meet fissile fuel needs.« less

Fusion breeder

DOE Office of Scientific and Technical Information (OSTI.GOV)

Moir, R.W.

1982-04-20

The fusion breeder is a fusion reactor designed with special blankets to maximize the transmutation by 14 MeV neutrons of uranium-238 to plutonium or thorium to uranium-233 for use as a fuel for fission reactors. Breeding fissile fuels has not been a goal of the US fusion energy program. This paper suggests it is time for a policy change to make the fusion breeder a goal of the US fusion program and the US nuclear energy program. The purpose of this paper is to suggest this policy change be made and tell why it should be made, and to outlinemore » specific research and development goals so that the fusion breeder will be developed in time to meet fissile fuel needs.« less

Integrated process modeling for the laser inertial fusion energy (LIFE) generation system

NASA Astrophysics Data System (ADS)

Meier, W. R.; Anklam, T. M.; Erlandson, A. C.; Miles, R. R.; Simon, A. J.; Sawicki, R.; Storm, E.

2010-08-01

A concept for a new fusion-fission hybrid technology is being developed at Lawrence Livermore National Laboratory. The primary application of this technology is base-load electrical power generation. However, variants of the baseline technology can be used to "burn" spent nuclear fuel from light water reactors or to perform selective transmutation of problematic fission products. The use of a fusion driver allows very high burn-up of the fission fuel, limited only by the radiation resistance of the fuel form and system structures. As a part of this process, integrated process models have been developed to aid in concept definition. Several models have been developed. A cost scaling model allows quick assessment of design changes or technology improvements on cost of electricity. System design models are being used to better understand system interactions and to do design trade-off and optimization studies. Here we describe the different systems models and present systems analysis results. Different market entry strategies are discussed along with potential benefits to US energy security and nuclear waste disposal. Advanced technology options are evaluated and potential benefits from additional R&D targeted at the different options is quantified.

Safety features of subcritical fluid fueled systems

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bell, C.R.

1995-10-01

Accelerator-driven transmutation technology has been under study at Los Alamos for several years for application to nuclear waste treatment, tritium production, energy generation, and recently, to the disposition of excess weapons plutonium. Studies and evaluations performed to date at Los Alamos have led to a current focus on a fluid-fuel, fission system operating in a neutron source-supported subcritical mode, using molten salt reactor technology and accelerator-driven proton-neutron spallation. In this paper, the safety features and characteristics of such systems are explored from the perspective of the fundamental nuclear safety objectives that any reactor-type system should address. This exploration is qualitativemore » in nature and uses current vintage solid-fueled reactors as a baseline for comparison. Based on the safety perspectives presented, such systems should be capable of meeting the fundamental nuclear safety objectives. In addition, they should be able to provide the safety robustness desired for advanced reactors. However, the manner in which safety objectives and robustness are achieved is very different from that associated with conventional reactors. Also, there are a number of safety design and operational challenges that will have to be addressed for the safety potential of such systems to be credible.« less

Dual neutral particle induced transmutation in CINDER2008

NASA Astrophysics Data System (ADS)

Martin, W. J.; de Oliveira, C. R. E.; Hecht, A. A.

2014-12-01

Although nuclear transmutation methods for fission have existed for decades, the focus has been on neutron-induced reactions. Recent novel concepts have sought to use both neutrons and photons for purposes such as active interrogation of cargo to detect the smuggling of highly enriched uranium, a concept that would require modeling the transmutation caused by both incident particles. As photonuclear transmutation has yet to be modeled alongside neutron-induced transmutation in a production code, new methods need to be developed. The CINDER2008 nuclear transmutation code from Los Alamos National Laboratory is extended from neutron applications to dual neutral particle applications, allowing both neutron- and photon-induced reactions for this modeling with a focus on fission. Following standard reaction modeling, the induced fission reaction is understood as a two-part reaction, with an entrance channel to the excited compound nucleus, and an exit channel from the excited compound nucleus to the fission fragmentation. Because photofission yield data-the exit channel from the compound nucleus-are sparse, neutron fission yield data are used in this work. With a different compound nucleus and excitation, the translation to the excited compound state is modified, as appropriate. A verification and validation of these methods and data has been performed. This has shown that the translation of neutron-induced fission product yield sets, and their use in photonuclear applications, is appropriate, and that the code has been extended correctly.

Insight into the Am-O Phase Equilibria: A Thermodynamic Study Coupling High-Temperature XRD and CALPHAD Modeling.

PubMed

Epifano, Enrica; Guéneau, Christine; Belin, Renaud C; Vauchy, Romain; Lebreton, Florent; Richaud, Jean-Christophe; Joly, Alexis; Valot, Christophe; Martin, Philippe M

2017-07-03

In the frame of minor actinide transmutation, americium can be diluted in UO 2 and (U, Pu)O 2 fuels burned in fast neutron reactors. The first mandatory step to foresee the influence of Am on the in-reactor behavior of transmutation targets or fuel is to have fundamental knowledge of the Am-O binary system and, in particular, of the AmO 2-x phase. In this study, we coupled HT-XRD (high-temperature X-ray diffraction) experiments with CALPHAD thermodynamic modeling to provide new insights into the structural properties and phase equilibria in the AmO 2-x -AmO 1.61+x -Am 2 O 3 domain. Because of this approach, we were able for the first time to assess the relationships between temperature, lattice parameter, and hypostoichiometry for fcc AmO 2-x . We showed the presence of a hyperstoichiometric existence domain for the bcc AmO 1.61+x phase and the absence of a miscibility gap in the fcc AmO 2-x phase, contrary to previous representations of the phase diagram. Finally, with the new experimental data, a new CALPHAD thermodynamic model of the Am-O system was developed, and an improved version of the phase diagram is presented.

Use of freeze-casting in advanced burner reactor fuel design

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lang, A. L.; Yablinsky, C. A.; Allen, T. R.

2012-07-01

This paper will detail the modeling of a fast reactor with fuel pins created using a freeze-casting process. Freeze-casting is a method of creating an inert scaffold within a fuel pin. The scaffold is created using a directional solidification process and results in open porosity for emplacement of fuel, with pores ranging in size from 300 microns to 500 microns in diameter. These pores allow multiple fuel types and enrichments to be loaded into one fuel pin. Also, each pore could be filled with varying amounts of fuel to allow for the specific volume of fission gases created by thatmore » fuel type. Currently fast reactors, including advanced burner reactors (ABR's), are not economically feasible due to the high cost of operating the reactors and of reprocessing the fuel. However, if the fuel could be very precisely placed, such as within a freeze-cast scaffold, this could increase fuel performance and result in a valid design with a much lower cost per megawatt. In addition to competitive costs, freeze-cast fuel would also allow for selective breeding or burning of actinides within specific locations in fast reactors. For example, fast flux peak locations could be utilized on a minute scale to target specific actinides for transmutation. Freeze-cast fuel is extremely flexible and has great potential in a variety of applications. This paper performs initial modeling of freeze-cast fuel, with the generic fast reactor parameters for this model based on EBR-II. The core has an assumed power of 62.5 MWt. The neutronics code used was Monte Carlo N-Particle (MCNP5) transport code. Uniform pore sizes were used in increments of 100 microns. Two different freeze-cast scaffold materials were used: ceramic (MgO-ZrO{sub 2}) and steel (SS316L). Separate models were needed for each material because the freeze-cast ceramic and metal scaffolds have different structural characteristics and overall porosities. Basic criticality results were compiled for the various models. Preliminary results show that criticality is achievable with freeze-cast fuel pins despite the significant amount of inert fuel matrix. Freeze casting is a promising method to achieve very precise fuel placement within fuel pins. (authors)« less

The role of accelerators in the nuclear fuel cycle

DOE Office of Scientific and Technical Information (OSTI.GOV)

Takahashi, Hiroshi.

1990-01-01

The use of neutrons produced by the medium energy proton accelerator (1 GeV--3 GeV) has considerable potential in reconstructing the nuclear fuel cycle. About 1.5 {approximately} 2.5 ton of fissile material can be produced annually by injecting a 450 MW proton beam directly into fertile materials. A source of neutrons, produced by a proton beam, to supply subcritical reactors could alleviate many of the safety problems associated with critical assemblies, such as positive reactivity coefficients due to coolant voiding. The transient power of the target can be swiftly controlled by controlling the power of the proton beam. Also, the usemore » of a proton beam would allow more flexibility in the choice of fuel and structural materials which otherwise might reduce the reactivity of reactors. This paper discusses the rate of accelerators in the transmutation of radioactive wastes of the nuclear fuel cycles. 34 refs., 17 figs., 9 tabs.« less

Spatial heterogeneity of tungsten transmutation in a fusion device

NASA Astrophysics Data System (ADS)

Gilbert, M. R.; Sublet, J.-Ch.; Dudarev, S. L.

2017-04-01

Accurately quantifying the transmutation rate of tungsten (W) under neutron irradiation is a necessary requirement in the assessment of its performance as an armour material in a fusion power plant. The usual approach of calculating average responses, assuming large, homogenised material volumes, is insufficient to capture the full complexity of the transmutation picture in the context of a realistic fusion power plant design, particularly for rhenium (Re) production from W. Combined neutron transport and inventory simulations for representative spatially heterogeneous high-resolution models of a fusion power plant show that the production rate of Re is strongly influenced by the surrounding local spatial environment. Localised variation in neutron moderation (slowing down) due to structural steel and coolant, particularly water, can dramatically increase Re production because of the huge cross sections of giant resolved resonances in the neutron-capture reaction of 186W at low neutron energies. Calculations using cross section data corrected for temperature (Doppler) effects suggest that temperature may have a relatively lesser influence on transmutation rates.

Integral nuclear data validation using experimental spent nuclear fuel compositions

DOE PAGES

Gauld, Ian C.; Williams, Mark L.; Michel-Sendis, Franco; ...

2017-07-19

Measurements of the isotopic contents of spent nuclear fuel provide experimental data that are a prerequisite for validating computer codes and nuclear data for many spent fuel applications. Under the auspices of the Organisation for Economic Co-operation and Development (OECD) Nuclear Energy Agency (NEA) and guidance of the Expert Group on Assay Data of Spent Nuclear Fuel of the NEA Working Party on Nuclear Criticality Safety, a new database of expanded spent fuel isotopic compositions has been compiled. The database, Spent Fuel Compositions (SFCOMPO) 2.0, includes measured data for more than 750 fuel samples acquired from 44 different reactors andmore » representing eight different reactor technologies. Measurements for more than 90 isotopes are included. This new database provides data essential for establishing the reliability of code systems for inventory predictions, but it also has broader potential application to nuclear data evaluation. Furthermore, the database, together with adjoint based sensitivity and uncertainty tools for transmutation systems developed to quantify the importance of nuclear data on nuclide concentrations, are described.« less

Integral nuclear data validation using experimental spent nuclear fuel compositions

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gauld, Ian C.; Williams, Mark L.; Michel-Sendis, Franco

Measurements of the isotopic contents of spent nuclear fuel provide experimental data that are a prerequisite for validating computer codes and nuclear data for many spent fuel applications. Under the auspices of the Organisation for Economic Co-operation and Development (OECD) Nuclear Energy Agency (NEA) and guidance of the Expert Group on Assay Data of Spent Nuclear Fuel of the NEA Working Party on Nuclear Criticality Safety, a new database of expanded spent fuel isotopic compositions has been compiled. The database, Spent Fuel Compositions (SFCOMPO) 2.0, includes measured data for more than 750 fuel samples acquired from 44 different reactors andmore » representing eight different reactor technologies. Measurements for more than 90 isotopes are included. This new database provides data essential for establishing the reliability of code systems for inventory predictions, but it also has broader potential application to nuclear data evaluation. Furthermore, the database, together with adjoint based sensitivity and uncertainty tools for transmutation systems developed to quantify the importance of nuclear data on nuclide concentrations, are described.« less

Update and evaluation of decay data for spent nuclear fuel analyses

NASA Astrophysics Data System (ADS)

Simeonov, Teodosi; Wemple, Charles

2017-09-01

Studsvik's approach to spent nuclear fuel analyses combines isotopic concentrations and multi-group cross-sections, calculated by the CASMO5 or HELIOS2 lattice transport codes, with core irradiation history data from the SIMULATE5 reactor core simulator and tabulated isotopic decay data. These data sources are used and processed by the code SNF to predict spent nuclear fuel characteristics. Recent advances in the generation procedure for the SNF decay data are presented. The SNF decay data includes basic data, such as decay constants, atomic masses and nuclide transmutation chains; radiation emission spectra for photons from radioactive decay, alpha-n reactions, bremsstrahlung, and spontaneous fission, electrons and alpha particles from radioactive decay, and neutrons from radioactive decay, spontaneous fission, and alpha-n reactions; decay heat production; and electro-atomic interaction data for bremsstrahlung production. These data are compiled from fundamental (ENDF, ENSDF, TENDL) and processed (ESTAR) sources for nearly 3700 nuclides. A rigorous evaluation procedure of internal consistency checks and comparisons to measurements and benchmarks, and code-to-code verifications is performed at the individual isotope level and using integral characteristics on a fuel assembly level (e.g., decay heat, radioactivity, neutron and gamma sources). Significant challenges are presented by the scope and complexity of the data processing, a dearth of relevant detailed measurements, and reliance on theoretical models for some data.

Radiogenic lead as coolant, reflector and moderator in advanced fast reactors

NASA Astrophysics Data System (ADS)

Kulikov, E. G.

2017-01-01

Main purpose of the study is assessing reasonability for recovery, production and application of radiogenic lead as a coolant, neutron moderator and neutron reflector in advanced fast reactors. When performing the study, thermal, physical and neutron-physical properties of natural and radiogenic lead were analyzed. The following results were obtained: 1. Radiogenic lead with high content of isotope 208Pb can be extracted from thorium or mixed thorium-uranium ores because 208Pb is a final product of 232Th natural decay chain. 2. The use of radiogenic lead with high 208Pb content in advanced fast reactors and accelerator-driven systems (ADS) makes it possible to improve significantly their neutron-physical and thermal-hydraulic parameters. 3. The use of radiogenic lead with high 208Pb content in advanced fast reactors as a coolant opens the possibilities for more intense fuel breeding and for application of well-known oxide fuel instead of the promising but not tested enough nitride fuel under the same safety parameters. 4. The use of radiogenic lead with high 208Pb content in ADS as a coolant can upgrade substantially the level of neutron flux in the ADS blanket, which enables effective transmutation of radioactive wastes with low cross-sections of radiative neutron capture.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Yim, J. S.; Tahk, Y. W.; Oh, J. Y.

In order to cope with global shortage of Mo-99 supplies and with growing demand of neutron transmutation doping, KJRR construction plan has been launched since April 2012 to provide self-sufficiency of domestic RI demand, and to extend Si doping capacity for power device market growth. Through comprehensive surveillance of the fuels in-reactor behavior, KAERI has selected the fuel meat of U-7%Mo dispersion in an aluminum matrix with 5wt%Si for the KJRR fuel. As part of the efforts for fuel licensing and qualification of the KJRR fuel, an LTA irradiation test at the ATR started from November 2015 was successfully completedmore » by reaching at 219 EFPD in the end of February 2017. Together with the results of HAMP-1 already completed irradiation and PIE, the successful irradiation of the LTA also demonstrates the fuel integrity under more rigorous conditions than the KJRR operation conditions. This paper updates the current status of the KJRR U7Mo (8 g-U/cm3) LTA irradiation and PIE plan up to date as of February 2017.« less

System analyses on advanced nuclear fuel cycle and waste management

NASA Astrophysics Data System (ADS)

Cheon, Myeongguk

To evaluate the impacts of accelerator-driven transmutation of waste (ATW) fuel cycle on a geological repository, two mathematical models are developed: a reactor system analysis model and a high-level waste (HLW) conditioning model. With the former, fission products and residual trans-uranium (TRU) contained in HLW generated from a reference ATW plant operations are quantified and the reduction of TRU inventory included in commercial spent-nuclear fuel (CSNF) is evaluated. With the latter, an optimized waste loading and composition in solidification of HLW are determined and the volume reduction of waste packages associated with CSNF is evaluated. WACOM, a reactor system analysis code developed in this study for burnup calculation, is validated by ORIGEN2.1 and MCNP. WACOM is used to perform multicycle analysis for the reference lead-bismuth eutectic (LBE) cooled transmuter. By applying the results of this analysis to the reference ATW deployment scenario considered in the ATW roadmap, the HLW generated from the ATW fuel cycle is quantified and the reduction of TRU inventory contained in CSNF is evaluated. A linear programming (LP) model has been developed for determination of an optimized waste loading and composition in solidification of HLW. The model has been applied to a US-defense HLW. The optimum waste loading evaluated by the LP model was compared with that estimated by the Defense Waste Processing Facility (DWPF) in the US and a good agreement was observed. The LP model was then applied to the volume reduction of waste packages associated with CSNF. Based on the obtained reduction factors, the expansion of Yucca Mountain Repository (YMR) capacity is evaluated. It is found that with the reference ATW system, the TRU contained in CSNF could be reduced by a factor of ˜170 in terms of inventory and by a factor of ˜40 in terms of toxicity under the assumed scenario. The number of waste packages related to CSNF could be reduced by a factor of ˜8 in terms of volume and by factor of ˜10 on the basis of electricity generation when a sufficient cooling time for discharged spent fuel and zero process chemicals in HLW are assumed. The expansion factor of Yucca Mountain Repository capacity is estimated to be a factor of 2.4, much smaller than the reduction factor of CSNF waste packages, due to the existence of DOE-owned spent fuel and HLW. The YMR, however, could support 10 times greater electricity generation as long as the statutory capacity of DOE-owned SNF and HLW remains unchanged. This study also showed that the reduction of the number of waste packages could strongly be subject to the heat generation rate of HLW and the amount of process chemicals contained in HLW. For a greater reduction of the number of waste packages, a sufficient cooling time for discharged fuel and efforts to minimize the amount of process chemicals contained in HLW are crucial.

Research on stellarator-mirror fission-fusion hybrid

NASA Astrophysics Data System (ADS)

Moiseenko, V. E.; Kotenko, V. G.; Chernitskiy, S. V.; Nemov, V. V.; Ågren, O.; Noack, K.; Kalyuzhnyi, V. N.; Hagnestål, A.; Källne, J.; Voitsenya, V. S.; Garkusha, I. E.

2014-09-01

The development of a stellarator-mirror fission-fusion hybrid concept is reviewed. The hybrid comprises of a fusion neutron source and a powerful sub-critical fast fission reactor core. The aim is the transmutation of spent nuclear fuel and safe fission energy production. In its fusion part, neutrons are generated in deuterium-tritium (D-T) plasma, confined magnetically in a stellarator-type system with an embedded magnetic mirror. Based on kinetic calculations, the energy balance for such a system is analyzed. Neutron calculations have been performed with the MCNPX code, and the principal design of the reactor part is developed. Neutron outflux at different outer parts of the reactor is calculated. Numerical simulations have been performed on the structure of a magnetic field in a model of the stellarator-mirror device, and that is achieved by switching off one or two coils of toroidal field in the Uragan-2M torsatron. The calculations predict the existence of closed magnetic surfaces under certain conditions. The confinement of fast particles in such a magnetic trap is analyzed.

New infrastructure for studies of transmutation and fast systems concepts

NASA Astrophysics Data System (ADS)

Panza, Fabio; Firpo, Gabriele; Lomonaco, Guglielmo; Osipenko, Mikhail; Ricco, Giovanni; Ripani, Marco; Saracco, Paolo; Viberti, Carlo Maria

2017-09-01

In this work we report initial studies on a low power Accelerator-Driven System as a possible experimental facility for the measurement of relevant integral nuclear quantities. In particular, we performed Monte Carlo simulations of minor actinides and fission products irradiation and estimated the fission rate within fission chambers in the reactor core and the reflector, in order to evaluate the transmutation rates and the measurement sensitivity. We also performed a photo-peak analysis of available experimental data from a research reactor, in order to estimate the expected sensitivity of this analysis method on the irradiation of samples in the ADS considered.

A low power ADS for transmutation studies in fast systems

NASA Astrophysics Data System (ADS)

Panza, Fabio; Firpo, Gabriele; Lomonaco, Guglielmo; Osipenko, Mikhail; Ricco, Giovanni; Ripani, Marco; Saracco, Paolo; Viberti, Carlo Maria

2017-12-01

In this work, we report studies on a fast low power accelerator driven system model as a possible experimental facility, focusing on its capabilities in terms of measurement of relevant integral nuclear quantities. In particular, we performed Monte Carlo simulations of minor actinides and fission products irradiation and estimated the fission rate within fission chambers in the reactor core and the reflector, in order to evaluate the transmutation rates and the measurement sensitivity. We also performed a photo-peak analysis of available experimental data from a research reactor, in order to estimate the expected sensitivity of this analysis method on the irradiation of samples in the ADS considered.

Design and Analysis of Thorium-fueled Reduced Moderation Boiling Water Reactors

NASA Astrophysics Data System (ADS)

Gorman, Phillip Michael

The Resource-renewable Boiling Water Reactors (RBWRs) are a set of light water reactors (LWRs) proposed by Hitachi which use a triangular lattice and high void fraction to incinerate fuel with an epithermal spectrum, which is highly atypical of LWRs. The RBWRs operate on a closed fuel cycle, which is impossible with a typical thermal spectrum reactor, in order to accomplish missions normally reserved for sodium fast reactors (SFRs)--either fuel self-sufficiency or waste incineration. The RBWRs also axially segregate the fuel into alternating fissile "seed" regions and fertile "blanket" regions in order to enhance breeding and leakage probability upon coolant voiding. This dissertation focuses on thorium design variants of the RBWR: the self-sufficient RBWR-SS and the RBWR-TR, which consumes reprocessed transuranic (TRU) waste from PWR used nuclear fuel. These designs were based off of the Hitachi-designed RBWR-AC and the RBWR-TB2, respectively, which use depleted uranium (DU) as the primary fertile fuel. The DU-fueled RBWRs use a pair of axially segregated seed sections in order to achieve a negative void coefficient; however, several concerns were raised with this multi-seed approach, including difficulty with controlling the reactor and unacceptably high axial power peaking. Since thorium-uranium fuel tends to have much more negative void feedback than uranium-plutonium fuels, the thorium RBWRs were designed to use a single elongated seed to avoid these issues. A series of parametric studies were performed in order to find the design space for the thorium RBWRs, and optimize the designs while meeting the required safety constraints. The RBWR-SS was optimized to maximize the discharge burnup, while the RBWR-TR was optimized to maximize the TRU transmutation rate. These parametric studies were performed on an assembly level model using the MocDown simulator, which calculates an equilibrium fuel composition with a specified reprocessing scheme. A full core model was then created for each design, using the Serpent/PARCS 3-D core simulator, and the full core performance was assessed. The RBWR-SS benefited from a harder spectrum than the RBWR-TR; a hard spectrum promotes breeding and increases the discharge burnup, but reduces the TRU transmutation rate. This led the RBWR-SS to have a very tight lattice, which has a lot of experimental uncertainty in the thermal hydraulic correlations. Two different RBWR-SS designs were created assuming different thermal hydraulic assumptions: the RBWR-SSH used the same assumptions as Hitachi used for the RBWR-AC, while the RBWR-SSM used more conservative correlations recommended by collaborators at MIT. However, the void feedback of the pure Th-fed system was too strongly negative, even with a single elongated seed. Therefore, instead of using just thorium, the self-sustaining designs were fed with a mix of between 30% and 50% DU and the rest thorium in order to keep the void feedback as close to zero as possible. This was not necessary for the RBWR-TR, as the external TRU feed fulfilled a similar role. Unfortunately, it was found that the RBWR-SSM could not sustain a critical cycle without either significantly downgrading the power or supplying an external feed of fissile material. While the RBWR-SSH and the RBWR-TR could reach similar burnups and transmutation rates to their DU-fueled counterparts as designed by Hitachi, the thorium designs were unable to simultaneously have negative void feedback and sufficient shutdown margin to shut down the core. The multi-seed approach of the Hitachi designs allowed their reactors to have much lower magnitudes of Doppler feedback than the single-seed designs, which helps them to have sufficient shutdown margin. It is expected that thorium-fueled RBWRs designed to have multiple seeds would permit adequate shutdown margin, although care would need to be taken in order to avoid running into the same issues as the DU fueled RBWRs. Alternatively, it may be possible to increase the amount of boron in the control blades by changing the assembly and core design. Nonetheless, the uncertainties in the multiplication factor due to nuclear data and void fraction uncertainty were assessed for the RBWR-SSH and the RBWR-TR, as well as for the RBWR-TB2. In addition, the uncertainty associated with the change in reactor states (such as the reactivity insertion in flooding the core) due to nuclear data uncertainties was quantified. The thorium RBWRs have much larger uncertainty of their DU-fueled counterparts as designed by Hitachi, as the fission cross section of 233U has very large uncertainty in the epithermal energy range. The uncertainty in the multiplication factor at reference conditions was about 1350 pcm for the RBWR-SSH, while it was about 900 pcm for the RBWR-TR. The uncertainty in the void coefficient of reactivity for both reactors is between 8 and 10 pcm/% void, which is on the same order of magnitude as the full core value. Finally, since sharp linear heat rate spikes were observed in the RBWR-TB2 simulation, the RBWR-TB2 unit cell was simulated using a much finer mesh than is possible using deterministic codes. It was found that the thermal neutrons reflecting back from the reflectors and the blankets were causing extreme spikes in the power density near the axial boundaries of the seeds, which were artificially smoothed out when using coarser meshes. It is anticipated that these spikes will cause melting in both seeds in the RBWR-TB2, unless design changes--such as reducing the enrichment level near the axial boundaries of the seeds--are made.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Maloy, Stuart Andrew; Pestovich, Kimberly Shay; Anderoglu, Osman

The Fuel Cycle Research and Development program is investigating methods of transmuting minor actinides in various fuel cycle options. To achieve this goal, new fuels and cladding materials must be developed and tested to high burnup levels (e.g. >20%) requiring cladding to withstand very high doses (greater than 200 dpa) while in contact with the coolant and the fuel. To develop and qualify materials to a total fluence greater than 200 dpa requires development of advanced alloys and irradiations in fast reactors to test these alloys. Recent results from testing numerous ferritic/martensitic steels at low temperatures suggest that improvements inmore » low temperature radiation tolerance can be achieved through carefully controlling the nitrogen content in these alloys. Thus, four new heats of HT-9 were produced with controlled nitrogen content: two by Metalwerks and two by Sophisticated Alloys. Initial results on these new alloys are presented including microstructural analysis and hardness testing. Future testing will include irradiation testing with ions and in reactor.« less

Neutron Transmutation Doped (NTD) germanium thermistors for sub-mm bolometer applications

NASA Technical Reports Server (NTRS)

Haller, E. E.; Itoh, K. M.; Beeman, J. W.

1996-01-01

Recent advances in the development of neutron transmutation doped (NTD) semiconductor thermistors fabricated from natural and controlled isotopic composition germanium are reported. The near ideal doping uniformity that can be achieved with the NTD process, the device simplicity of NTD Ge thermistors and the high performance of cooled junction field effect transistor preamplifiers led to the widespread acceptance of these thermal sensors in ground-based, airborne and spaceborne radio telescopes. These features made possible the development of efficient bolometer arrays.

On the influence of the americium isotopic vector on the cooling time of minor actinides bearing blankets in fast reactors

NASA Astrophysics Data System (ADS)

Kooyman, Timothée; Buiron, Laurent; Rimpault, Gerald

2018-05-01

In the heterogeneous minor actinides transmutation approach, the nuclei to be transmuted are loaded in dedicated targets often located at the core periphery, so that long-lived heavy nuclides are turned into shorter-lived fission products by fission. To compensate for low flux level at the core periphery, the minor actinides content in the targets is set relatively high (around 20 at.%), which has a negative impact on the reprocessing of the targets due to their important decay heat level. After a complete analysis of the main contributors to the heat load of the irradiated targets, it is shown here that the choice of the reprocessing order of the various feeds of americium from the fuel cycle depends on the actual limit for fuel reprocessing. If reprocessing of hot targets is possible, it is more interesting to reprocess first the americium feed with a high 243Am content in order to limit the total cooling time of the targets, while if reprocessing of targets is limited by their decay heat, it is more interesting to wait for an increase in the 241Am content before loading the americium in the core. An optimization of the reprocessing order appears to lead to a decrease of the total cooling time by 15 years compared to a situation where all the americium feeds are mixed together when two feeds from SFR are considered with a high reprocessing limit.

Radiolytic degradation of a new diglycol-diamide ligand for actinide and lanthanide co-extraction from spent nuclear fuel

NASA Astrophysics Data System (ADS)

Ossola, Annalisa; Macerata, Elena; Tinonin, Dario A.; Faroldi, Federica; Giola, Marco; Mariani, Mario; Casnati, Alessandro

2016-07-01

Within the Partitioning and Transmutation strategies, great efforts have been devoted in the last decades to the development of lipophilic ligands able to co-extract trivalent Lanthanides (Ln) and Actinides (An) from spent nuclear fuel. Because of the harsh working conditions these ligands undergo, it is important to prove their chemical and radiolytic stability during the counter-current multi-stage extraction process. In the present work the hydrolytic and radiolytic resistance of the freshly prepared and aged organic solutions containing the new ligand (2,6-bis[(N-methyl-N-dodecyl)carboxamide]-4-methoxy-tetrahydro-pyran) were investigated in order to evaluate the impact on the safety and efficiency of the process. Liquid-liquid extraction tests with spiked solutions showed that the ligand extracting performances are strongly impaired by storing the samples at room temperature and in the light. Moreover, the extracting efficiency of the irradiated samples resulted to be influenced by gamma irradiation, while selectivity remains unchanged. Preliminary mass spectrometric data showed that degradation is mainly due to the acid-catalysed reaction of the ligand carboxamide and ether groups with the 1-octanol present in the diluent.

LWR First Recycle of TRU with Thorium Oxide for Transmutation and Cross Sections

DOE Office of Scientific and Technical Information (OSTI.GOV)

Andrea Alfonsi; Gilles Youinou; Sonat Sen

2013-02-01

Thorium has been considered as an option to uranium-based fuel, based on considerations of resource utilization (thorium is approximately three times more plentiful than uranium) and as a result of concerns about proliferation and waste management (e.g. reduced production of plutonium, etc.). Since the average composition of natural Thorium is dominated (100%) by the fertile isotope Th-232, Thorium is only useful as a resource for breeding new fissile materials, in this case U-233. Consequently a certain amount of fissile material must be present at the start-up of the reactor in order to guarantee its operation. The thorium fuel can bemore » used in both once-through and recycle options, and in both fast and thermal spectrum systems. The present study has been aimed by the necessity of investigating the option of using reprocessed plutonium/TRU, from a once-through reference LEU scenario (50 GWd/ tIHM), mixed with natural thorium and the need of collect data (mass fractions, cross-sections etc.) for this particular fuel cycle scenario. As previously pointed out, the fissile plutonium is needed to guarantee the operation of the reactor. Four different scenarios have been considered: • Thorium – recycled Plutonium; • Thorium – recycled Plutonium/Neptunium; • Thorium – recycled Plutonium/Neptunium/Americium; • Thorium – recycled Transuranic. The calculations have been performed with SCALE6.1-TRITON.« less

LWR First Recycle of TRU with Thorium Oxide for Transmutation and Cross Sections

DOE Office of Scientific and Technical Information (OSTI.GOV)

Andrea Alfonsi; Gilles Youinou

2012-07-01

Thorium has been considered as an option to uranium-based fuel, based on considerations of resource utilization (thorium is approximately three times more plentiful than uranium) and as a result of concerns about proliferation and waste management (e.g. reduced production of plutonium, etc.). Since the average composition of natural Thorium is dominated (100%) by the fertile isotope Th-232, Thorium is only useful as a resource for breeding new fissile materials, in this case U-233. Consequently a certain amount of fissile material must be present at the start-up of the reactor in order to guarantee its operation. The thorium fuel can bemore » used in both once-through and recycle options, and in both fast and thermal spectrum systems. The present study has been aimed by the necessity of investigating the option of using reprocessed plutonium/TRU, from a once-through reference LEU scenario (50 GWd/ tIHM), mixed with natural thorium and the need of collect data (mass fractions, cross-sections etc.) for this particular fuel cycle scenario. As previously pointed out, the fissile plutonium is needed to guarantee the operation of the reactor. Four different scenarios have been considered: • Thorium – recycled Plutonium; • Thorium – recycled Plutonium/Neptunium; • Thorium – recycled Plutonium/Neptunium/Americium; • Thorium – recycled Transuranic. The calculations have been performed with SCALE6.1-TRITON.« less

Radiological assessment of target materials for accelerator transmutation of waste (ATW) applications

NASA Astrophysics Data System (ADS)

Vickers, Linda Diane

This dissertation issues the first published document of the radiation absorbed dose rate (rad-h-1) to tissue from radioactive spallation products in Ta, W, Pb, Bi, and LBE target materials used in Accelerator Transmutation of Waste (ATW) applications. No previous works have provided an estimate of the absorbed dose rate (rad-h-1) from activated targets for ATW applications. The results of this dissertation are useful for planning the radiological safety assessment to personnel, and for the design, construction, maintenance, and disposition of target materials of high-energy particle accelerators for ATW applications (Charlton, 1996). In addition, this dissertation provides the characterization of target materials of high-energy particle accelerators for the parameters of: (1) spallation neutron yield (neutrons/proton), (2) spallation products yield (nuclides/proton), (3) energy-dependent spallation neutron fluence distribution, (4) spallation neutron flux, (5) identification of radioactive spallation products for consideration in safety of personnel to high radiation dose rates, and (6) identification of the optimum geometrical dimensions for the target applicable to the maximum radial spallation neutron leakage from the target. Pb and Bi target materials yielded the lowest absorbed dose rates (rad-h -1) for a 10-year irradiation/50-year decay scheme, and would be the preferred target materials for consideration of the radiological safety of personnel during ATW operations. A beneficial characteristic of these target materials is that they do not produce radioactive transuranic isotopes, which have very long half-lives and require special handling and disposition requirements. Furthermore, the targets are not considered High-Level Waste (HLW) such as reactor spent fuel for disposal purposes. It is a basic ATW system requirement that the spallation target after it has been expended should be disposable as Class C low-level radioactive waste. Therefore, the disposal of Pb and Bi targets would be optimally beneficial to the economy and environment. Future studies should relate the target performance to other system parameters, specifically solid and liquid blanket systems that contain the radioactive waste to be transmuted. The methodology of this dissertation may be applied to any target material of a high-energy particle accelerator.

Feasibility study of nuclear transmutation by negative muon capture reaction using the PHITS code

NASA Astrophysics Data System (ADS)

Abe, Shin-ichiro; Sato, Tatsuhiko

2016-06-01

Feasibility of nuclear transmutation of fission products in high-level radioactive waste by negative muon capture reaction is investigated using the Particle and Heave Ion Transport code System (PHITS). It is found that about 80 % of stopped negative muons contribute to transmute target nuclide into stable or short-lived nuclide in the case of 135Cs, which is one of the most important nuclide in the transmutation. The simulation result also indicates that the position of transmutation is controllable by changing the energy of incident negative muon. Based on our simulation, it takes approximately 8.5 × 108years to transmute 500 g of 135Cs by negative muon beam with the highest intensity currently available.

Application of neutron transmutation doping method to initially p-type silicon material.

PubMed

Kim, Myong-Seop; Kang, Ki-Doo; Park, Sang-Jun

2009-01-01

The neutron transmutation doping (NTD) method was applied to the initially p-type silicon in order to extend the NTD applications at HANARO. The relationship between the irradiation neutron fluence and the final resistivity of the initially p-type silicon material was investigated. The proportional constant between the neutron fluence and the resistivity was determined to be 2.3473x10(19)nOmegacm(-1). The deviation of the final resistivity from the target for almost all the irradiation results of the initially p-type silicon ingots was at a range from -5% to 2%. In addition, the burn-up effect of the boron impurities, the residual (32)P activity and the effect of the compensation characteristics for the initially p-type silicon were studied. Conclusively, the practical methodology to perform the neutron transmutation doping of the initially p-type silicon ingot was established.

Physics of hydride fueled PWR

NASA Astrophysics Data System (ADS)

Ganda, Francesco

The first part of the work presents the neutronic results of a detailed and comprehensive study of the feasibility of using hydride fuel in pressurized water reactors (PWR). The primary hydride fuel examined is U-ZrH1.6 having 45w/o uranium: two acceptable design approaches were identified: (1) use of erbium as a burnable poison; (2) replacement of a fraction of the ZrH1.6 by thorium hydride along with addition of some IFBA. The replacement of 25 v/o of ZrH 1.6 by ThH2 along with use of IFBA was identified as the preferred design approach as it gives a slight cycle length gain whereas use of erbium burnable poison results in a cycle length penalty. The feasibility of a single recycling plutonium in PWR in the form of U-PuH2-ZrH1.6 has also been assessed. This fuel was found superior to MOX in terms of the TRU fractional transmutation---53% for U-PuH2-ZrH1.6 versus 29% for MOX---and proliferation resistance. A thorough investigation of physics characteristics of hydride fuels has been performed to understand the reasons of the trends in the reactivity coefficients. The second part of this work assessed the feasibility of multi-recycling plutonium in PWR using hydride fuel. It was found that the fertile-free hydride fuel PuH2-ZrH1.6, enables multi-recycling of Pu in PWR an unlimited number of times. This unique feature of hydride fuels is due to the incorporation of a significant fraction of the hydrogen moderator in the fuel, thereby mitigating the effect of spectrum hardening due to coolant voiding accidents. An equivalent oxide fuel PuO2-ZrO2 was investigated as well and found to enable up to 10 recycles. The feasibility of recycling Pu and all the TRU using hydride fuels were investigated as well. It was found that hydride fuels allow recycling of Pu+Np at least 6 times. If it was desired to recycle all the TRU in PWR using hydrides, the number of possible recycles is limited to 3; the limit is imposed by positive large void reactivity feedback.

Coupled thermochemical, isotopic evolution and heat transfer simulations in highly irradiated UO2 nuclear fuel

NASA Astrophysics Data System (ADS)

Piro, M. H. A.; Banfield, J.; Clarno, K. T.; Simunovic, S.; Besmann, T. M.; Lewis, B. J.; Thompson, W. T.

2013-10-01

Predictive capabilities for simulating irradiated nuclear fuel behavior are enhanced in the current work by coupling thermochemistry, isotopic evolution and heat transfer. Thermodynamic models that are incorporated into this framework not only predict the departure from stoichiometry of UO2, but also consider dissolved fission and activation products in the fluorite oxide phase, noble metal inclusions, secondary oxides including uranates, zirconates, molybdates and the gas phase. Thermochemical computations utilize the spatial and temporal evolution of the fission and activation product inventory in the pellet, which is typically neglected in nuclear fuel performance simulations. Isotopic computations encompass the depletion, decay and transmutation of more than 2000 isotopes that are calculated at every point in space and time. These computations take into consideration neutron flux depression and the increased production of fissile plutonium near the fuel pellet periphery (i.e., the so-called “rim effect”). Thermochemical and isotopic predictions are in very good agreement with reported experimental measurements of highly irradiated UO2 fuel with an average burnup of 102 GW d t(U)-1. Simulation results demonstrate that predictions are considerably enhanced when coupling thermochemical and isotopic computations in comparison to empirical correlations. Notice: This manuscript has been authored by UT-Battelle, LLC, under Contract No. DE-AC05-00OR22725 with the U.S. Department of Energy. The United States Government retains and the publisher, by accepting the article for publication, acknowledges that the United States Government retains a non-exclusive, paid-up, irrevocable, world-wide license to publish or reproduce the published form of this manuscript, or allow others to do so, for United States Government purposes.

Accelerator-Driven Subcritical System for Disposing of the U.S. Spent Nuclear Fuel Inventory

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gohar, Yousry; Cao, Yan; Kraus, Adam R.

The current United States inventory of the spent nuclear fuel (SNF) is ~80,000 metric tons of heavy metal (MTHM), including ~131 tons of minor actinides (MAs) and ~669 tons of plutonium. This study describes a conceptual design of an accelerator-driven subcritical (ADS) system for disposing of this SNF inventory by utilizing the 131 tons of MAs inventory and a fraction of the plutonium inventory for energy production, and transmuting some long-lived fission products. An ADS system with a homogeneous subcritical fission blanket was first examined. A spallation neutron source is used to drive the blanket and it is produced frommore » the interaction of a 1-GeV proton beam with a lead-bismuth eutectic (LBE) target. The blanket has a liquid mobile fuel using LBE as the fuel carrier. The fuel materials are dissolved, mixed, or suspended in the liquid fuel carrier. Monte Carlo analyses were performed to determine the overall parameters of the concept. Steady-state Monte Carlo simulations were performed for three similar fission blankets. Except for, the loaded amount of actinide materials in the LBE is either 5, 7, or 10% of the total volume of the blanket, respectively. The neutron multiplication factors of the three blankets are ~0.98 and the initial MAs blanket inventories are ~10 tons. In addition, Monte Carlo burnup simulations using the MCB5 code were performed to analyze the performance of the three conceptual ADS systems. During operation, fresh fuel was fed into the fission blanket to adjust its reactivity and to control the system power. The burnup analysis shows that the three ADS concepts consume about 1.2 tons of actinides per full power year and produce 3 GW thermal power, with a proton beam power of 25 MW. For the blankets with 5, 7, or 10% actinide fuel particles loaded in the LBE, assuming that the ADS systems can be operated for 35 full-power years, the total MA materials consumed in the three ADS systems are about 30.6, 35.3, and 37.2 tons, respectively. Thus, the corresponding numbers of ADS systems to utilize the 131 tons of MA materials of the SNF inventory are 4.3, 3.7, or 3.5, respectively. ADS concepts with tube bundles inserted in the fission blanket were analyzed to overcome the disadvantages of the homogeneous blanket concept. The liquid lead is used as the target material, the mobile fuel carrier, and the primary coolant to avoid the polonium production from bismuth. Reactor physics and thermal-hydraulic analyses were coupled to determine the parameters of the heterogeneous fission blanket. The engineering requirements for a satisfactory operation performance of the HT-9 ferritic steel structure material have been realized. Two heterogeneous concepts of the subcritical fission blanket with the liquid lead mobile fuel inside or outside the tube bundles were considered. The heterogeneous configuration with the mobile fuel inside the tubes showed better performance than the configuration with mobile fuel outside the bundle tubes. The Monte Carlo burnup codes, MCB5 and SERPENT were both used to simulate the fuel burnup in the ADS concepts with the mobile fuels inside the tubes. The burnup analyses were carried out for 35 full power years. The results show that 5 ADS systems can dispose of the total United States inventory of the spent nuclear fuel.« less

Copper Doping of Zinc Oxide by Nuclear Transmutation

DTIC Science & Technology

2014-03-27

Copper Doping of Zinc Oxide by Nuclear Transmutation THESIS Matthew C. Recker, Captain, USAF AFIT-ENP-14-M-30 DEPARTMENT OF THE AIR FORCE AIR...NUCLEAR TRANSMUTATION THESIS Presented to the Faculty Department of Engineering Physics Graduate School of Engineering and Management Air Force...COPPER DOPING OF ZINC OXIDE BY NUCLEAR TRANSMUTATION Matthew C. Recker, BS Captain, USAF Approved: //signed// 27 February 2014 John W. McClory, PhD

Neutron-transmutation-doped germanium bolometers

NASA Technical Reports Server (NTRS)

Palaio, N. P.; Rodder, M.; Haller, E. E.; Kreysa, E.

1983-01-01

Six slices of ultra-pure germanium were irradiated with thermal neutron fluences between 7.5 x 10 to the 16th and 1.88 x 10 to the 18th per sq cm. After thermal annealing the resistivity was measured down to low temperatures (less than 4.2 K) and found to follow the relationship rho = rho sub 0 exp(Delta/T) in the hopping conduction regime. Also, several junction FETs were tested for noise performance at room temperature and in an insulating housing in a 4.2 K cryostat. These FETs will be used as first stage amplifiers for neutron-transmutation-doped germanium bolometers.

Concept of DT fuel cycle for a fusion neutron source

DOE Office of Scientific and Technical Information (OSTI.GOV)

Anan'ev, S.; Spitsyn, A.V.; Kuteev, B.V.

2015-03-15

A concept of DT-fusion neutron source (FNS) with the neutron yield higher than 10{sup 18} neutrons per second is under design in Russia. Such a FNS is of interest for many applications: 1) basic and applied research (neutron scattering, etc); 2) testing the structural materials for fusion reactors; 3) control of sub-critical nuclear systems and 4) nuclear waste processing (including transmutation of minor actinides). This paper describes the fuel cycle concept of a compact fusion neutron source based on a small spherical tokamak (FNS-ST) with a MW range of DT fusion power and considers the key physics issues of thismore » device. The major and minor radii are ∼0.5 and ∼0.3 m, magnetic field ∼1.5 T, heating power less than 15 MW and plasma current 1-2 MA. The system provides the fuel mixture with equal fractions of D and T (D:T = 1:1) for all FNS technology systems. (authors)« less

Determining the minimum required uranium carbide content for HTGR UCO fuel kernels

DOE PAGES

McMurray, Jacob W.; Lindemer, Terrence B.; Brown, Nicholas R.; ...

2017-03-10

There are three important failure mechanisms that must be controlled in high-temperature gas-cooled reactor (HTGR) fuel for certain higher burnup applications are SiC layer rupture, SiC corrosion by CO, and coating compromise from kernel migration. All are related to high CO pressures stemming from free O generated when uranium present as UO 2 fissions and the O is not subsequently bound by other elements. Furthermore, in the HTGR UCO kernel design, CO buildup from excess O is controlled by the inclusion of additional uranium in the form of a carbide, UC x. An approach for determining the minimum UC xmore » content to ensure negligible CO formation was developed and demonstrated using CALPHAD models and the Serpent 2 reactor physics and depletion analysis tool. Our results are intended to be more accurate than previous estimates by including more nuclear and chemical factors, in particular the effect of transmutation products on the oxygen distribution as the fuel kernel composition evolves with burnup.« less

Role of (n,2n) reactions in transmutation of long-lived fission products

DOE Office of Scientific and Technical Information (OSTI.GOV)

Apse, V. A.; Kulikov, G. G., E-mail: ggkulikov@mephi.ru; Kulikov, E. G.

2016-12-15

The conditions under which (n,γ) and (n,2n) reactions can help or hinder each other in neutron transmutation of long-lived fission products (LLFPs) are considered. Isotopic and elemental transmutation for the main long-lived fission products, {sup 79}Se, {sup 93}Zr, {sup 99}Tc, {sup 107}Pd, {sup 126}Sn, {sup 129}I, and {sup 135}Cs, are considered. The effect of (n,2n) reactions on the equilibrium amount of nuclei of the transmuted isotope and the neutron consumption required for the isotope processing is estimated. The aim of the study is to estimate the influence of (n,2n) reactions on efficiency of neutron LLFP transmutation. The code TIME26 andmore » the libraries of evaluated nuclear data ABBN-93, JEF-PC, and JANIS system are applied. The following results are obtained: (1) The effect of (n,2n) reactions on the minimum number of neutrons required for transmutation and the equilibrium amount of LLFP nuclei is estimated. (2) It is demonstrated that, for three LLFP isotopes ({sup 126}Sn, {sup 129}I, and {sup 135}Cs), (n,γ) and (n,2n) reactions are partners facilitating neutron transmutation. The strongest effect of (n,2n) reaction is found for {sup 126}Sn transmutation (reduction of the neutron consumption by 49% and the equilibrium amount of nuclei by 19%).« less

Development and applications of methodologies for the neutronic design of the Pebble Bed Advanced High Temperature Reactor (PB-AHTR)

NASA Astrophysics Data System (ADS)

Fratoni, Massimiliano

This study investigated the neutronic characteristics of the Pebble Bed Advanced High Temperature Reactor (PB-AHTR), a novel nuclear reactor concept that combines liquid salt (7LiF-BeF2---flibe) cooling and TRISO coated-particle fuel technology. The use of flibe enables operation at high power density and atmospheric pressure and improves passive decay-heat removal capabilities, but flibe, unlike conventional helium coolant, is not transparent to neutrons. The flibe occupies 40% of the PB-AHTR core volume and absorbs ˜8% of the neutrons, but also acts as an effective neutron moderator. Two novel methodologies were developed for calculating the time dependent and equilibrium core composition: (1) a simplified single pebble model that is relatively fast; (2) a full 3D core model that is accurate and flexible but computationally intensive. A parametric analysis was performed spanning a wide range of fuel kernel diameters and graphite-to-heavy metal atom ratios to determine the attainable burnup and reactivity coefficients. Using 10% enriched uranium ˜130 GWd/tHM burnup was found to be attainable, when the graphite-to-heavy metal atom ratio (C/HM) is in the range of 300 to 400. At this or smaller C/HM ratio all reactivity coefficients examined---coolant temperature, coolant small and full void, fuel temperature, and moderator temperature, were found to be negative. The PB-AHTR performance was compared to that of alternative options for HTRs, including the helium-cooled pebble-bed reactor and prismatic fuel reactors, both gas-cooled and flibe-cooled. The attainable burnup of all designs was found to be similar. The PB-AHTR generates at least 30% more energy per pebble than the He-cooled pebble-bed reactor. Compared to LWRs the PB-AHTR requires 30% less natural uranium and 20% less separative work per unit of electricity generated. For deep burn TRU fuel made from recycled LWR spent fuel, it was found that in a single pass through the core ˜66% of the TRU can be transmuted; this burnup is slightly superior to that attainable in helium-cooled reactors. A preliminary analysis of the modular variant for the PB-AHTR investigated the triple heterogeneity of this design and determined its performance characteristics.

International programs related to the transmutation of transuranics

DOE Office of Scientific and Technical Information (OSTI.GOV)

Newman, C.

1991-04-01

This report is an account of current programs outside the U.S. relating to the transmutation of transuranics. This work was performed under contract to EPRI. The investigation was based on literature surveys, personal discussions, and visits to European research establishments that are currently active in the area. Research in actinide (uranium plus transuranics) partitioning and transmutation (P-T) is actively promoted in Japan, where the largest program in research on P-T is currently underway; however, following years of relative inactivity, the concept is being revisited elsewhere. Additionally, a significant amount of research in reprocessing and advanced reactors has produced results withmore » interesting possibilities for P-T. Foreign research activities relevant to actinide burning is presented in two sections: foreign national programs, and international programs and working groups. In order to provide the reader with an ability to assess the motivators for or against development of P-T, background on political and institutional trends relating to nuclear waste management is also provided. 38 refs., 17 figs.« less

Concept of grouping in partitioning of HLW for self-consistent fuel cycle

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kitamoto, A.; Mulyanto

1993-12-31

A concept of grouping for partitioning of HLW has been developed in order to examine the possibility of a self-consistent fuel recycle. The concept of grouping of radionuclides is proposed herein, such as Group MA1 (MA below Cm), Group MA2 (Cm and higher MA), Group A ({sup 99}Tc and I), Group B (Cs and Sr) and Group R (the partitioned remain of HLW). Group B is difficult to be transmuted by neutron reaction, so a radiation application in an industrial scale should be developed in the future. Group A and Group MA1 can be burned by a thermal reactor, onmore » the other hand Group MA2 should be burned by a fast reactor. P-T treatment can be optimized for the in-core and out-core system, respectively.« less

Innovative and Advanced Coupled Neutron Transport and Thermal Hydraulic Method (Tool) for the Design, Analysis and Optimization of VHTR/NGNP Prismatic Reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rahnema, Farzad; Garimeela, Srinivas; Ougouag, Abderrafi

2013-11-29

This project will develop a 3D, advanced coarse mesh transport method (COMET-Hex) for steady- state and transient analyses in advanced very high-temperature reactors (VHTRs). The project will lead to a coupled neutronics and thermal hydraulic (T/H) core simulation tool with fuel depletion capability. The computational tool will be developed in hexagonal geometry, based solely on transport theory without (spatial) homogenization in complicated 3D geometries. In addition to the hexagonal geometry extension, collaborators will concurrently develop three additional capabilities to increase the code’s versatility as an advanced and robust core simulator for VHTRs. First, the project team will develop and implementmore » a depletion method within the core simulator. Second, the team will develop an elementary (proof-of-concept) 1D time-dependent transport method for efficient transient analyses. The third capability will be a thermal hydraulic method coupled to the neutronics transport module for VHTRs. Current advancements in reactor core design are pushing VHTRs toward greater core and fuel heterogeneity to pursue higher burn-ups, efficiently transmute used fuel, maximize energy production, and improve plant economics and safety. As a result, an accurate and efficient neutron transport, with capabilities to treat heterogeneous burnable poison effects, is highly desirable for predicting VHTR neutronics performance. This research project’s primary objective is to advance the state of the art for reactor analysis.« less

Transmutation prospect of long-lived nuclear waste induced by high-charge electron beam from laser plasma accelerator

NASA Astrophysics Data System (ADS)

Wang, X. L.; Xu, Z. Y.; Luo, W.; Lu, H. Y.; Zhu, Z. C.; Yan, X. Q.

2017-09-01

Photo-transmutation of long-lived nuclear waste induced by a high-charge relativistic electron beam (e-beam) from a laser plasma accelerator is demonstrated. A collimated relativistic e-beam with a high charge of approximately 100 nC is produced from high-intensity laser interaction with near-critical-density (NCD) plasma. Such e-beam impinges on a high-Z convertor and then radiates energetic bremsstrahlung photons with flux approaching 1011 per laser shot. Taking a long-lived radionuclide 126Sn as an example, the resulting transmutation reaction yield is the order of 109 per laser shot, which is two orders of magnitude higher than obtained from previous studies. It is found that at lower densities, a tightly focused laser irradiating relatively longer NCD plasmas can effectively enhance the transmutation efficiency. Furthermore, the photo-transmutation is generalized by considering mixed-nuclide waste samples, which suggests that the laser-accelerated high-charge e-beam could be an efficient tool to transmute long-lived nuclear waste.

Report on the Synchrotron Characterization of U-Mo and U-Zr Alloys and the Modeling Results

DOE Office of Scientific and Technical Information (OSTI.GOV)

Okuniewski, Maria A.; Ganapathy, Varsha; Hamilton, Brenden

2016-09-01

ABSTRACT Uranium-molybdenum (U-Mo) and uranium-zirconium (U-Zr) are two promising fuel candidates for nuclear transmutation reactors which burn long-lived minor actinides and fission products within fast spectrum reactors. The objectives of this research are centered on understanding the early stages of fuel performance through the examination of the irradiation induced microstructural changes in U-Zr and U-Mo alloys subjected to low neutron fluences. Specimens that were analyzed include those that were previously irradiated in the Advanced Test Reactor at INL. This most recent work has focused on a sub-set of the irradiated specimens, specifically U-Zr and U-Mo alloys that were irradiated tomore » 0.01 dpa at temperatures ranging from (150-800oC). These specimens were analyzed with two types of synchrotron techniques, including X-ray absorption fine structure and X-ray diffraction. These techniques provide non-destructive microstructural analysis, including phase identification and quantitation, lattice parameters, crystallite sizes, as well as bonding, structure, and chemistry. Preliminary research has shown changes in the phase fractions, crystallite sizes, and lattice parameters as a function of irradiation and temperature. Future data analyses will continue to explore these microstructural changes.« less

The feasibility of using molten carbonate corrosion for separating a nuclear surrogate for plutonium oxide from silicon carbide inert matrix

NASA Astrophysics Data System (ADS)

Cheng, Ting; Baney, Ronald H.; Tulenko, James

2010-10-01

Silicon carbide is one of the prime candidates as a matrix material in inert matrix fuels (IMF) being designed to reduce the plutonium inventories. Since complete fission and transmutation is not practical in a single in-core run, it is necessary to separate the non-transmuted actinide materials from the silicon carbide matrix for recycling. In this work, SiC was corroded in sodium carbonate (Na 2CO 3) and potassium carbonate (K 2CO 3), to form water soluble sodium or potassium silicate. Separation of the transuranics was achieved by dissolving the SiC corrosion product in boiling water. Ceria (CeO 2), which was used as a surrogate for plutonium oxide (PuO 2), was not corroded in these molten salt environments. The molten salt depth, which is a distance between the salt/air interface to the upper surface of SiC pellets, significantly affected the rate of corrosion. The corrosion was faster in K 2CO 3 than in Na 2CO 3 molten salt at 1050 °C, when the initial molten salt depths were kept the same for both salts.

LOS ALAMOS NEUTRON SCIENCE CENTER CONTRIBUTIONS TO THE DEVELOPMENT OF FUTURE POWER REACTORS

DOE Office of Scientific and Technical Information (OSTI.GOV)

GAVRON, VICTOR I.; HILL, TONY S.; PITCHER, ERIC J.

The Los Alamos Neutron Science Center (LANSCE) is a large spallation neutron complex centered around an 800 MeV high-currently proton accelerator. Existing facilities include a highly-moderated neutron facility (Lujan Center) where neutrons between thermal and keV energies are produced, and the Weapons Neutron Research Center (WNR), where a bare spallation target produces neutrons between 0.1 and several hundred MeV.The LANSCE facility offers a unique capability to provide high precision nuclear data over a large energy region, including that for fast reactor systems. In an ongoing experimental program the fission and capture cross sections are being measured for a number ofmore » minor actinides relevant for Generation-IV reactors and transmutation technology. Fission experiments makes use of both the highly moderated spallation neutron spectrum at the Lujan Center, and the unmoderated high energy spectrum at WNR. By combininb measurements at these two facilities the differential fission cross section is measured relative to the {sup 235}U(n,f) standard from subthermal energies up to about 200 MeV. An elaborate data acquisition system is designed to deal with all the different types of background present when spanning 10 energy decades. The first isotope to be measured was {sup 237}Np, and the results were used to improve the current ENDF/B-VII evaluation. Partial results have also been obtained for {sup 240}Pu and {sup 242}Pu, and the final results are expected shortly. Capture cross sections are measured at LANSCE using the Detector for Advanced Neutron Capture Experiments (DANCE). This unique instrument is highly efficient in detecting radiative capture events, and can thus handle radioactive samples of half-lives as low as 100 years. A number of capture cross sections important to fast reaction applications have been measured with DANCE. The first measurement was on {sup 237}Np(n,{gamma}), and the results have been submitted for publication. Other capture measurements in progress include {sup 240}Pu and {sup 242}Pu. The United States recently announced the Global Nuclear Energy Partnership (GNEP), with the goal of closing the commercial nuclear fuel cycle while minimizing proliferation risk. GNEP achieves these goals using fast-spectrum nuclear reactors powered by new transmutation fuels that contain significant quantities of minor actinides. The proposed Materials Test Station (MTS) will provide the GNEP with a cost-effective means of obtaining domestic fast-spectrum irradiations of advanced transmutation fuel forms and structural materials, which is an important step in the fuels qualification process. The MTS will be located at the LANSCE, and will be driven by a 1.08-MW proton beam. Th epeak neutron flux in the irradiation region is 1.67 x 10{sup 15} n/cm{sup 2}/s, and the energy spectrum is similar to that of a fast reactor, with the addition of a high-energy tail. The facility is expected to operate at least 4,400 hours per year. Fuel burnup rates will exceed 4% per year, and the radiation damage rate in iron will be 18 dpa (displacements per atom) per year. The construction cost is estimated to be $73M (including 25% contingency), with annual operating costs in the range of $6M to $10M. Appropriately funded, the MTS could begin operation in 2010.« less

Neutron transport-burnup code MCORGS and its application in fusion fission hybrid blanket conceptual research

NASA Astrophysics Data System (ADS)

Shi, Xue-Ming; Peng, Xian-Jue

2016-09-01

Fusion science and technology has made progress in the last decades. However, commercialization of fusion reactors still faces challenges relating to higher fusion energy gain, irradiation-resistant material, and tritium self-sufficiency. Fusion Fission Hybrid Reactors (FFHR) can be introduced to accelerate the early application of fusion energy. Traditionally, FFHRs have been classified as either breeders or transmuters. Both need partition of plutonium from spent fuel, which will pose nuclear proliferation risks. A conceptual design of a Fusion Fission Hybrid Reactor for Energy (FFHR-E), which can make full use of natural uranium with lower nuclear proliferation risk, is presented. The fusion core parameters are similar to those of the International Thermonuclear Experimental Reactor. An alloy of natural uranium and zirconium is adopted in the fission blanket, which is cooled by light water. In order to model blanket burnup problems, a linkage code MCORGS, which couples MCNP4B and ORIGEN-S, is developed and validated through several typical benchmarks. The average blanket energy Multiplication and Tritium Breeding Ratio can be maintained at 10 and 1.15 respectively over tens of years of continuous irradiation. If simple reprocessing without separation of plutonium from uranium is adopted every few years, FFHR-E can achieve better neutronic performance. MCORGS has also been used to analyze the ultra-deep burnup model of Laser Inertial Confinement Fusion Fission Energy (LIFE) from LLNL, and a new blanket design that uses Pb instead of Be as the neutron multiplier is proposed. In addition, MCORGS has been used to simulate the fluid transmuter model of the In-Zinerater from Sandia. A brief comparison of LIFE, In-Zinerater, and FFHR-E will be given.

Friction Stir Welding of HT9 Ferritic-Martensitic Steel: An Assessment of Microstructure and Properties

DTIC Science & Technology

2013-06-01

X, where X represents lithium, sodium, beryllium, or transmutation products, such as tritium [47]. In this mechanism, the transmutation of lithium...Similar to the study by Williams, Farmer found that galvanic coupling, increased temperature and the formation of transmutation products (HF and TF), a

An improved transmutation method for quantitative determination of the components in multicomponent overlapping chromatograms.

PubMed

Shao, Xueguang; Yu, Zhengliang; Ma, Chaoxiong

2004-06-01

An improved method is proposed for the quantitative determination of multicomponent overlapping chromatograms based on a known transmutation method. To overcome the main limitation of the transmutation method caused by the oscillation generated in the transmutation process, two techniques--wavelet transform smoothing and the cubic spline interpolation for reducing data points--were adopted, and a new criterion was also developed. By using the proposed algorithm, the oscillation can be suppressed effectively, and quantitative determination of the components in both the simulated and experimental overlapping chromatograms is successfully obtained.

Unifying relations for scattering amplitudes

NASA Astrophysics Data System (ADS)

Cheung, Clifford; Shen, Chia-Hsien; Wen, Congkao

2018-02-01

We derive new amplitudes relations revealing a hidden unity among a wideranging variety of theories in arbitrary spacetime dimensions. Our results rely on a set of Lorentz invariant differential operators which transmute physical tree-level scattering amplitudes into new ones. By transmuting the amplitudes of gravity coupled to a dilaton and two-form, we generate all the amplitudes of Einstein-Yang-Mills theory, Dirac-Born-Infield theory, special Galileon, nonlinear sigma model, and biadjoint scalar theory. Transmutation also relates amplitudes in string theory and its variants. As a corollary, celebrated aspects of gluon and graviton scattering like color-kinematics duality, the KLT relations, and the CHY construction are inherited traits of the transmuted amplitudes. Transmutation recasts the Adler zero as a trivial consequence of the Weinberg soft theorem and implies new subleading soft theorems for certain scalar theories.

OECD/NEA Ongoing activities related to the nuclear fuel cycle

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cornet, S.M.; McCarthy, K.; Chauvin, N.

2013-07-01

As part of its role in encouraging international collaboration, the OECD Nuclear Energy Agency is coordinating a series of projects related to the Nuclear Fuel Cycle. The Nuclear Science Committee (NSC) Working Party on Scientific Issues of the Nuclear Fuel Cycle (WPFC) comprises five different expert groups covering all aspects of the fuel cycle from front to back-end. Activities related to fuels, materials, physics, separation chemistry, and fuel cycles scenarios are being undertaken. By publishing state-of-the-art reports and organizing workshops, the groups are able to disseminate recent research advancements to the international community. Current activities mainly focus on advanced nuclearmore » systems, and experts are working on analyzing results and establishing challenges associated to the adoption of new materials and fuels. By comparing different codes, the Expert Group on Advanced Fuel Cycle Scenarios is aiming at gaining further understanding of the scientific issues and specific national needs associated with the implementation of advanced fuel cycles. At the back end of the fuel cycle, separation technologies (aqueous and pyrochemical processing) are being assessed. Current and future activities comprise studies on minor actinides separation and post Fukushima studies. Regular workshops are also organized to discuss recent developments on Partitioning and Transmutation. In addition, the Nuclear Development Committee (NDC) focuses on the analysis of the economics of nuclear power across the fuel cycle in the context of changes of electricity markets, social acceptance and technological advances and assesses the availability of the nuclear fuel and infrastructure required for the deployment of existing and future nuclear power. The Expert Group on the Economics of the Back End of the Nuclear Fuel Cycle (EBENFC), in particular, is looking at assessing economic and financial issues related to the long term management of spent nuclear fuel. (authors)« less

An overview of DANCE: a 4II BaF[2] detector for neutron capture measurements at LANSCE.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ullmann, J. L.

2004-01-01

The Detector for Advanced Neutron Capture experiments (DANCE) is a 162-element, 4{pi} BaF{sub 2} array designed to make neutron capture cross-section measurements on rare or radioactive targets with masses as little as 1 mg. Accurate capture cross sections are needed in many research areas, including stellar nucleosynthesis, advanced nuclear fuel cycles, waste transmutation, and other applied programs. These cross sections are difficult to calculate accurately and must be measured. Up to now, except for a few long-lived nuclides there are essentially no differential capture measurements on radioactive nuclei. The DANCE array is located at the Lujan Neutron Scattering Center atmore » LANSCE, which is a continuous-spectrum neutron source with useable energies from below thermal to about 100 keV. Data acquisition is done with 320 fast waveform digitizers. The design and initial performance results, including background minimization, will be discussed.« less

Dissolution behavior of MgO based inert matrix fuel for the transmutation of minor actinides

NASA Astrophysics Data System (ADS)

Mühr-Ebert, E. L.; Lichte, E.; Bukaemskiy, A.; Finkeldei, S.; Klinkenberg, M.; Brandt, F.; Bosbach, D.; Modolo, G.

2018-07-01

This study explores the dissolution properties of magnesia-based inert matrix nuclear fuel (IMF) containing transuranium elements (TRU). Pure MgO pellets as well as MgO pellets containing CeO2, as surrogate for TRU oxides, and are considered as model systems for genuine magnesia based inert matrix fuel were fabricated. The aim of this study is to identify conditions at which the matrix material can be selectively dissolved during the head-end reprocessing step, allowing a separation of MgO from the actinides, whereas the actinides remain undissolved. The dissolution behavior was studied in macroscopic batch experiments as a function of nitric acid concentration, dissolution medium volume, temperature, stirring velocity, and pellet density (85, 90, 96, and 99%TD). To mimic pellets with various burn-ups the density of the here fabricated pellets was varied. MgO is soluble even under mild conditions (RT, 2.5 mol/L HNO3). The dissolution rates of MgO at different acid concentrations are rather similar, whereas the dissolution rate is strongly dependent on the temperature. Via a microscopic approach, a model was developed to describe the evolution of the pellet surface area during dissolution and determine a surface normalized dissolution rate. Moreover, dissolution rates of the inert matrix fuel containing CeO2 were determined as a function of the acid concentration and temperature. During the dissolution of MgO/CeO2 pellets the MgO dissolves completely, while CeO2 (>99%) remains undissolved. This study intends to provide a profound understanding of the chemical performance of magnesia based IMF containing fissile material. The feasibility of the dissolution of magnesia based IMF with nitric acid is discussed.

Transmutation Theory in the Greek Alchemical Corpus.

PubMed

Dufault, Olivier

2015-08-01

This paper studies transmutation theory as found in the texts attributed to Zosimus of Panopolis, "the philosopher Synesius," and "the philosopher Olympiodorus of Alexandria." It shows that transmutation theory (i.e. a theory explaining the complete transformation of substances) is mostly absent from the work attributed to these three authors. The text attributed to Synesius describes a gilding process, which is similar to those described by Pliny and Vitruvius. The commentary attributed to Olympiodorus is the only text studied here that describes something similar to a transmutation theory. It is unclear, however, if this was a theory of transmutation or if the writer meant something more like the literal meaning of the word "ekstrophē," a term used to describe the transformation of metals, as the "turning inside-out" of what is hidden in a substance. A similar conception of ekstrophē can be found in the works of Zosimus, who discussed transmutation to make an analogy with self-purification processes, which, from the perspective of his own anthropogony, consisted in the "turning inside-out" of the "inner human" (esō anthrōpos).

Accelerator Reactor Coupling for Energy Production in Advanced Nuclear Fuel Cycles

DOE PAGES

Brown, Nicholas R.; Heidet, Florent; Haj Tahar, Malek

2016-01-01

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 systemsmore » 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.« less

Accelerator–Reactor Coupling for Energy Production in Advanced Nuclear Fuel Cycles

DOE Office of Scientific and Technical Information (OSTI.GOV)

Heidet, Florent; Brown, Nicholas R.; Haj Tahar, Malek

2015-01-01

This article is a review of several accelerator-reactor interface issues and nuclear fuel cycle applications of accelerator-driven 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 focused on issues of interest, e.g. the impact of the energy required to run the accelerator and associated systems onmore » 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 a critically important. One interesting challenge is measurement of blanket subcriticality level during operation. We also reviewed 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 versus a critical fast reactor with recycle of uranium and plutonium.« less

Accelerator-Reactor Coupling for Energy Production in Advanced Nuclear Fuel Cycles

NASA Astrophysics Data System (ADS)

Heidet, Florent; Brown, Nicholas R.; Haj Tahar, Malek

This article is a review of several accelerator-reactor interface issues and nuclear fuel cycle applications of accelerator-driven 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.

Back-end of the fuel cycle - Indian scenario

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wattal, P.K.

Nuclear power has a key role in meeting the energy demands of India. This can be sustained by ensuring robust technology for the back end of the fuel cycle. Considering the modest indigenous resources of U and a huge Th reserve, India has adopted a three stage Nuclear Power Programme (NPP) based on 'closed fuel cycle' approach. This option on 'Recovery and Recycle' serves twin objectives of ensuring adequate supply of nuclear fuel and also reducing the long term radio-toxicity of the wastes. Reprocessing of the spent fuel by Purex process is currently employed. High Level Liquid Waste (HLW) generatedmore » during reprocessing is vitrified and undergoes interim storage. Back-end technologies are constantly modified to address waste volume minimization and radio-toxicity reduction. Long-term management of HLW in Indian context would involve partitioning of long lived minor actinides and recovery of valuable fission products specifically cesium. Recovery of minor actinides from HLW and its recycle is highly desirable for the sustained growth of India's NPPs. In this context, programme for developing and deploying partitioning technologies on industrial scale is pursued. The partitioned elements could be either transmuted in Fast Reactors (FRs)/Accelerated Driven Systems (ADS) as an integral part of sustainable Indian NPP. (authors)« less

Radiotoxicity Characterization of Multi-Recycled Thorium Fuel - 12394

DOE Office of Scientific and Technical Information (OSTI.GOV)

Franceschini, F.; Wenner, M.; Fiorina, C.

2012-07-01

As described in companion papers, Westinghouse is proposing the implementation of a thorium based fuel cycle to burn the transuranic (TRU) contained in the used nuclear fuel. The potential of thorium as a TRU burner is described in another paper presented at this conference. This paper analyzes the long-term impact of thorium on the front-end and backend of the fuel cycle. This is accomplished by an assessment of the isotopic make-up of Th in a closed cycle and its impact on representative metrics, such as radiotoxicity, decay heat and gamma heat. The behavior in both thermal and fast neutron energymore » ranges has been investigated. Irradiation in a Th fuel PWR has been assumed as representative of the thermal range, while a Th fuel fast reactor (FR) has been employed to characterize the behavior in the high-energy range. A comparison with a U-fuel closed-cycle FR has been undertaken in an attempt of a more comprehensive evaluation of each cycle's long-term potential. As the Th fuel undergoes multiple cycles of irradiation, the isotopic composition of the recycled fuel changes. Minor Th isotopes are produced; U-232 and Pa-231 build up; the U vector gradually shifts towards increasing amounts of U-234, U-235 etc., eventually leading to the production of non negligible amounts of TRU isotopes, especially Pu-238. The impact of the recycled fuel isotopic makeup on the in-core behavior is mild, and for some aspects beneficial, i.e. the reactivity swing during irradiation is reduced as the fertile characteristics of the fuel increase. On the other hand, the front and the back-end of the fuel cycle are negatively affected due to the presence of Th-228 and U-232 and the build-up of higher actinides (Pu-238 etc.). The presence of U-232 can also be seen as advantageous as it represents an obstacle to potential proliferators. Notwithstanding the increase in the short-term radiotoxicity and decay heat in the multi-recycled fuel, the Th closed cycle has some potentially substantial advantages compared to the U cycle, such as the smaller actinide radiotoxicity and decay heat for up to 25,000 years after irradiation. In order for these benefits to materialize, the capability to reprocess and remotely manufacture industrial amounts of recycled fuel appears to be the key. Westinghouse is proposing the implementation of a thorium based fuel cycle to burn the TRU contained in the current UNF. The general approach and the potential of thorium as TRU burner is described in other papers presented at this conference. The focus of this paper is to analyze the long-term potential of thorium, once the legacy TRU has been exhausted and the thorium reactor system will become self-sufficient. Therefore, a comparison of Th closed cycle, in fast and thermal neutron energy ranges, vs. U closed cycle, in the fast energy range, has been undertaken. The results presented focus on selected backend and front-end metrics: isotopic actinide composition and potential implications on ingested radiotoxicity, decay heat and gamma heat. The evaluation confirms potential substantial improvements in the backend of the fuel cycle by transitioning to a thorium closed cycle. These benefits are the result of a much lower TRU content, in particular Pu-241, Am-241 and Pu-240, characterizing the Th vs. U actinide inventories, and the ensuing process waste to be disposed. On the other hand, the larger gamma activity of Th recycled fuel, consisting predominantly of hard gammas from U-232's decay products, is a significant challenge for fuel handling, transportation and manufacturing but can be claimed as beneficial for the proliferation resistance of the fuel. It is worth remembering that in our perspective the Th closed cycle and the U closed cycle will follow a transmutation phase which will likely take place over several decades and dictate the technologies required. These will likely include remote fuel manufacturing, regardless of the specific system adopted for the transmutation, which could then be inherited for the ensuing closed cycles. Finally, specific data related to the fuel manufacturing and separation technologies and their performance in the prospected industrial scale deployment, are key for further quantification of the potential merits of the options explored. Further studies in this direction should be warranted before making definitive conclusion. (authors)« less

Transmutation doping of silicon solar cells

NASA Technical Reports Server (NTRS)

Wood, R. F.; Westbrook, R. D.; Young, R. T.; Cleland, J. W.

1977-01-01

Normal isotopic silicon contains 3.05% of Si-30 which transmutes to P-31 after thermal neutron absorption, with a half-life of 2.6 hours. This reaction is used to introduce extremely uniform concentrations of phosphorus into silicon, thus eliminating the areal and spatial inhomogeneities characteristic of chemical doping. Annealing of the lattice damage in the irradiated silicon does not alter the uniformity of dopant distribution. Transmutation doping also makes it possible to introduce phosphorus into polycrystalline silicon without segregation of the dopant at the grain boundaries. The use of neutron transmutation doped (NTD) silicon in solar cell research and development is discussed.

High Temperature Reactor (HTR) Deep Burn Core and Fuel Analysis: Design Selection for the Prismatic Block Reactor With Results from FY-2011 Activities

DOE Office of Scientific and Technical Information (OSTI.GOV)

Michael A. Pope

2011-10-01

The Deep Burn (DB) Project is a U.S. Department of Energy sponsored feasibility study of Transuranic Management using high burnup fuel in the high temperature helium cooled reactor (HTR). The DB Project consists of seven tasks: project management, core and fuel analysis, spent fuel management, fuel cycle integration, TRU fuel modeling, TRU fuel qualification, and HTR fuel recycle. In the Phase II of the Project, we conducted nuclear analysis of TRU destruction/utilization in the HTR prismatic block design (Task 2.1), deep burn fuel/TRISO microanalysis (Task 2.3), and synergy with fast reactors (Task 4.2). The Task 2.1 covers the core physicsmore » design, thermo-hydraulic CFD analysis, and the thermofluid and safety analysis (low pressure conduction cooling, LPCC) of the HTR prismatic block design. The Task 2.3 covers the analysis of the structural behavior of TRISO fuel containing TRU at very high burnup level, i.e. exceeding 50% of FIMA. The Task 4.2 includes the self-cleaning HTR based on recycle of HTR-generated TRU in the same HTR. Chapter IV contains the design and analysis results of the 600MWth DB-HTR core physics with the cycle length, the average discharged burnup, heavy metal and plutonium consumptions, radial and axial power distributions, temperature reactivity coefficients. Also, it contains the analysis results of the 450MWth DB-HTR core physics and the analysis of the decay heat of a TRU loaded DB-HTR core. The evaluation of the hot spot fuel temperature of the fuel block in the DB-HTR (Deep-Burn High Temperature Reactor) core under full operating power conditions are described in Chapter V. The investigated designs are the 600MWth and 460MWth DB-HTRs. In Chapter VI, the thermo-fluid and safety of the 600MWth DB-HTRs has been analyzed to investigate a thermal-fluid design performance at the steady state and a passive safety performance during an LPCC event. Chapter VII describes the analysis results of the TRISO fuel microanalysis of the 600MWth and 450MWth DB-HTRs. The TRISO fuel microanalysis covers the gas pressure buildup in a coated fuel particle including helium production, the thermo-mechanical behavior of a CFP, the failure probabilities of CFPs, the temperature distribution in a CPF, and the fission product (FP) transport in a CFP and a graphite. In Chapter VIII, it contains the core design and analysis of sodium cooled fast reactor (SFR) with deep burn HTR reactor. It considers a synergistic combination of the DB-MHR and an SFR burner for a safe and efficient transmutation of the TRUs from LWRs. Chapter IX describes the design and analysis results of the self-cleaning (or self-recycling) HTR core. The analysis is considered zero and 5-year cooling time of the spent LWR fuels.« less

High Temperature Reactor (HTR) Deep Burn Core and Fuel Analysis: Design Selection for the Prismatic Block Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Francesco Venneri; Chang-Keun Jo; Jae-Man Noh

2010-09-01

The Deep Burn (DB) Project is a U.S. Department of Energy sponsored feasibility study of Transuranic Management using high burnup fuel in the high temperature helium cooled reactor (HTR). The DB Project consists of seven tasks: project management, core and fuel analysis, spent fuel management, fuel cycle integration, TRU fuel modeling, TRU fuel qualification, and HTR fuel recycle. In the Phase II of the Project, we conducted nuclear analysis of TRU destruction/utilization in the HTR prismatic block design (Task 2.1), deep burn fuel/TRISO microanalysis (Task 2.3), and synergy with fast reactors (Task 4.2). The Task 2.1 covers the core physicsmore » design, thermo-hydraulic CFD analysis, and the thermofluid and safety analysis (low pressure conduction cooling, LPCC) of the HTR prismatic block design. The Task 2.3 covers the analysis of the structural behavior of TRISO fuel containing TRU at very high burnup level, i.e. exceeding 50% of FIMA. The Task 4.2 includes the self-cleaning HTR based on recycle of HTR-generated TRU in the same HTR. Chapter IV contains the design and analysis results of the 600MWth DB-HTR core physics with the cycle length, the average discharged burnup, heavy metal and plutonium consumptions, radial and axial power distributions, temperature reactivity coefficients. Also, it contains the analysis results of the 450MWth DB-HTR core physics and the analysis of the decay heat of a TRU loaded DB-HTR core. The evaluation of the hot spot fuel temperature of the fuel block in the DB-HTR (Deep-Burn High Temperature Reactor) core under full operating power conditions are described in Chapter V. The investigated designs are the 600MWth and 460MWth DB-HTRs. In Chapter VI, the thermo-fluid and safety of the 600MWth DB-HTRs has been analyzed to investigate a thermal-fluid design performance at the steady state and a passive safety performance during an LPCC event. Chapter VII describes the analysis results of the TRISO fuel microanalysis of the 600MWth and 450MWth DB-HTRs. The TRISO fuel microanalysis covers the gas pressure buildup in a coated fuel particle including helium production, the thermo-mechanical behavior of a CFP, the failure probabilities of CFPs, the temperature distribution in a CPF, and the fission product (FP) transport in a CFP and a graphite. In Chapter VIII, it contains the core design and analysis of sodium cooled fast reactor (SFR) with deep burn HTR reactor. It considers a synergistic combination of the DB-MHR and an SFR burner for a safe and efficient transmutation of the TRUs from LWRs. Chapter IX describes the design and analysis results of the self-cleaning (or self-recycling) HTR core. The analysis is considered zero and 5-year cooling time of the spent LWR fuels.« less

Object kinetic Monte Carlo model for neutron and ion irradiation in tungsten: Impact of transmutation and carbon impurities

NASA Astrophysics Data System (ADS)

Castin, N.; Bonny, G.; Bakaev, A.; Ortiz, C. J.; Sand, A. E.; Terentyev, D.

2018-03-01

We upgrade our object kinetic Monte Carlo (OKMC) model, aimed at describing the microstructural evolution in tungsten (W) under neutron and ion irradiation. Two main improvements are proposed based on recently published atomistic data: (a) interstitial carbon impurities, and their interaction with radiation-induced defects (point defect clusters and loops), are more accurately parameterized thanks to ab initio findings; (b) W transmutation to rhenium (Re) upon neutron irradiation, impacting the diffusivity of radiation defects, is included, also relying on recent atomistic data. These essential amendments highly improve the portability of our OKMC model, providing a description for the formation of SIA-type loops under different irradiation conditions. The model is applied to simulate neutron and ion irradiation in pure W samples, in a wide range of fluxes and temperatures. We demonstrate that it performs a realistic prediction of the population of TEM-visible voids and loops, as compared to experimental evidence. The impact of the transmutation of W to Re, and of carbon trapping, is assessed.

Modeling Radioactive Decay Chains with Branching Fraction Uncertainties

DTIC Science & Technology

2013-03-01

moments methods with transmutation matrices. Uncertainty from both half-lives and branching fractions is carried through these calculations by Monte...moment methods, method for sampling from normal distributions for half- life uncertainty, and use of transmutation matrices were leveraged. This...distributions for half-life and branching fraction uncertainties, building decay chains and generating the transmutation matrix (T-matrix

PRELIMINARY DATA CALL REPORT ADVANCED BURNER REACTOR START UP FUEL FABRICATION FACILITY

DOE Office of Scientific and Technical Information (OSTI.GOV)

S. T. Khericha

2007-04-01

The purpose of this report is to provide data for preparation of a NEPA Environmental Impact Statement in support the U. S. Department of Energy (DOE) Global Nuclear Energy Partnership (GNEP). One of the GNEP objectives is to reduce the inventory of long lived actinide from the light water reactor (LWR) spent fuel. The LWR spent fuel contains Plutonium (Pu) -239 and other transuranics (TRU) such as Americium-241. One of the options is to transmute or burn these actinides in fast neutron spectra as well as generate the electricity. A sodium-cooled Advanced Recycling Reactor (ARR) concept has been proposed tomore » achieve this goal. However, fuel with relatively high TRU content has not been used in the fast reactor. To demonstrate the utilization of TRU fuel in a fast reactor, an Advanced Burner Reactor (ABR) prototype of ARR is proposed, which would necessarily be started up using weapons grade (WG) Pu fuel. The WG Pu is distinguished by relatively highest proportions of Pu-239 and lesser amount of other actinides. The WG Pu will be used as the startup fuel along with TRU fuel in lead test assemblies. Because such fuel is not currently being produced in the US, a new facility (or new capability in an existing facility) is being considered for fabrication of WG Pu fuel for the ABR. This report is provided in response to ‘Data Call’ for the construction of startup fuel fabrication facility. It is anticipated that the facility will provide the startup fuel for 10-15 years and will take to 3 to 5 years to construct.« less

A Blueprint for GNEP Advanced Burner Reactor Startup Fuel Fabrication Facility

DOE Office of Scientific and Technical Information (OSTI.GOV)

S. Khericha

2010-12-01

The purpose of this article is to identify the requirements and issues associated with design of GNEP Advanced Burner Reactor Fuel Facility. The report was prepared in support of providing data for preparation of a NEPA Environmental Impact Statement in support the U. S. Department of Energy (DOE) Global Nuclear Energy Partnership (GNEP). One of the GNEP objectives was to reduce the inventory of long lived actinide from the light water reactor (LWR) spent fuel. The LWR spent fuel contains Plutonium (Pu) -239 and other transuranics (TRU) such as Americium-241. One of the options is to transmute or burn thesemore » actinides in fast neutron spectra as well as generate the electricity. A sodium-cooled Advanced Recycling Reactor (ARR) concept was proposed to achieve this goal. However, fuel with relatively high TRU content has not been used in the fast reactor. To demonstrate the utilization of TRU fuel in a fast reactor, an Advanced Burner Reactor (ABR) prototype of ARR was proposed, which would necessarily be started up using weapons grade (WG) Pu fuel. The WG Pu is distinguished by relatively highest proportions of Pu-239 and lesser amount of other actinides. The WG Pu was assumed to be used as the startup fuel along with TRU fuel in lead test assemblies. Because such fuel is not currently being produced in the US, a new facility (or new capability in an existing facility) was being considered for fabrication of WG Pu fuel for the ABR. It was estimated that the facility will provide the startup fuel for 10-15 years and would take 3 to 5 years to construct.« less

Helium in inert matrix dispersion fuels

NASA Astrophysics Data System (ADS)

van Veen, A.; Konings, R. J. M.; Fedorov, A. V.

2003-07-01

The behaviour of helium, an important decay product in the transmutation chains of actinides, in dispersion-type inert matrix fuels is discussed. A phenomenological description of its accumulation and release in CERCER and CERMET fuel is given. A summary of recent He-implantation studies with inert matrix metal oxides (ZrO 2, MgAl 2O 4, MgO and Al 2O 3) is presented. A general picture is that for high helium concentrations helium and vacancy defects form helium clusters which convert into over-pressurized bubbles. At elevated temperature helium is released from the bubbles. On some occasions thermal stable nano-cavities or nano-pores remain. On the basis of these results the consequences for helium induced swelling and helium storage in oxide matrices kept at 800-1000 °C will be discussed. In addition, results of He-implantation studies for metal matrices (W, Mo, Nb and V alloys) will be presented. Introduction of helium in metals at elevated temperatures leads to clustering of helium to bubbles. When operational temperatures are higher than 0.5 melting temperature, swelling and helium embrittlement might occur.

Recent upgrades and new scientific infrastructure of MARIA research reactor, Otwock-Swierk, Poland

DOE Office of Scientific and Technical Information (OSTI.GOV)

NONE

The MARIA reactor is open-pool type, water and beryllium moderated. It has two independent primary cooling systems: fuel and pool cooling system. Each fuel assembly is cooled down separately in pressurized channels with individual performances characterization. The fuel assemblies consist of five layers of bent plates or six concentric tubes. Currently it is one of the most powerful research reactors in Europe with operation availability at least up to 2030. Its nominal thermal power is 30 MW. It is characterized by high neutron flux density: up to 3x10{sup 14} n cm{sup -2} s{sup -1} in case of thermal neutrons, andmore » up to 2x10{sup 13} n cm{sup -2} s{sup -1} in case of fast neutrons. The reactor is operated for ca. 4000 h per year. The reactor facility is equipped with fully equipped three hot cells with shielding up to 10{sup 15} Bq. Adjacent to the reactor facility, the radio-pharmaceutics plant (POLATOM) and Material Research Laboratory are located. They are equipped with a number of hot cells with instrumentation. The transport system of radioactive materials from reactor facility to Material Research Laboratory is available. During 2014 the MARIA reactor has been operated with three different types of fuel the same time: previous 36% enriched fuel, and two types of new LEU fuels. In the meantime, molybdenum irradiation programme has been developed. Maria is a multifunctional research tool, with a notable application in production of radioisotopes, radio-pharmaceutics manufacturing (ca. 600 TBq/y), {sup 99}Mo for medical scintigraphy (ca. 6000 TBq/y), neutron transmutation doping of silicon single crystals, wide scientific research based on neutron beams utilization. From the beginning MARIA reactor was intended for loop and fuel testing research activities. Currently it is used mostly as material testing and irradiation facility and for that reason it has wide experimental capabilities. There are eight horizontal irradiation channels from among whom six of them are equipped with instrumentation for condensed matter physics research: - H3 - spectrometer and diffractometer with double monochromator; - H4 - small angle scattering spectrometer; - H5 - polarized neutrons spectrometer; - H6, H7 - two 3-axial crystal neutron spectrometers; - H8 - neutron radiography stand. For two horizontal channels are ongoing exploitation programs: - H2 - station with epithermal neutron beam produced in uranium converter is being developed. Intelligent converter will be installed on the periphery of reactor core. The intensity of the beam will be at the level 2x10{sup 9} n cm{sup -2}s{sup -1} what makes the beam unique in the Europe. - H1 - special pneumatic horizontal mail is being developed for irradiation material samples in the vicinity of the core i.e. in the distal part of the H1 channel. The number of neutron irradiation facilities in MARIA reactor is increasing every year. Numerous of thermal neutron irradiation channels including fast hydraulic rabbit system and large size channels for fast neutron irradiation are used routinely. Recently new in-pile facility with ITER-like neutron energy spectrum for 14 MeV neutron irradiation has been constructed. Taking into account its performance and ability of almost incessant operation the facility appears as one of the most powerful 14 MeV neutron sources. The facility shall be used for material research connected with thermonuclear devices (ITER) and 4. generation nuclear reactors. The system of independent fuels channels used in MARIA reactor appear to be very flexible and very convenient to be used as irradiation channels for uranium targets for {sup 99}Mo production. Currently, MARIA reactor supplies ca. 18% world production of {sup 99}Mo. The MARIA reactor research activities are still extended. The current scientific projects are connected e.g. with silicon neutron transmutation doping, in-pile gamma heating measurements, French calculation codes implementation (TRIPOLI4, APOLLO2). The horizontal neutron beams utilization is also developed. The MARIA reactor, due to its primary application connected with loop and fuel testing, is very convenient for testing the nuclear instrumentation, control and measurement systems.« less

Experimental demonstration of free-space optical vortex transmutation with polygonal lenses.

PubMed

Gao, Nan; Xie, Changqing

2012-08-01

Vortex transmutation was predicted to take place when vortices interact with systems possessing discrete rotational symmetries of finite order [Phys. Rev. Lett.95, 123901 (2005)]. Here we report what is believed to be the first experimental demonstration of vortex transmutation. We show that in free space, by simply inserting polygonal lenses into the optical path, the central vorticity of a coaxially incident optical vortex can be changed following the modular transmutation rule. We generate the wavefront at the exit face of the lenses with computer generated holograms and measure the output vorticity using the interference patterns at the focal plane. The results agree well with theoretical predictions.

Infrared absorption study of neutron-transmutation-doped germanium

NASA Technical Reports Server (NTRS)

Park, I. S.; Haller, E. E.

1988-01-01

Using high-resolution far-infrared Fourier transform absorption spectroscopy and Hall effect measurements, the evolution of the shallow acceptor and donor impurity levels in germanium during and after the neutron transmutation doping process was studied. The results show unambiguously that the gallium acceptor level concentration equals the concentration of transmutated Ge-70 atoms during the whole process indicating that neither recoil during transmutation nor gallium-defect complex formation play significant roles. The arsenic donor levels appear at full concentration only after annealing for 1 h at 450 C. It is shown that this is due to donor-radiation-defect complex formation. Again, recoil does not play a significant role.

Method to Reduce Long-lived Fission Products by Nuclear Transmutations with Fast Spectrum Reactors.

PubMed

Chiba, Satoshi; Wakabayashi, Toshio; Tachi, Yoshiaki; Takaki, Naoyuki; Terashima, Atsunori; Okumura, Shin; Yoshida, Tadashi

2017-10-24

Transmutation of long-lived fission products (LLFPs: 79 Se, 93 Zr, 99 Tc, 107 Pd, 129 I, and 135 Cs) into short-lived or non-radioactive nuclides by fast neutron spectrum reactors without isotope separation has been proposed as a solution to the problem of radioactive wastes disposal. Despite investigation of many methods, such transmutation remains technologically difficult. To establish an effective and efficient transmutation system, we propose a novel neutron moderator material, yttrium deuteride (YD 2 ), to soften the neutron spectrum leaking from the reactor core. Neutron energy spectra and effective half-lives of LLFPs, transmutation rates, and support ratios were evaluated with the continuous-energy Monte Carlo code MVP-II/MVP-BURN and the JENDL-4.0 cross section library. With the YD 2 moderator in the radial blanket and shield regions, effective half-lives drastically decreased from 106 to 102 years and the support ratios reached 1.0 for all six LLFPs. This successful development and implementation of a transmutation system for LLFPs without isotope separation contributes to a the ability of fast spectrum reactors to reduce radioactive waste by consuming their own LLFPs.

Nuclear transmutation in steels

NASA Astrophysics Data System (ADS)

Belozerova, A. R.; Shimanskii, G. A.; Belozerov, S. V.

2009-05-01

The investigations of the effects of nuclear transmutation in steels that are widely used in nuclear power and research reactors and in steels that are planned for the application in thermonuclear fusion plants, which are employed under the conditions of a prolonged action of neutron irradiation with different spectra, made it possible to study the effects of changes in the isotopic and chemical composition on the tendency of changes in the structural stability of these steels. For the computations of nuclear transmutation in steels, we used a program complex we have previously developed on the basis of algorithms for constructing branched block-type diagrams of nuclide transformations and for locally and globally optimizing these diagrams with the purpose of minimizing systematic errors in the calculation of nuclear transmutation. The dependences obtained were applied onto a Schaeffler diagram for steels used for structural elements of reactors. For the irradiation in fission reactors, we observed only a weak influence of the effects of nuclear transmutation in steels on their structural stability. On the contrary, in the case of irradiation with fusion neutrons, a strong influence of the effects of nuclear transmutation in steels on their structural stability has been noted.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Nordborg, C.

A new improved version of the OECD Nuclear Energy Agency (NEA) co-ordinated Joint Evaluated Fission and Fusion (JEFF) data library, JEFF-3.1, was released in May 2005. It comprises a general purpose library and the following five special purpose libraries: activation; thermal scattering law; radioactive decay; fission yield; and proton library. The objective of the previous version of the library (JEFF-2.2) was to achieve improved performance for existing reactors and fuel cycles. In addition to this objective, the JEFF-3.1 library aims to provide users with data for a wider range of applications. These include innovative reactor concepts, transmutation of radioactive waste,more » fusion, and various other energy and non-energy related industrial applications. Initial benchmark testing has confirmed the expected very good performance of the JEFF-3.1 library. Additional benchmarking of the libraries is underway, both for the general purpose and for the special purpose libraries. A new three-year mandate to continue developing the JEFF library was recently granted by the NEA. For the next version of the library, JEFF-3.2, it is foreseen to put more effort into fission product and minor actinide evaluations, as well as the inclusion of more covariance data. (authors)« less

Measurement of 235U(n,n'γ) and 235U(n,2nγ) reaction cross sections

NASA Astrophysics Data System (ADS)

Kerveno, M.; Thiry, J. C.; Bacquias, A.; Borcea, C.; Dessagne, P.; Drohé, J. C.; Goriely, S.; Hilaire, S.; Jericha, E.; Karam, H.; Negret, A.; Pavlik, A.; Plompen, A. J. M.; Romain, P.; Rouki, C.; Rudolf, G.; Stanoiu, M.

2013-02-01

The design of generation IV nuclear reactors and the studies of new fuel cycles require knowledge of the cross sections of various nuclear reactions. Our research is focused on (n,xnγ) reactions occurring in these new reactors. The aim is to measure unknown cross sections and to reduce the uncertainty on present data for reactions and isotopes of interest for transmutation or advanced reactors. The present work studies the 235U(n,n'γ) and 235U(n,2nγ) reactions in the fast neutron energy domain (up to 20 MeV). The experiments were performed with the Geel electron linear accelerator GELINA, which delivers a pulsed white neutron beam. The time characteristics enable measuring neutron energies with the time-of-flight (TOF) technique. The neutron induced reactions [in this case inelastic scattering and (n,2n) reactions] are identified by on-line prompt γ spectroscopy with an experimental setup including four high-purity germanium (HPGe) detectors. A fission ionization chamber is used to monitor the incident neutron flux. The experimental setup and analysis methods are presented and the model calculations performed with the TALYS-1.2 code are discussed.

Selective Separation of Trivalent Actinides from Lanthanides by Aqueous Processing with Introduction of Soft Donor Atoms

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kenneth L. Nash

2009-09-22

Implementation of a closed loop nuclear fuel cycle requires the utilization of Pu-containing MOX fuels with the important side effect of increased production of the transplutonium actinides, most importantly isotopes of Am and Cm. Because the presence of these isotopes significantly impacts the long-term radiotoxicity of high level waste, it is important that effective methods for their isolation and/or transmutation be developed. Furthermore, since transmutation is most efficiently done in the absence of lanthanide fission products (high yield species with large thermal neutron absorption cross sections) it is important to have efficient procedures for the mutual separation of Am andmore » Cm from the lanthanides. The chemistries of these elements are nearly identical, differing only in the slightly stronger strength of interaction of trivalent actinides with ligand donor atoms softer than O (N, Cl-, S). Research being conducted around the world has led to the development of new reagents and processes with considerable potential for this task. However, pilot scale testing of these reagents and processes has demonstrated the susceptibility of the new classes of reagents to radiolytic and hydrolytic degradation. In this project, separations of trivalent actinides from fission product lanthanides have been investigated in studies of 1) the extraction and chemical stability properties of a class of soft-donor extractants that are adapted from water-soluble analogs, 2) the application of water soluble soft-donor complexing agents in tandem with conventional extractant molecules emphasizing fundamental studies of the TALSPEAK Process. This research was conducted principally in radiochemistry laboratories at Washington State University. Collaborators at the Radiological Processing Laboratory (RPL) at the Pacific Northwest National Laboratory (PNNL) have contributed their unique facilities and capabilities, and have supported student internships at PNNL to broaden their academic experience. New information has been developed to qualify the extraction potential of a class of pyridine-functionalized tetraaza complexants indicating potential single contact Am-Nd separation factors of about 40. The methodology developed for characterization will find further application in our continuing efforts to synthesize and characterize new reagents for this separation. Significant new insights into the performance envelope and supporting information on the TALSPEAK process has also been developed.« less

Electrochemical/Pyrometallurgical Waste Stream Processing and Waste Form Fabrication

DOE Office of Scientific and Technical Information (OSTI.GOV)

Steven Frank; Hwan Seo Park; Yung Zun Cho

This report summarizes treatment and waste form options being evaluated for waste streams resulting from the electrochemical/pyrometallurgical (pyro ) processing of used oxide nuclear fuel. The technologies that are described are South Korean (Republic of Korea – ROK) and United States of America (US) ‘centric’ in the approach to treating pyroprocessing wastes and are based on the decade long collaborations between US and ROK researchers. Some of the general and advanced technologies described in this report will be demonstrated during the Integrated Recycle Test (IRT) to be conducted as a part of the Joint Fuel Cycle Study (JFCS) collaboration betweenmore » US Department of Energy (DOE) and ROK national laboratories. The JFCS means to specifically address and evaluated the technological, economic, and safe guard issues associated with the treatment of used nuclear fuel by pyroprocessing. The IRT will involve the processing of commercial, used oxide fuel to recover uranium and transuranics. The recovered transuranics will then be fabricated into metallic fuel and irradiated to transmutate, or burn the transuranic elements to shorter lived radionuclides. In addition, the various process streams will be evaluated and tested for fission product removal, electrolytic salt recycle, minimization of actinide loss to waste streams and waste form fabrication and characterization. This report specifically addresses the production and testing of those waste forms to demonstrate their compatibility with treatment options and suitability for disposal.« less

Th and U fuel photofission study by NTD for AD-MSR subcritical assembly

NASA Astrophysics Data System (ADS)

Sajo-Bohus, Laszlo; Greaves, Eduardo D.; Davila, Jesus; Barros, Haydn; Pino, Felix; Barrera, Maria T.; Farina, Fulvio

2015-07-01

During the last decade a considerable effort has been devoted for developing energy generating systems based on advanced nuclear technology within the design concepts of GEN-IV. Thorium base fuel systems such as accelerator driven nuclear reactors are one of the often mentioned attractive and affordable options. Several radiotherapy linear accelerators are on the market and due to their reliability, they could be employed as drivers for subcritical liquid fuel assemblies. Bremsstrahlung photons with energies above 5.5MeV, induce (γ,n) and (e,e'n) reactions in the W-target. Resulting gamma radiation and photo or fission neutrons may be absorbed in target materials such as thorium and uranium isotopes to induce sustained fission or nuclear transmutation in waste radioactive materials. Relevant photo driven and photo-fission reaction cross sections are important for actinides 232Th, 238U and 237Np in the radiotherapy machines energy range of 10-20 MV. In this study we employ passive nuclear track detectors (NTD) to determine fission rates and neutron production rates with the aim to establish the feasibility for gamma and photo-neutron driven subcritical assemblies. To cope with these objectives a 20 MV radiotherapy machine has been employed with a mixed fuel target. Results will support further development for a subcritical assembly employing a thorium containing liquid fuel. It is expected that acquired technological knowledge will contribute to the Venezuelan nuclear energy program.

An investigation into the feasibility of thorium fuels utilization in seed-blanket configurations for TRIGA PUSPATI Reactor (RTP)

NASA Astrophysics Data System (ADS)

Damahuri, Abdul Hannan Bin; Mohamed, Hassan; Aziz Mohamed, Abdul; Idris, Faridah

2018-01-01

Thorium is one of the elements that needs to be explored for nuclear fuel research and development. One of the popular core configurations of thorium fuel is seed-blanket configuration or also known as Radkowsky Thorium Fuel concept. The seed will act as a supplier of neutrons, which will be placed inside of the core. The blanket, on the other hand, is the consumer of neutrons that is located at outermost of the core. In this work, a neutronic analysis of seed-blanket configuration for the TRIGA PUSPATI Reactor (RTP) is carried out using Monte Carlo method. The reactor, which has been operated since 1982 use uranium zirconium hydride (U-ZrH1.6) as the fuel and have multiple uranium weight which are 8.5, 12 and 20 wt.%. The pool type reactor is one and only research reactor that located in Malaysia. The design of core included the Uranium Zirconium Hydride located at the centre of the core that will act as the seed to supply neutron. The thorium oxide that will act as blanket situated outside of seed region will receive neutron to transmute 232Th to 233U. The neutron multiplication factor or criticality of each configuration is estimated. Results show that the highest initial criticality achieved is 1.30153.

Determination of Trace Concentration in TMD Detectors using PGAA

NASA Astrophysics Data System (ADS)

Tomandl, I.; Viererbl, L.; Kudějová, P.; Lahodová, Z.; Klupák, V.; Fikrle, M.

2015-05-01

Transmutation detectors could be alternative to the traditional activation detector method for neutron fluence dosimetry at power nuclear reactors. This new method require an isotopically highly-sensitive, non-destructive in sense of compactness as well as isotopic content, precise and standardly used analytical method for trace concentration determination. The capability of Prompt Gamma-ray Activation Analysis (PGAA) for determination of trace concentrations of transmuted stable nuclides in the metallic foils of Ni, Au, Cu and Nb, which were irradiated for 21 days in the reactor core at the LVR-15 research reactor in Řež, is reported. The PGAA measurements of these activation foils were performed at the PGAA facility at Forschungs-Neutronenquelle Heinz Maier-Leibnitz (FRMII) in Garching.

Monitoring of the Irradiated Neutron Fluence in the Neutron Transmutation Doping Process of Hanaro

NASA Astrophysics Data System (ADS)

Kim, Myong-Seop; Park, Sang-Jun

2009-08-01

Neutron transmutation doping (NTD) for silicon is a process of the creation of phosphorus impurities in intrinsic or extrinsic silicon by neutron irradiation to obtain silicon semiconductors with extremely uniform dopant distribution. HANARO has two vertical holes for the NTD, and the irradiation for 5 and 6 inch silicon ingots has been going on at one hole. In order to achieve the accurate neutron fluence corresponding to the target resistivity, the real time neutron flux is monitored by self-powered neutron detectors. After irradiation, the total irradiation fluence is confirmed by measuring the absolute activity of activation detectors. In this work, a neutron fluence monitoring method using zirconium foils with the mass of 10 ~ 50 mg was applied to the NTD process of HANARO. We determined the proportional constant of the relationship between the resistivity of the irradiated silicon and the neutron fluence determined by using zirconium foils. The determined constant for the initially n-type silicon was 3.126 × 1019 n·Ω/cm. It was confirmed that the difference between this empirical value and the theoretical one was only 0.5%. Conclusively, the practical methodology to perform the neutron transmutation doping of silicon was established.

Energy Production and Transmutation of Nuclear Waste by Accelerator Driven Systems

NASA Astrophysics Data System (ADS)

Zhivkov, P. K.

2018-05-01

There is a significant amount of highly radiotoxic long-life nuclear waste (NW) produced by NPP (Nuclear Power Plants). Transmutation is a process which transforms NW into less radiotoxic nuclides with a shorter period of half-life by spallation neutrons or radiative capture of neutrons produced by ADS (Accelerator Driven System). In the processes of transmutation new radioactive nuclides are produced. ADS is big energy consumer equipment. It is a method for production of a high-flux and high-energy neutron field. All these processes occur in ADS simultaneously. ADS is able to transmute actinides and produce energy simultaneously. The article considers the energy production problems in ADS. Several ideas are developed regarding the solution of the global energy supply.

Fuel cycle for a fusion neutron source

NASA Astrophysics Data System (ADS)

Ananyev, S. S.; Spitsyn, A. V.; Kuteev, B. V.

2015-12-01

The concept of a tokamak-based stationary fusion neutron source (FNS) for scientific research (neutron diffraction, etc.), tests of structural materials for future fusion reactors, nuclear waste transmutation, fission reactor fuel production, and control of subcritical nuclear systems (fusion-fission hybrid reactor) is being developed in Russia. The fuel cycle system is one of the most important systems of FNS that provides circulation and reprocessing of the deuterium-tritium fuel mixture in all fusion reactor systems: the vacuum chamber, neutral injection system, cryogenic pumps, tritium purification system, separation system, storage system, and tritium-breeding blanket. The existing technologies need to be significantly upgraded since the engineering solutions adopted in the ITER project can be only partially used in the FNS (considering the capacity factor higher than 0.3, tritium flow up to 200 m3Pa/s, and temperature of reactor elements up to 650°C). The deuterium-tritium fuel cycle of the stationary FNS is considered. The TC-FNS computer code developed for estimating the tritium distribution in the systems of FNS is described. The code calculates tritium flows and inventory in tokamak systems (vacuum chamber, cryogenic pumps, neutral injection system, fuel mixture purification system, isotope separation system, tritium storage system) and takes into account tritium loss in the fuel cycle due to thermonuclear burnup and β decay. For the two facility versions considered, FNS-ST and DEMO-FNS, the amount of fuel mixture needed for uninterrupted operation of all fuel cycle systems is 0.9 and 1.4 kg, consequently, and the tritium consumption is 0.3 and 1.8 kg per year, including 35 and 55 g/yr, respectively, due to tritium decay.

Inert matrix fuel neutronic, thermal-hydraulic, and transient behavior in a light water reactor

NASA Astrophysics Data System (ADS)

Carmack, W. J.; Todosow, M.; Meyer, M. K.; Pasamehmetoglu, K. O.

2006-06-01

Currently, commercial power reactors in the United States operate on a once-through or open cycle, with the spent nuclear fuel eventually destined for long-term storage in a geologic repository. Since the fissile and transuranic (TRU) elements in the spent nuclear fuel present a proliferation risk, limit the repository capacity, and are the major contributors to the long-term toxicity and dose from the repository, methods and systems are needed to reduce the amount of TRU that will eventually require long-term storage. An option to achieve a reduction in the amount, and modify the isotopic composition of TRU requiring geological disposal is 'burning' the TRU in commercial light water reactors (LWRs) and/or fast reactors. Fuel forms under consideration for TRU destruction in light water reactors (LWRs) include mixed-oxide (MOX), advanced mixed-oxide, and inert matrix fuels. Fertile-free inert matrix fuel (IMF) has been proposed for use in many forms and studied by several researchers. IMF offers several advantages relative to MOX, principally it provides a means for reducing the TRU in the fuel cycle by burning the fissile isotopes and transmuting the minor actinides while producing no new TRU elements from fertile isotopes. This paper will present and discuss the results of a four-bundle, neutronic, thermal-hydraulic, and transient analyses of proposed inert matrix materials in comparison with the results of similar analyses for reference UOX fuel bundles. The results of this work are to be used for screening purposes to identify the general feasibility of utilizing specific inert matrix fuel compositions in existing and future light water reactors. Compositions identified as feasible using the results of these analyses still require further detailed neutronic, thermal-hydraulic, and transient analysis study coupled with rigorous experimental testing and qualification.

Joint tests at INL and CEA of a transient hot wire needle probe for in-pile thermal conductivity measurement

DOE Office of Scientific and Technical Information (OSTI.GOV)

Daw, J.E.; Knudson, D.L.; Villard, J.F.

2015-07-01

Thermal conductivity is a key property that must be known for proper design, testing, and deployment of new fuels and structural materials in nuclear reactors. Thermal conductivity is highly dependent on the physical structure, chemical composition, and the state of the material. Typically, thermal conductivity changes that occur during irradiation are currently measured out-of-pile using a 'cook and look' approach. But repeatedly removing samples from a test reactor to make measurements is expensive, has the potential to disturb phenomena of interest, and only provides understanding of the sample's end state when each measurement is made. There are also limited thermo-physicalmore » property data available for advanced fuels; and such data are needed for simulation codes, the development of next generation reactors, and advanced fuels for existing nuclear plants. Being able to quickly characterize fuel thermal conductivity during irradiation can improve the fidelity of data, reduce costs of post-irradiation examinations, increase understanding of how fuels behave under irradiation, and confirm or improve existing thermal conductivity measurement techniques. This paper discusses efforts to develop and evaluate an innovative in-pile thermal conductivity sensor based on the transient hot wire thermal conductivity method (THWM), using a single needle probe (NP) containing a line heat source and thermocouple embedded in the fuel. The sensor that has been designed and manufactured by the Idaho National Laboratory (INL) includes a unique combination of materials, geometry, and fabrication techniques that make the hot wire method suitable for in-pile applications. In particular, efforts were made to minimize the influence of the sensor and maximize fuel hot-wire heating. The probe has a thermocouple-like construction with high temperature resistant materials that remain ductile while resisting transmutation and materials interactions. THWM-NP prototypes were fabricated for both room temperature proof-of-concept evaluations and high temperature testing. Evaluations have been performed jointly by the INL and the French Alternative Energies and Atomic Energy Commission (CEA), both in Idaho Falls (USA) and in Cadarache (France), in the framework of a collaborative program for instrumentation of Material Testing Reactors. Initial tests were conducted on samples with a large range of thermal conductivities and temperatures ranging from 20 deg. C to 600 deg. C. Particularly, tests were recently performed on a sample having thermal conductivity and dimensions similar to UO{sub 2} and MOX nuclear fuels, in order to validate the ability of this sensor to operate for in-pile characterization of Light Water Reactors fuels. The results of the tests already completed at INL and CEA indicate that the Transient Hot Wire Needle Probe offers an enhanced method for in-pile detection of thermal conductivity. (authors)« less

Characterization of Neutron-Induced Defects in Isotopically Enriched Lithium Tetraborate

DTIC Science & Technology

2011-03-01

that efficiently captures and transmutes neutrons into more readily detected forms of material or energy. Neutron detection is desirable to detect...be used to transmute neutrons into a more readily detectable particle or energy. Upon absorbing a thermal neutron, 6Li undergoes the reaction, 6 1...both 6Li and 10B in natural abundances unless deliberately enriched. In addition to the direct reactions, 6Li or 7Li and 16O can transmute neutrons

The Soviet Central Asian Challenge: A Neo-Gramscian Analysis.

DTIC Science & Technology

1986-09-01

transmutated into the Soviet Union. This point is fundamental to understanding why the Russians are the ruling nationality group in the Soviet Union. The Great...initial years, force and coercion were instrumental for ensuring the continued existence of the transmuted Russian Empire. The new Soviet Union also...information on .Muslim national communism s1 l (Reft. 31, i33. 26F1or an excellent article on Russian nationalism’s transmutation to Soviet communism and the

Synthesis and characterisation of PuPO4 - a potential analytical standard for EPMA actinide quantification

NASA Astrophysics Data System (ADS)

Wright, K. E.; Popa, K.; Pöml, P.

2018-01-01

Transmutation nuclear fuels contain weight percentage quantities of actinide elements, including Pu, Am and Np. Because of the complex spectra presented by actinide elements using electron probe microanalysis (EPMA), it is necessary to have relatively pure actinide element standards to facilitate overlap correction and accurate quantitation. Synthesis of actinide oxide standards is complicated by their multiple oxidation states, which can result in inhomogeneous standards or standards that are not stable at atmospheric conditions. Synthesis of PuP4 results in a specimen that exhibits stable oxidation-reduction chemistry and is sufficiently homogenous to serve as an EPMA standard. This approach shows promise as a method for producing viable actinide standards for microanalysis.

The End of Alchemy? The Repudiation and Persistence of Chrysopoeia at the Académie Royale des Sciences in the Eighteenth Century.

PubMed

Principe, Lawrence M

2014-01-01

The general abandonment of serious endeavor toward metallic transmutation represents a major development in the history of chemistry, yet its exact causes and timing remain unclear. This essay examines the fate of chrysopoeia at the eighteenth-century Académie Royale des Sciences. It reveals a long-standing tension between Académie chemists, who pursued transmutation, and administrators, who tried to suppress it. This tension provides background for Etienne-François Geoffroy's 1722 paper describing fraudulent practices around transmutation. Although transmutation seems to disappear after Geoffroy's paper, manuscripts reveal that most of the institution's chemists continued to pursue it privately until at least the 1760s, long after widely accepted dates for the "demise of alchemy" in learned circles.

DOE Office of Scientific and Technical Information (OSTI.GOV)

McMurray, Jacob W.; Lindemer, Terrence B.; Brown, Nicholas R.

There are three important failure mechanisms that must be controlled in high-temperature gas-cooled reactor (HTGR) fuel for certain higher burnup applications are SiC layer rupture, SiC corrosion by CO, and coating compromise from kernel migration. All are related to high CO pressures stemming from free O generated when uranium present as UO 2 fissions and the O is not subsequently bound by other elements. Furthermore, in the HTGR UCO kernel design, CO buildup from excess O is controlled by the inclusion of additional uranium in the form of a carbide, UC x. An approach for determining the minimum UC xmore » content to ensure negligible CO formation was developed and demonstrated using CALPHAD models and the Serpent 2 reactor physics and depletion analysis tool. Our results are intended to be more accurate than previous estimates by including more nuclear and chemical factors, in particular the effect of transmutation products on the oxygen distribution as the fuel kernel composition evolves with burnup.« less

Transmuted of Rayleigh Distribution with Estimation and Application on Noise Signal

NASA Astrophysics Data System (ADS)

Ahmed, Suhad; Qasim, Zainab

2018-05-01

This paper deals with transforming one parameter Rayleigh distribution, into transmuted probability distribution through introducing a new parameter (λ), since this studied distribution is necessary in representing signal data distribution and failure data model the value of this transmuted parameter |λ| ≤ 1, is also estimated as well as the original parameter (⊖) by methods of moments and maximum likelihood using different sample size (n=25, 50, 75, 100) and comparing the results of estimation by statistical measure (mean square error, MSE).

A brief history of design studies on innovative nuclear reactors

NASA Astrophysics Data System (ADS)

Sekimoto, Hiroshi

2014-09-01

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.

Resolving and quantifying overlapped chromatographic bands by transmutation

PubMed

Malinowski

2000-09-15

A new chemometric technique called "transmutation" is developed for the purpose of sharpening overlapped chromatographic bands in order to quantify the components. The "transmutation function" is created from the chromatogram of the pure component of interest, obtained from the same instrument, operating under the same experimental conditions used to record the unresolved chromatogram of the sample mixture. The method is used to quantify mixtures containing toluene, ethylbenzene, m-xylene, naphthalene, and biphenyl from unresolved chromatograms previously reported. The results are compared to those obtained using window factor analysis, rank annihilation factor analysis, and matrix regression analysis. Unlike the latter methods, the transmutation method is not restricted to two-dimensional arrays of data, such as those obtained from HPLC/DAD, but is also applicable to chromatograms obtained from single detector experiments. Limitations of the method are discussed.

Transmutation: The Roots of the Dream.

ERIC Educational Resources Information Center

Karpenko, Vladimir

1995-01-01

Examines the history of alchemical attempts at transmutation and classifies them by differing approaches and techniques. Traces the development of alchemy in Asia, Europe, and the Middle East, and compares alchemy with craftsmanship. (18 references) (DDR)

Electronic Transmutation (ET): Chemically Turning One Element into Another.

PubMed

Zhang, Xinxing; Lundell, Katie A; Olson, Jared K; Bowen, Kit H; Boldyrev, Alexander I

2018-03-08

The concept of electronic transmutation (ET) depicts the processes that by acquiring an extra electron, an element with the atomic number Z begins to have properties that were known to only belong to its neighboring element with the atomic number Z+1. Based on ET, signature compounds and chemical bonds that are composed of certain elements can now be designed and formed by other electronically transmutated elements. This Minireview summarizes the recent developments and applications of ET on both the theoretical and experimental fronts. Examples on the ET of Group 13 elements into Group 14 elements, Group 14 elements into Group 15 elements, and Group 15 elements into Group 16 elements are discussed. Compounds and chemical bonding composed of carbon, silicon, germanium, phosphorous, oxygen and sulfur now have analogues using transmutated boron, aluminum, gallium, silicon, nitrogen, and phosphorous. © 2018 Wiley-VCH Verlag GmbH & Co. KGaA, Weinheim.

Transmutation studies at CEA in frame of the SPIN program objectives, results and future trends

DOE Office of Scientific and Technical Information (OSTI.GOV)

Salvatores, M.; Prunier, C.; Guerin, Y.

1995-10-01

In order to respond to the public concern about wastes and in particular the long-lived high level ones, a French law issued on December 30, 1991 identified the major objectives of research for the next fifteen years, before a new debate and possibly a decision on final wastes disposal in Parliament. These objectives are: (1) improvement of the wastes conditioning; (2) extraction and transmutation of the long-lived wastes in order to minimize their long term toxicity; (3) research performed in underground laboratories in order to characterize the capacity of geological structures to confine radioactive wastes (two sites have to bemore » selected for these underground laboratories, in concertation with the local population); (4) last, the study of conditioning and prolonged surface storage of wastes.« less

"Can Simple Metals Be Transmuted into Gold?" Teaching Science through a Historical Approach.

ERIC Educational Resources Information Center

Mamlok, Rachel; Ben-Zvi, Ruth; Menis, Joseph; Penick, John E.

2000-01-01

Describes the development and enactment of a new teaching unit, "Can simple metals be transmuted into gold?", through an historical approach as well as teacher preparation to teach this unit. (Contains 16 references.) (ASK)

CY2013 Annual Report for DOE-ITU INERI 2010-006-E

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kennedy, J. Rory; Rondinella, Vincenzo V.

2014-12-01

New concepts for nuclear energy development are considered in both the USA and Europe within the framework of the Generation-IV International Forum (GIF) as well as in various US-DOE programs (e.g. the Fuel Cycle Research and Development - FCRD) and as part of the European Sustainable Nuclear Energy Technology Platform (SNE-TP). Since most new fuel cycle concepts envisage the adoption of a closed nuclear fuel cycle employing fast reactors, the fuel behavior characteristics of the various proposed advanced fuel forms must be effectively investigated using state of the art experimental techniques before implementation. More rapid progress can be achieved ifmore » effective synergy with advanced (multi-scale) modeling efforts can be achieved. The fuel systems to be considered include minor actinide (MA) transmutation fuel types such as advanced MOX, advanced metal alloy, inert matrix fuel (IMF), and other ceramic fuels like nitrides, carbides, etc., for fast neutronic spectrum conditions. Most of the advanced fuel compounds have already been the object of past examination programs, which included irradiations in research reactors. The knowledge derived from previous experience constitutes a significant, albeit incomplete body of data. New or upgraded experimental tools are available today that can extend the scientific and technological knowledge towards achieving the objectives associated with the new generation of nuclear reactors and fuels. The objectives of this project will be three-fold: (1) to extend the available knowledge on properties and irradiation behavior of high burnup and minor actinide bearing advanced fuel systems; (2) to establish a synergy with multi-scale and code development efforts in which experimental data and expertise on the irradiation behavior of nuclear fuels is properly conveyed for the upgrade/development of advanced modeling tools; (3) to promote the effective use of international resources to the characterization of irradiated fuel through exchange of expertise and information among leading experimental facilities. The priorities in this project will be set according to the down selection procedure of U.S. and European development programs.« less

Method and apparatus for transmutation of atomic nuclei

DOEpatents

Maenchen, John Eric; Ruiz, Carlos Leon

1998-01-01

Insuring a constant supply of radioisotopes is of great importance to medicine and industry. This invention addresses this problem, and helps to solve it by introducing a new apparatus for transmutation of isotopes which enables swift and flexible production on demand.

Method and apparatus for transmutation of atomic nuclei

DOEpatents

Maenchen, J.E.; Ruiz, C.L.

1998-12-08

Insuring a constant supply of radioisotopes is of great importance to medicine and industry. This invention addresses this problem, and helps to solve it by introducing a new apparatus for transmutation of isotopes which enables swift and flexible production on demand. 9 figs.

Method and apparatus for transmutation of atomic nuclei

DOEpatents

Maenchen, J.E.; Ruiz, C.L.

1998-06-09

Insuring a constant supply of radioisotopes is of great importance to medicine and industry. This invention addresses this problem, and helps to solve it by introducing a new apparatus for transmutation of isotopes which enables swift and flexible production on demand. 9 figs.

Reflection, transmutation, annihilation, and resonance in two-component kink collisions

NASA Astrophysics Data System (ADS)

Alonso-Izquierdo, A.

2018-02-01

In this paper, the study of collisions between kinks arising in the family of MSTB models is addressed. Phenomena such as elastic kink reflection, mutual annihilation, kink-antikink transmutation and inelastic reflection are found and depend on the impact velocity.

Santa Muerte: Threatening the U.S. Homeland

DTIC Science & Technology

2011-03-08

7 Religious Transmutation ...Religious Transmutation Similarly, a small number of anthropologists do not support the notorious claims about Santa Muerte. They believe that...Muerte, it is no wonder that the religion is spreading globally. Random Terror Slave holding, sexual activity with minors, kidnapping, and

Copper doping of ZnO crystals by transmutation of 64Zn to 65Cu: An electron paramagnetic resonance and gamma spectroscopy study

NASA Astrophysics Data System (ADS)

Recker, M. C.; McClory, J. W.; Holston, M. S.; Golden, E. M.; Giles, N. C.; Halliburton, L. E.

2014-06-01

Transmutation of 64Zn to 65Cu has been observed in a ZnO crystal irradiated with neutrons. The crystal was characterized with electron paramagnetic resonance (EPR) before and after the irradiation and with gamma spectroscopy after the irradiation. Major features in the gamma spectrum of the neutron-irradiated crystal included the primary 1115.5 keV gamma ray from the 65Zn decay and the positron annihilation peak at 511 keV. Their presence confirmed the successful transmutation of 64Zn nuclei to 65Cu. Additional direct evidence for transmutation was obtained from the EPR of Cu2+ ions (where 63Cu and 65Cu hyperfine lines are easily resolved). A spectrum from isolated Cu2+ (3d9) ions acquired after the neutron irradiation showed only hyperfine lines from 65Cu nuclei. The absence of 63Cu lines in this Cu2+ spectrum left no doubt that the observed 65Cu signals were due to transmuted 65Cu nuclei created as a result of the neutron irradiation. Small concentrations of copper, in the form of Cu+-H complexes, were inadvertently present in our as-grown ZnO crystal. These Cu+-H complexes are not affected by the neutron irradiation, but they dissociate when a crystal is heated to 900 °C. This behavior allowed EPR to distinguish between the copper initially in the crystal and the copper subsequently produced by the neutron irradiation. In addition to transmutation, a second major effect of the neutron irradiation was the formation of zinc and oxygen vacancies by displacement. These vacancies were observed with EPR.

A beamline systems model for Accelerator-Driven Transmutation Technology (ADTT) facilities

DOE Office of Scientific and Technical Information (OSTI.GOV)

Todd, A.M.M.; Paulson, C.C.; Peacock, M.A.

1995-10-01

A beamline systems code, that is being developed for Accelerator-Driven Transmutation Technology (ADTT) facility trade studies, is described. The overall program is a joint Grumman, G.H. Gillespie Associates (GHGA) and Los Alamos National Laboratory effort. The GHGA Accelerator Systems Model (ASM) has been adopted as the framework on which this effort is based. Relevant accelerator and beam transport models from earlier Grumman systems codes are being adapted to this framework. Preliminary physics and engineering models for each ADTT beamline component have been constructed. Examples noted include a Bridge Coupled Drift Tube Linac (BCDTL) and the accelerator thermal system. A decisionmore » has been made to confine the ASM framework principally to beamline modeling, while detailed target/blanket, balance-of-plant and facility costing analysis will be performed externally. An interfacing external balance-of-plant and facility costing model, which will permit the performance of iterative facility trade studies, is under separate development. An ABC (Accelerator Based Conversion) example is used to highlight the present models and capabilities.« less

A beamline systems model for Accelerator-Driven Transmutation Technology (ADTT) facilities

DOE Office of Scientific and Technical Information (OSTI.GOV)

Todd, Alan M. M.; Paulson, C. C.; Peacock, M. A.

1995-09-15

A beamline systems code, that is being developed for Accelerator-Driven Transmutation Technology (ADTT) facility trade studies, is described. The overall program is a joint Grumman, G. H. Gillespie Associates (GHGA) and Los Alamos National Laboratory effort. The GHGA Accelerator Systems Model (ASM) has been adopted as the framework on which this effort is based. Relevant accelerator and beam transport models from earlier Grumman systems codes are being adapted to this framework. Preliminary physics and engineering models for each ADTT beamline component have been constructed. Examples noted include a Bridge Coupled Drift Tube Linac (BCDTL) and the accelerator thermal system. Amore » decision has been made to confine the ASM framework principally to beamline modeling, while detailed target/blanket, balance-of-plant and facility costing analysis will be performed externally. An interfacing external balance-of-plant and facility costing model, which will permit the performance of iterative facility trade studies, is under separate development. An ABC (Accelerator Based Conversion) example is used to highlight the present models and capabilities.« less

Separations in the STATS report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Choppin, G.R.

1996-12-31

The Separations Technology and Transmutation Systems (STATS) Committee formed a Subcommittee on Separations. This subcommittee was charged with evaluating the separations proposed for the several reactor and accelerator transmutation systems. It was also asked to review the processing options for the safe management of high-level waste generated by the defense programs, in particular, the special problems involved in dealing with the waste at the U.S. Department of Energy (DOE) facility in Hanford, Washington. Based on the evaluations from the Subcommittee on Separations, the STATS Committee concluded that for the reactor transmutation programs, aqueous separations involving a combination of PUREX andmore » TRUEX solvent extraction processes could be used. However, additional research and development (R&D) would be required before full plant-scale use of the TRUEX technology could be employed. Alternate separations technology for the reactor transmutation program involves pyroprocessing. This process would require a significant amount of R&D before its full-scale application can be evaluated.« less

Exploring Chemical Space with the Alchemical Derivatives.

PubMed

Balawender, Robert; Welearegay, Meressa A; Lesiuk, Michał; De Proft, Frank; Geerlings, Paul

2013-12-10

In this paper, we verify the usefulness of the alchemical derivatives in the prediction of chemical properties. We concentrate on the stability of the transmutation products, where the term "transmutation" means the change of the nuclear charge at an atomic site at constant number of electrons. As illustrative transmutations showing the potential of the method in exploring chemical space, we present some examples of increasing complexity starting with the deprotonation, continuing with the transmutation of the nitrogen molecule, and ending with the substitution of isoelectronic B-N units for C-C units and N units for C-H units in carbocyclic systems. The basis set influence on the qualitative and quantitative accuracies of the alchemical predictions was investigated. The alchemical deprotonation energy (from the second order Taylor expansion) correlates well with the vertical deprotonation energy and can be used as a preliminary indicator for the experimental deprotonation energy. The results of calculations for the BN derivatives of benzene and pyrene show that this method has great potential for efficient and accurate scanning of chemical space.

The Fast-spectrum Transmutation Experimental Facility FASTEF: Main design achievements (part 2: Reactor building design and plant layout) within the FP7-CDT collaborative project of the European Commission

DOE Office of Scientific and Technical Information (OSTI.GOV)

De Bruyn, D.; Engelen, J.; Ortega, A.

MYRRHA (Multi-purpose hybrid Research Reactor for High-tech Applications) is the flexible experimental accelerator-driven system (ADS) in development at SCK-CEN in replacement of its material testing reactor BR2. SCK-CEN in association with 17 European partners from industry, research centres and academia, responded to the FP7 (Seventh Framework Programme) call from the European Commission to establish a Central Design Team (CDT) for the design of a Fast Spectrum Transmutation Experimental Facility (FASTEF) able to demonstrate efficient transmutation and associated technology through a system working in subcritical and/or critical mode. The project has started on April 01, 2009 for a period of threemore » years. In this paper, we present the latest concept of the reactor building and the plant layout. The FASTEF facility has evolved quite a lot since the intermediate reporting done at the ICAPP'10 and ICAPP'11 conferences 1,2. Many iterations have been performed to take into account the safety requirements. The present configuration enables an easy operation and maintenance of the facility, including the possibility to change large components of the reactor. In a companion paper 3, we present the latest configuration of the reactor core and primary system. (authors)« less

Transmutation of singularities and zeros in graded index optical instruments: a methodology for designing practical devices.

PubMed

Hooper, I R; Philbin, T G

2013-12-30

We describe a design methodology for modifying the refractive index profile of graded-index optical instruments that incorporate singularities or zeros in their refractive index. The process maintains the device performance whilst resulting in graded profiles that are all-dielectric, do not require materials with unrealistic values, and that are impedance matched to the bounding medium. This is achieved by transmuting the singularities (or zeros) using the formalism of transformation optics, but with an additional boundary condition requiring the gradient of the co-ordinate transformation be continuous. This additional boundary condition ensures that the device is impedance matched to the bounding medium when the spatially varying permittivity and permeability profiles are scaled to realizable values. We demonstrate the method in some detail for an Eaton lens, before describing the profiles for an "invisible disc" and "multipole" lenses.

Average fast neutron flux in three energy ranges in the Quinta assembly irradiated by two types of beams

NASA Astrophysics Data System (ADS)

Strugalska-Gola, Elzbieta; Bielewicz, Marcin; Kilim, Stanislaw; Szuta, Marcin; Tyutyunnikov, Sergey

2017-03-01

This work was performed within the international project "Energy plus Transmutation of Radioactive Wastes" (E&T - RAW) for investigations of energy production and transmutation of radioactive waste of the nuclear power industry. 89Y (Yttrium 89) samples were located in the Quinta assembly in order to measure an average high neutron flux density in three different energy ranges using deuteron and proton beams from Dubna accelerators. Our analysis showed that the neutron density flux for the neutron energy range 20.8 - 32.7 MeV is higher than for the neutron energy range 11.5 - 20.8 MeV both for protons with an energy of 0.66 GeV and deuterons with an energy of 2 GeV, while for deuteron beams of 4 and 6 GeV we did not observe this.

A brief history of design studies on innovative nuclear reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sekimoto, Hiroshi, E-mail: hsekimot@gmail.com

2014-09-30

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 USAmore » 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.« less

Conjugated Gammadion Chiral Metamaterial with Uniaxial Optical Activity and Negative Refractive Index

DTIC Science & Technology

2011-01-10

in Fig. 4, we discuss a procedure of transmutation from the simple -particle chiral element to the conjugated gammadion chiral metamaterial. The...the transmutation from the simple -particle chiral element to the conjugated gammadion chiral metamaterial. The procedure shows how the magnetic or

Transmutation of All German Transuranium under Nuclear Phase Out Conditions – Is This Feasible from Neutronic Point of View?

PubMed Central

Merk, Bruno; Litskevich, Dzianis

2015-01-01

The German government has decided for the nuclear phase out, but a decision on a strategy for the management of the highly radioactive waste is not defined yet. Partitioning and Transmutation (P&T) could be considered as a technological option for the management of highly radioactive waste, therefore a wide study has been conducted. In the study group objectives for P&T and the boundary conditions of the phase out have been discussed. The fulfillment of the given objectives is analyzed from neutronics point of view using simulations of a molten salt reactor with fast neutron spectrum. It is shown that the efficient transmutation of all existing transuranium isotopes would be possible from neutronic point of view in a time frame of about 60 years. For this task three reactors of a mostly new technology would have to be developed and a twofold life cycle consisting of a transmuter operation and a deep burn phase would be required. A basic insight for the optimization of the time duration of the deep burn phase is given. Further on, a detailed balance of different isotopic inventories is given to allow a deeper understanding of the processes during transmutation in the molten salt fast reactor. The effect of modeling and simulation is investigated based on three different modeling strategies and two different code versions. PMID:26717509

Transmutation of All German Transuranium under Nuclear Phase Out Conditions - Is This Feasible from Neutronic Point of View?

PubMed

Merk, Bruno; Litskevich, Dzianis

2015-01-01

The German government has decided for the nuclear phase out, but a decision on a strategy for the management of the highly radioactive waste is not defined yet. Partitioning and Transmutation (P&T) could be considered as a technological option for the management of highly radioactive waste, therefore a wide study has been conducted. In the study group objectives for P&T and the boundary conditions of the phase out have been discussed. The fulfillment of the given objectives is analyzed from neutronics point of view using simulations of a molten salt reactor with fast neutron spectrum. It is shown that the efficient transmutation of all existing transuranium isotopes would be possible from neutronic point of view in a time frame of about 60 years. For this task three reactors of a mostly new technology would have to be developed and a twofold life cycle consisting of a transmuter operation and a deep burn phase would be required. A basic insight for the optimization of the time duration of the deep burn phase is given. Further on, a detailed balance of different isotopic inventories is given to allow a deeper understanding of the processes during transmutation in the molten salt fast reactor. The effect of modeling and simulation is investigated based on three different modeling strategies and two different code versions.

Neutron transmutation doped Ge bolometers

NASA Technical Reports Server (NTRS)

Haller, E. E.; Kreysa, E.; Palaio, N. P.; Richards, P. L.; Rodder, M.

1983-01-01

Some conclusions reached are as follow. Neutron Transmutation Doping (NTD) of high quality Ge single crystals provides perfect control of doping concentration and uniformity. The resistivity can be tailored to any given bolometer operating temperature down to 0.1 K and probably lower. The excellent uniformity is advantaged for detector array development.

Nuclear waste disposal utilizing a gaseous core reactor

NASA Technical Reports Server (NTRS)

Paternoster, R. R.

1975-01-01

The feasibility of a gaseous core nuclear reactor designed to produce power to also reduce the national inventories of long-lived reactor waste products through nuclear transmutation was examined. Neutron-induced transmutation of radioactive wastes is shown to be an effective means of shortening the apparent half life.

Comparison of Fission Product Yields and Their Impact

DOE Office of Scientific and Technical Information (OSTI.GOV)

S. Harrison

2006-02-01

This memorandum describes the Naval Reactors Prime Contractor Team (NRPCT) Space Nuclear Power Program (SNPP) interest in determining the expected fission product yields from a Prometheus-type reactor and assessing the impact of these species on materials found in the fuel element and balance of plant. Theoretical yield calculations using ORIGEN-S and RACER computer models are included in graphical and tabular form in Attachment, with focus on the desired fast neutron spectrum data. The known fission product interaction concerns are the corrosive attack of iron- and nickel-based alloys by volatile fission products, such as cesium, tellurium, and iodine, and the radiologicalmore » transmutation of krypton-85 in the coolant to rubidium-85, a potentially corrosive agent to the coolant system metal piping.« less

Updated and revised neutron reaction data for 237Np

NASA Astrophysics Data System (ADS)

Chen, Guochang; Wang, Jimin; Cao, Wentian; Tang, Guoyou; Yu, Baosheng

2017-09-01

Nuclear data with high accuracy for minor actinides play an important role in nuclear technology applications, including reactor design and operation, fuel cycle, estimation of the amount of minor actinides in high burn-up reactors and the minor actinides transmutation. Based on the evaluated experimental data, the updated and revised evaluation of a full set of n+237Np nuclear data from 10-5 eV ˜ 20 MeV are carried out and recommended. Mainly revised quantities are neutron multiplicities from fission reaction, inelastic, fission, (n, 2n) and (n, γ) reaction cross sections as well as angular distribution and so on. The promising results are obtained when the renewal evaluated data of 237Np will be used to instead of the evaluated data in CENDL-3.1 database.

Updated and revised neutron reaction data for 236,238Np

NASA Astrophysics Data System (ADS)

Chen, Guochang; Wang, Jimin; Cao, Wentian; Tang, Guoyou; Yu, Baosheng

2017-09-01

Nuclear data with high accuracy for minor actinides play an important role in nuclear technology applications, including reactor design and operation, fuel cycle, estimation of the amount of minor actinides in high burn-up reactors and the minor actinides transmutation. Based on a new set of neutron optical model parameter and the reaction cross section systematics of fissile isotopes, a full set of 236,238Np neutron reaction data from 10-5 eV ˜20 MeV are updated and improved through theoretical calculation. Mainly revised quantities include the total, elastic, inelastic, fission, (n, 2n) and (n, γ) reaction cross sections as well as angular distribution etc. The promising results are obtained when the renewal evaluated data of 236,238Np will replace the evaluated data in CENDL-3.1 database.

Apparatus for nuclear transmutation and power production using an intense accelerator-generated thermal neutron flux

DOEpatents

Bowman, C.D.

1992-11-03

Apparatus for nuclear transmutation and power production using an intense accelerator-generated thermal neutron flux. High thermal neutron fluxes generated from the action of a high power proton accelerator on a spallation target allows the efficient burn-up of higher actinide nuclear waste by a two-step process. Additionally, rapid burn-up of fission product waste for nuclides having small thermal neutron cross sections, and the practicality of small material inventories while achieving significant throughput derive from employment of such high fluxes. Several nuclear technology problems are addressed including 1. nuclear energy production without a waste stream requiring storage on a geological timescale, 2. the burn-up of defense and commercial nuclear waste, and 3. the production of defense nuclear material. The apparatus includes an accelerator, a target for neutron production surrounded by a blanket region for transmutation, a turbine for electric power production, and a chemical processing facility. In all applications, the accelerator power may be generated internally from fission and the waste produced thereby is transmuted internally so that waste management might not be required beyond the human lifespan.

Generation of Escher Arts with Dual Perception.

PubMed

Lin, Shih-Syun; Morace, Charles C; Lin, Chao-Hung; Hsu, Li-Fong; Lee, Tong-Yee

2018-02-01

Escher transmutation is a graphic art that smoothly transforms one tile pattern into another tile pattern with dual perception. A classic example is the artwork called Sky and Water, in which a compelling figure-ground arrangement is applied to portray the transmutation of a bird in sky and a fish in water. The shape of a bird is progressively deformed and dissolves into the background while the background gradually reveals the shape of a fish. This paper introduces a system to create a variety of Escher-like transmutations, which includes the algorithms for initializing a tile pattern with dual figure-ground arrangement, for searching for the best matched shape of a user-specified motif from a database, and for transforming the content and shapes of tile patterns using a content-aware warping technique. The proposed system, integrating the graphic techniques of tile initialization, shape matching, and shape warping, allows users to create various Escher-like transmutations with minimal user interaction. Experimental results and conducted user studies demonstrate the feasibility and flexibility of the proposed system in Escher art generation.

Apparatus for nuclear transmutation and power production using an intense accelerator-generated thermal neutron flux

DOEpatents

Bowman, Charles D.

1992-01-01

Apparatus for nuclear transmutation and power production using an intense accelerator-generated thermal neutron flux. High thermal neutron fluxes generated from the action of a high power proton accelerator on a spallation target allows the efficient burn-up of higher actinide nuclear waste by a two-step process. Additionally, rapid burn-up of fission product waste for nuclides having small thermal neutron cross sections, and the practicality of small material inventories while achieving significant throughput derive from employment of such high fluxes. Several nuclear technology problems are addressed including 1. nuclear energy production without a waste stream requiring storage on a geological timescale, 2. the burn-up of defense and commercial nuclear waste, and 3. the production of defense nuclear material. The apparatus includes an accelerator, a target for neutron production surrounded by a blanket region for transmutation, a turbine for electric power production, and a chemical processing facility. In all applications, the accelerator power may be generated internally from fission and the waste produced thereby is transmuted internally so that waste management might not be required beyond the human lifespan.

Shifts in Selective Pressures on Snake Phototransduction Genes Associated with Photoreceptor Transmutation and Dim-Light Ancestry.

PubMed

Schott, Ryan K; Van Nynatten, Alexander; Card, Daren C; Castoe, Todd A; S W Chang, Belinda

2018-06-01

The visual systems of snakes are heavily modified relative to other squamates, a condition often thought to reflect their fossorial origins. Further modifications are seen in caenophidian snakes, where evolutionary transitions between rod and cone photoreceptors, termed photoreceptor transmutations, have occurred in many lineages. Little previous work, however, has focused on the molecular evolutionary underpinnings of these morphological changes. To address this, we sequenced seven snake eye transcriptomes and utilized new whole-genome and targeted capture sequencing data. We used these data to analyze gene loss and shifts in selection pressures in phototransduction genes that may be associated with snake evolutionary origins and photoreceptor transmutation. We identified the surprising loss of rhodopsin kinase (GRK1), despite a low degree of gene loss overall and a lack of relaxed selection early during snake evolution. These results provide some of the first evolutionary genomic corroboration for a dim-light ancestor that lacks strong fossorial adaptations. Our results also indicate that snakes with photoreceptor transmutation experienced significantly different selection pressures from other reptiles. Significant positive selection was found primarily in cone-specific genes, but not rod-specific genes, contrary to our expectations. These results reveal potential molecular adaptations associated with photoreceptor transmutation and also highlight unappreciated functional differences between rod- and cone-specific phototransduction proteins. This intriguing example of snake visual system evolution illustrates how the underlying molecular components of a complex system can be reshaped in response to changing selection pressures.

Scent Transmutation: A New Way to Teach on Chemical Equilibrium, Distillation, and Dynamic Combinatorial Chemistry

ERIC Educational Resources Information Center

Ji, Qing; El-Hamdi, Nadia S.; Miljanic´, Ognjen S?.

2014-01-01

Esters are volatile and pleasantly smelling compounds, commonly used as food additives. Using Ti(OBu)[subscript 4]-catalyzed acyl exchange, we demonstrate a scent transmutation experiment, in which two fragrant esters swap their acyl and alkoxy substituents and are, during the course of a reactive distillation, quantitatively converted into two…

Statistical transmutation in doped quantum dimer models.

PubMed

Lamas, C A; Ralko, A; Cabra, D C; Poilblanc, D; Pujol, P

2012-07-06

We prove a "statistical transmutation" symmetry of doped quantum dimer models on the square, triangular, and kagome lattices: the energy spectrum is invariant under a simultaneous change of statistics (i.e., bosonic into fermionic or vice versa) of the holes and of the signs of all the dimer resonance loops. This exact transformation enables us to define the duality equivalence between doped quantum dimer Hamiltonians and provides the analytic framework to analyze dynamical statistical transmutations. We investigate numerically the doping of the triangular quantum dimer model with special focus on the topological Z(2) dimer liquid. Doping leads to four (instead of two for the square lattice) inequivalent families of Hamiltonians. Competition between phase separation, superfluidity, supersolidity, and fermionic phases is investigated in the four families.

New fixed-point mini-cell to investigate thermocouple drift in a high-temperature environment under neutron irradiation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Laurie, M.; Vlahovic, L.; Rondinella, V.V.

Temperature measurements in the nuclear field require a high degree of reliability and accuracy. Despite their sheathed form, thermocouples subjected to nuclear radiations undergo changes due to radiation damage and transmutation that lead to significant EMF drift during long-term fuel irradiation experiment. For the purpose of a High Temperature Reactor fuel irradiation to take place in the High Flux Reactor Petten, a dedicated fixed-point cell was jointly developed by LNE-Cnam and JRC-IET. The developed cell to be housed in the irradiation rig was tailor made to quantify the thermocouple drift during the irradiation (about two year duration) and withstand highmore » temperature (in the range 950 deg. C - 1100 deg. C) in the presence of contaminated helium in a graphite environment. Considering the different levels of temperature achieved in the irradiation facility and the large palette of thermocouple types aimed at surveying the HTR fuel pebble during the qualification test both copper (1084.62 deg. C) and gold (1064.18 deg. C) fixed-point materials were considered. The aim of this paper is to first describe the fixed-point mini-cell designed to be embedded in the reactor rig and to discuss the preliminary results achieved during some out of pile tests as much as some robustness tests representative of the reactor scram scenarios. (authors)« less

Investigation of the Thermal Stability of Nd(x)Sc(y)Zr(1-x-y)O(2-δ) Materials Proposed for Inert Matrix Fuel Applications.

PubMed

Hayes, John R; Grosvenor, Andrew P; Saoudi, Mouna

2016-02-01

Inert matrix fuels (IMF) consist of transuranic elements (i.e., Pu, Am, Np, Cm) embedded in a neutron transparent (inert) matrix and can be used to "burn up" (transmute) these elements in current or Generation IV nuclear reactors. Yttria-stabilized zirconia has been extensively studied for IMF applications, but the low thermal conductivity of this material limits its usefulness. Other elements can be used to stabilize the cubic zirconia structure, and the thermal conductivity of the fuel can be increased through the use of a lighter stabilizing element. To this end, a series of Nd(x)Sc(y)Zr(1-x-y)O(2-δ) materials has been synthesized via a co-precipitation reaction and characterized by multiple techniques (Nd was used as a surrogate for Am). The long-range and local structures of these materials were studied using powder X-ray diffraction, scanning electron microscopy, and X-ray absorption spectroscopy. Additionally, the stability of these materials over a range of temperatures has been studied by annealing the materials at 1100 and 1400 °C. It was shown that the Nd(x)Sc(y)Zr(1-x-y)O(2-δ) materials maintained a single cubic phase upon annealing at high temperatures only when both Nd and Sc were present with y ≥ 0.10 and x + y > 0.15.

VISION User Guide - VISION (Verifiable Fuel Cycle Simulation) Model

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jacob J. Jacobson; Robert F. Jeffers; Gretchen E. Matthern

2009-08-01

The purpose of this document is to provide a guide for using the current version of the Verifiable Fuel Cycle Simulation (VISION) model. This is a complex model with many parameters; the user is strongly encouraged to read this user guide before attempting to run the model. This model is an R&D work in progress and may contain errors and omissions. It is based upon numerous assumptions. This model is intended to assist in evaluating “what if” scenarios and in comparing fuel, reactor, and fuel processing alternatives at a systems level for U.S. nuclear power. The model is not intendedmore » as a tool for process flow and design modeling of specific facilities nor for tracking individual units of fuel or other material through the system. The model is intended to examine the interactions among the components of a fuel system as a function of time varying system parameters; this model represents a dynamic rather than steady-state approximation of the nuclear fuel system. VISION models the nuclear cycle at the system level, not individual facilities, e.g., “reactor types” not individual reactors and “separation types” not individual separation plants. Natural uranium can be enriched, which produces enriched uranium, which goes into fuel fabrication, and depleted uranium (DU), which goes into storage. Fuel is transformed (transmuted) in reactors and then goes into a storage buffer. Used fuel can be pulled from storage into either separation of disposal. If sent to separations, fuel is transformed (partitioned) into fuel products, recovered uranium, and various categories of waste. Recycled material is stored until used by its assigned reactor type. Note that recovered uranium is itself often partitioned: some RU flows with recycled transuranic elements, some flows with wastes, and the rest is designated RU. RU comes out of storage if needed to correct the U/TRU ratio in new recycled fuel. Neither RU nor DU are designated as wastes. VISION is comprised of several Microsoft Excel input files, a Powersim Studio core, and several Microsoft Excel output files. All must be co-located in the same folder on a PC to function. We use Microsoft Excel 2003 and have not tested VISION with Microsoft Excel 2007. The VISION team uses both Powersim Studio 2005 and 2009 and it should work with either.« less

Modulated electromagnetic fields in inhomogeneous media, hyperbolic pseudoanalytic functions, and transmutations

DOE Office of Scientific and Technical Information (OSTI.GOV)

Khmelnytskaya, Kira V., E-mail: khmel@uaq.edu.mx; Kravchenko, Vladislav V., E-mail: vkravchenko@math.cinvestav.edu.mx; Torba, Sergii M., E-mail: storba@math.cinvestav.edu.mx

2016-05-15

The time-dependent Maxwell system describing electromagnetic wave propagation in inhomogeneous isotropic media in the one-dimensional case reduces to a Vekua-type equation for bicomplex-valued functions of a hyperbolic variable, see Kravchenko and Ramirez [Adv. Appl. Cliord Algebr. 21(3), 547–559 (2011)]. Using this relation, we solve the problem of the transmission through an inhomogeneous layer of a normally incident electromagnetic time-dependent plane wave. The solution is written in terms of a pair of Darboux-associated transmutation operators [Kravchenko, V. V. and Torba, S. M., J. Phys. A: Math. Theor. 45, 075201 (2012)], and combined with the recent results on their construction [Kravchenko, V.more » V. and Torba, S. M., Complex Anal. Oper. Theory 9, 379-429 (2015); Kravchenko, V. V. and Torba, S. M., J. Comput. Appl. Math. 275, 1–26 (2015)] can be used for efficient computation of the transmitted modulated signals. We develop the corresponding numerical method and illustrate its performance with examples.« less

Nuclear Fuel Reprocessing

DOE Office of Scientific and Technical Information (OSTI.GOV)

Harold F. McFarlane; Terry Todd

2013-11-01

Reprocessing is essential to closing nuclear fuel cycle. Natural uranium contains only 0.7 percent 235U, the fissile (see glossary for technical terms) isotope that produces most of the fission energy in a nuclear power plant. Prior to being used in commercial nuclear fuel, uranium is typically enriched to 3–5% in 235U. If the enrichment process discards depleted uranium at 0.2 percent 235U, it takes more than seven tonnes of uranium feed to produce one tonne of 4%-enriched uranium. Nuclear fuel discharged at the end of its economic lifetime contains less one percent 235U, but still more than the natural ore.more » Less than one percent of the uranium that enters the fuel cycle is actually used in a single pass through the reactor. The other naturally occurring isotope, 238U, directly contributes in a minor way to power generation. However, its main role is to transmute into plutoniumby neutron capture and subsequent radioactive decay of unstable uraniumand neptuniumisotopes. 239Pu and 241Pu are fissile isotopes that produce more than 40% of the fission energy in commercially deployed reactors. It is recovery of the plutonium (and to a lesser extent the uranium) for use in recycled nuclear fuel that has been the primary focus of commercial reprocessing. Uraniumtargets irradiated in special purpose reactors are also reprocessed to obtain the fission product 99Mo, the parent isotope of technetium, which is widely used inmedical procedures. Among the fission products, recovery of such expensive metals as platinum and rhodium is technically achievable, but not economically viable in current market and regulatory conditions. During the past 60 years, many different techniques for reprocessing used nuclear fuel have been proposed and tested in the laboratory. However, commercial reprocessing has been implemented along a single line of aqueous solvent extraction technology called plutonium uranium reduction extraction process (PUREX). Similarly, hundreds of types of reactor fuels have been irradiated for different purposes, but the vast majority of commercial fuel is uranium oxide clad in zirconium alloy tubing. As a result, commercial reprocessing plants have relatively narrow technical requirements for used nuclear that is accepted for processing.« less

ADVANCED NUCLEAR FUEL CYCLE EFFECTS ON THE TREATMENT OF UNCERTAINTY IN THE LONG-TERM ASSESSMENT OF GEOLOGIC DISPOSAL SYSTEMS - EBS INPUT

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sutton, M; Blink, J A; Greenberg, H R

2012-04-25

The Used Fuel Disposition (UFD) Campaign within the Department of Energy's Office of Nuclear Energy (DOE-NE) Fuel Cycle Technology (FCT) program has been tasked with investigating the disposal of the nation's spent nuclear fuel (SNF) and high-level nuclear waste (HLW) for a range of potential waste forms and geologic environments. The planning, construction, and operation of a nuclear disposal facility is a long-term process that involves engineered barriers that are tailored to both the geologic environment and the waste forms being emplaced. The UFD Campaign is considering a range of fuel cycles that in turn produce a range of wastemore » forms. The UFD Campaign is also considering a range of geologic media. These ranges could be thought of as adding uncertainty to what the disposal facility design will ultimately be; however, it may be preferable to thinking about the ranges as adding flexibility to design of a disposal facility. For example, as the overall DOE-NE program and industrial actions result in the fuel cycles that will produce waste to be disposed, and the characteristics of those wastes become clear, the disposal program retains flexibility in both the choice of geologic environment and the specific repository design. Of course, other factors also play a major role, including local and State-level acceptance of the specific site that provides the geologic environment. In contrast, the Yucca Mountain Project (YMP) repository license application (LA) is based on waste forms from an open fuel cycle (PWR and BWR assemblies from an open fuel cycle). These waste forms were about 90% of the total waste, and they were the determining waste form in developing the engineered barrier system (EBS) design for the Yucca Mountain Repository design. About 10% of the repository capacity was reserved for waste from a full recycle fuel cycle in which some actinides were extracted for weapons use, and the remaining fission products and some minor actinides were encapsulated in borosilicate glass. Because the heat load of the glass was much less than the PWR and BWR assemblies, the glass waste form was able to be co-disposed with the open cycle waste, by interspersing glass waste packages among the spent fuel assembly waste packages. In addition, the Yucca Mountain repository was designed to include some research reactor spent fuel and naval reactor spent fuel, within the envelope that was set using the commercial reactor assemblies as the design basis waste form. This milestone report supports Sandia National Laboratory milestone M2FT-12SN0814052, and is intended to be a chapter in that milestone report. The independent technical review of this LLNL milestone was performed at LLNL and is documented in the electronic Information Management (IM) system at LLNL. The objective of this work is to investigate what aspects of quantifying, characterizing, and representing the uncertainty associated with the engineered barrier are affected by implementing different advanced nuclear fuel cycles (e.g., partitioning and transmutation scenarios) together with corresponding designs and thermal constraints.« less

New Quantum Diffusion Monte Carlo Method for strong field time dependent problems

NASA Astrophysics Data System (ADS)

Kalinski, Matt

2017-04-01

We have recently formulated the Quantum Diffusion Quantum Monte Carlo (QDMC) method for the solution of the time-dependent Schrödinger equation when it is equivalent to the reaction-diffusion system coupled by the highly nonlinear potentials of the type of Shay. Here we formulate a new Time Dependent QDMC method free of the nonlinearities described by the constant stochastic process of the coupled diffusion with transmutation. As before two kinds of diffusing particles (color walkers) are considered but which can further also transmute one into the other. Each of the species undergoes the hypothetical Einstein random walk progression with transmutation. The progressed particles transmute into the particles of the other kind before contributing to or annihilating the other particles density. This fully emulates the Time Dependent Schrödinger equation for any number of quantum particles. The negative sign of the real and the imaginary parts of the wave function is handled by the ``spinor'' densities carrying the sign as the degree of freedom. We apply the method for the exact time-dependent observation of our discovered two-electron Langmuir configurations in the magnetic and circularly polarized fields.

Simulating irradiation hardening in tungsten under fast neutron irradiation including Re production by transmutation

NASA Astrophysics Data System (ADS)

Huang, Chen-Hsi; Gilbert, Mark R.; Marian, Jaime

2018-02-01

Simulations of neutron damage under fusion energy conditions must capture the effects of transmutation, both in terms of accurate chemical inventory buildup as well as the physics of the interactions between transmutation elements and irradiation defect clusters. In this work, we integrate neutronics, primary damage calculations, molecular dynamics results, Re transmutation calculations, and stochastic cluster dynamics simulations to study neutron damage in single-crystal tungsten to mimic divertor materials. To gauge the accuracy and validity of the simulations, we first study the material response under experimental conditions at the JOYO fast reactor in Japan and the High Flux Isotope Reactor at Oak Ridge National Laboratory, for which measurements of cluster densities and hardening levels up to 2 dpa exist. We then provide calculations under expected DEMO fusion conditions. Several key mechanisms involving Re atoms and defect clusters are found to govern the accumulation of irradiation damage in each case. We use established correlations to translate damage accumulation into hardening increases and compare our results to the experimental measurements. We find hardening increases in excess of 5000 MPa in all cases, which casts doubts about the integrity of W-based materials under long-term fusion exposure.

Neutron induced fission of 237Np - status, challenges and opportunities

NASA Astrophysics Data System (ADS)

Ruskov, Ivan; Goverdovski, Andrei; Furman, Walter; Kopatch, Yury; Shcherbakov, Oleg; Hambsch, Franz-Josef; Oberstedt, Stephan; Oberstedt, Andreas

2018-03-01

Nowadays, there is an increased interest in a complete study of the neutron-induced fission of 237Np. This is due to the need of accurate and reliable nuclear data for nuclear science and technology. 237Np is generated (and accumulated) in the nuclear reactor core during reactor operation. As one of the most abundant long-lived isotopes in spent fuel ("waste"), the incineration of 237Np becomes an important issue. One scenario for burning of 237Np and other radio-toxic minor actinides suggests they are to be mixed into the fuel of future fast-neutron reactors, employing the so-called transmutation and partitioning technology. For testing present fission models, which are at the basis of new generation nuclear reactor developments, highly accurate and detailed neutron-induced nuclear reaction data is needed. However, the EXFOR nuclear database for 237Np on neutron-induced capture cross-section, σγ, and fission cross-section, σf, as well as on the characteristics of capture and fission resonance parameters (Γγ, Γf, σoΓf, fragments mass-energy yield distributions, multiplicities of neutrons vn and γ-rays vγ), has not been updated for decades.

A review and overview of nuclear waste management

DOE Office of Scientific and Technical Information (OSTI.GOV)

Murray, R.L.

1984-12-31

An understanding of the status and issues in the management of radioactive wastes is based on technical information on radioactivity, radiation, biological hazard of radiation exposure, radiation standards, and methods of protection. The fission process gives rise to radioactive fission products and neutron bombardment gives activation products. Radioactive wastes are classified according to source: defense, commercial, industrial, and institutional; and according to physical features: uranium mill tailings, high-level, transuranic, and low-level. The nuclear fuel cycle, which contributes a large fraction of annual radioactive waste, starts with uranium ore, includes nuclear reactor use for electrical power generation, and ends with ultimatemore » disposal of residues. The relation of spent fuel storage and reprocessing is governed by technical, economic, and political considerations. Waste has been successfully solidified in glass and other forms and choices of the containers for the waste form are available. Methods of disposal of high-level waste that have been investigated are transmutation by neutron bombardment, shipment to Antartica, deep-hole insertion, subseabed placement, transfer by rocket to an orbit in space, and disposal in a mined cavity. The latter is the favored method. The choices of host geological media are salt, basalt, tuff, and granite.« less

Transmutation of Matter in Byzantium: The Case of Michael Psellos, the Alchemist

NASA Astrophysics Data System (ADS)

Katsiampoura, Gianna

2008-06-01

There is thus nothing paradoxical about the inclusion of alchemy in the ensemble of the physical sciences nor in the preoccupation with it on the part of learned men engaged in scientific study. In the context of the Medieval model, where discourse on the physical world was ambiguous, often unclear, and lacking the support of experimental verification, the transmutation of matter, which was the subject of alchemy, even if not attended by a host of occult features, was a process that was thought to have a probable basis in reality. What is interesting in this connection is the utilization of the scientific categories of the day for discussion of transmutation of matter and the attempt to avoid, in most instances in the texts that survive, of methods reminiscent of magic.

Development of Improved Burnable Poisons for Commercial Nuclear Power Reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Renier, J.A.

2002-04-17

Burnable poisons are used in all modern nuclear reactors to permit higher loading of fuel without the necessity of an overly large control rod system. This not only permits a longer core life but can also be used to level the power distribution. Commercial nuclear reactors commonly use B{sub 4}C in separate non-fueled rods and more recently, zirconium boride coatings on the fuel pellets or gadolinium oxide mixed with the fuel. Although the advantages are great, there are problems with using these materials. Boron, which is an effective neutron absorber, transmutes to lithium and helium upon absorption of a neutron.more » Helium is insoluble and is eventually released to the interior of the fuel rod, where it produces an internal pressure. When sufficiently high, this pressure stress could cause separation of the cladding from the fuel, causing overly high centerline temperatures. Gadolinium has several very strongly absorbing isotopes, but not all have large cross sections and result in residual burnable poison reactivity worth at the end of the fuel life. Even if the amount of this residual absorber is small and the penalty in operation small, the cost of this penalty, even if only several days, can be very high. The objective of this investigation was to study the performance of single isotopes in order to reduce the residual negative reactivity left over at the end of the fuel cycle. Since the behavior of burnable poisons can be strongly influenced by their configuration, four forms for the absorbers were studied: homogeneously mixed with the fuel, mixed with only the outer one-third of the fuel pellet, coated on the perimeter of the fuel pellets, and alloyed with the cladding. In addition, the numbers of fuel rods containing burnable poison were chosen as 8, 16, 64, and 104. Other configurations were chosen for a few special cases. An enrichment of 4.5 wt% {sup 235}U was chosen for most cases for study in order to achieve a 4-year fuel cycle. A standard pressurized water reactor fuel core was chosen for the study, and state-of-the-art neutronic reactor core computer codes were used for analysis. Power distribution, fuel burnup, reactivity due to burnable poisons and other fission products, spectrum shift, core reactivity, moderator void coefficients, as well as other parameters were calculated as a function of time and fuel burnup. The results not only showed advantages of separation of burnable poison isotopes but revealed benefits to be achieved by careful selection of the configuration of even naturally occurring elements used as burnable poisons. The savings in terms of additional days of operation is shown in Figure 1, where the savings is plotted for each of six favorable isotopes in the four configurations. The benefit of isotope separation is most dramatic for dysprosium, but even the time savings in the case of gadolinium is several days. For a modern nuclear plant, one day's worth of electricity is worth about one million dollars, so the resulting savings of only a few days is considerable. It is also apparent that the amount of savings depends upon the configuration of the burnable poison.« less

Transmutation of Isotopes --- Ecological and Energy Production Aspects

NASA Astrophysics Data System (ADS)

Gudowski, Waclaw

2000-01-01

This paper describes principles of Accelerator-Driven Transmutation of Nuclear Wastes (ATW) and gives some flavour of the most important topics which are today under investigations in many countries. An assessment of the potential impact of ATW on a future of nuclear energy is also given. Nuclear reactors based on self-sustained fission reactions --- after spectacular development in fifties and sixties, that resulted in deployment of over 400 power reactors --- are wrestling today more with public acceptance than with irresolvable technological problems. In a whole spectrum of reasons which resulted in today's opposition against nuclear power few of them are very relevant for the nuclear physics community and they arose from the fact that development of nuclear power had been handed over to the nuclear engineers and technicians with some generically unresolved problems, which should have been solved properly by nuclear scientists. In a certain degree of simplification one can say, that most of the problems originate from very specific features of a fission phenomenon: self-sustained chain reaction in fissile materials and very strong radioactivity of fission products and very long half-life of some of the fission and activation products. And just this enormous concentration of radioactive fission products in the reactor core is the main problem of managing nuclear reactors: it requires unconditional guarantee for the reactor core integrity in order to avoid radioactive contamination of the environment; it creates problems to handle decay heat in the reactor core and finally it makes handling and/or disposal of spent fuel almost a philosophical issue, due to unimaginable long time scales of radioactive decay of some isotopes. A lot can be done to improve the design of conventional nuclear reactors (like Light Water Reactors); new, better reactors can be designed but it seems today very improbable to expect any radical change in the public perception of conventional nuclear power. In this context a lot of hopes and expectations have been expressed for novel systems called Accelerator-Driven Systems, Accelerator-Driven Transmutation of Waste or just Hybrid Reactors. All these names are used for description of the same nuclear system combining a powerful particle accelerator with a subcritical reactor. A careful analysis of possible environmental impact of ATW together with limitation of this technology is presented also in this paper.

The Euratom Seventh Framework Programme FP7 (2007-2011)

NASA Astrophysics Data System (ADS)

Garbil, R.

2010-10-01

The objective of the Seventh Euratom Framework Program in the area of nuclear fission and radiation protection is to establish a sound scientific and technical basis to accelerate practical developments of nuclear energy related to resource efficiency, enhancing safety performance, cost-effectiveness and safer management of long-lived radioactive waste. Key cross-cutting topics such as the nuclear fuel cycle, actinide chemistry, risk analysis, safety assessment, even societal and governance issues are linked to the individual technical areas. Research need to explore new scientific and techno- logical opportunities and to respond in a flexible way to new policy needs that arise. The following activities are to be pursued. (a) Management of radioactive waste, research on partitioning and transmutation and/or other concepts aimed at reducing the amount and/or hazard of the waste for disposal; (b) Reactor systems research to underpin the con- tinued safe operation of all relevant types of existing reactor systems (including fuel cycle facilities), life-time extension, development of new advanced safety assessment methodologies and waste-management aspects of future reactor systems; (c) Radiation protection research in particular on the risks from low doses on medical uses and on the management of accidents; (d) Infrastructures and support given to the availability of, and cooperation between, research infrastructures necessary to maintain high standards of technical achievement, innovation and safety in the European nuclear sector and Research Area. (e) Human resources, mobility and training support to be provided for the retention and further development of scientific competence, human capacity through joint training activities in order to guarantee the availability of suitably qualified researchers, engineers and employees in the nuclear sector over the longer term.

Survey of Materials for Fusion Fission Hybrid Reactors Vol 1 Rev. 0

DOE Office of Scientific and Technical Information (OSTI.GOV)

Farmer, Joseph Collin

2007-07-03

Materials for fusion-fission hybrid reactors fall into several broad categories, including fuels, blanket and coolant materials, cladding, structural materials, shielding, and in the specific case of inertial-confinement fusion systems, laser and optical materials. This report surveys materials in all categories of materials except for those required for lasers and optics. Preferred collants include two molten salt mixtures known as FLIBE (Li2BeF4) and FLINABE (LiNaBeF4). In the case of homogenous liquid fuels, UF4 can be dissolved in these molten salt mixtures. The transmutation of lithium in this coolant produces very corrosive hydrofluoric acid species (HF and TF), which can rapidly degrademore » structural materials. Broad ranges of high-melting radiation-tolerant structural material have been proposed for fusion-fission reactor structures. These include a wide variety of steels and refractory alloys. Ferritic steels with oxide-dispersion strengthening and graphite have been given particular attention. Refractory metals are found in Groups IVB and VB of the periodic table, and include Nb, Ta, Cr, Mo, and W, as serve as the basis of refractory alloys. Stable high-melting composites and amorphous metals may also be useful. Since amorphous metals have no lattice structure, neutron bombardment cannot dislodge atoms from lattice sites, and the materials would be immune from this specific mode of degradation. The free energy of formation of fluorides of the alloying elements found in steels and refractory alloys can be used to determine the relative stability of these materials in molten salts. The reduction of lithium transmutation products (H + and T +) drives the electrochemical corrosion process, and liberates aggressive fluoride ions that pair with ions formed from dissolved structural materials. Corrosion can be suppressed through the use of metallic Be and Li, though the molten salt becomes laden with colloidal suspensions of Be and Li corrosion products in the process. Alternatively, imposed currents and other high-temperature cathodic protection systems are envisioned for protection of the structural materials. This novel concept could prove to be enabling technology for such high-temperature molten-salt reactors. The use of UF 4 as a liquid-phase homogenous fuel is also complicated by redox control. For example, the oxidation of tetravalent uranium to hexavalent uranium could result in the formation of volatile UF 6. This too could be controlled through electrochemically manipulated oxidation and reduction reactions. In situ studies of pertinent electrochemical reactions in the molten salts are proposed, and are relevant to both the corrosive attack of structural materials, as well as the volatilization of fuel. Some consideration is given to the potential advantages of gravity fed falling-film blankets. Such systems may be easier to control than vortex systems, but would require that cylindrical reaction vessels be oriented with the centerline normal to the gravitational field.« less

Cone-like rhodopsin expressed in the all-cone retina of the colubrid pine snake as a potential adaptation to diurnality.

PubMed

Bhattacharyya, Nihar; Darren, Benedict; Schott, Ryan K; Tropepe, Vincent; Chang, Belinda S W

2017-07-01

Colubridae is the largest and most diverse family of snakes, with visual systems that reflect this diversity, encompassing a variety of retinal photoreceptor organizations. The transmutation theory proposed by Walls postulates that photoreceptors could evolutionarily transition between cell types in squamates, but few studies have tested this theory. Recently, evidence for transmutation and rod-like machinery in an all-cone retina has been identified in a diurnal garter snake ( Thamnophis ), and it appears that the rhodopsin gene at least may be widespread among colubrid snakes. However, functional evidence supporting transmutation beyond the existence of the rhodopsin gene remains rare. We examined the all-cone retina of another colubrid, Pituophis melanoleucus , thought to be more secretive/burrowing than Thamnophis We found that P. melanoleucus expresses two cone opsins (SWS1, LWS) and rhodopsin (RH1) within the eye. Immunohistochemistry localized rhodopsin to the outer segment of photoreceptors in the all-cone retina of the snake and all opsin genes produced functional visual pigments when expressed in vitro Consistent with other studies, we found that P. melanoleucus rhodopsin is extremely blue-shifted. Surprisingly, P. melanoleucus rhodopsin reacted with hydroxylamine, a typical cone opsin characteristic. These results support the idea that the rhodopsin-containing photoreceptors of P. melanoleucus are the products of evolutionary transmutation from rod ancestors, and suggest that this phenomenon may be widespread in colubrid snakes. We hypothesize that transmutation may be an adaptation for diurnal, brighter-light vision, which could result in increased spectral sensitivity and chromatic discrimination with the potential for colour vision. © 2017. Published by The Company of Biologists Ltd.

Summary of Research Activities. Academic Departments, 1979-1980.

DTIC Science & Technology

1979-10-01

studied, including recycling, geologic storage, transmutation , ejection from earth, and seabed disposal. Currently, the most favored methods are...official society. (4) The poem ultimately celebrates in a sort of poetic eucharist the regenerative power of poetry to transmute the bread and wine of...Union prevails. Such a view becomes commonplace after Burr’s political enemies attack him as a Catiline, Cain, and sexual predator. Influenced by these

The Changing Nature of Warfare, the Factors Mediating Future Conflict, and Implications for SOF

DTIC Science & Technology

2006-04-01

most commonly used vernacular today, one is describing the person’s sexual orientation.13 The English language is replete with other words that have...countries and those that are technologically more advanced, primarily due to labor cost differentials. Globaliza- tion has transmuted economics from a...fascination with forensic sci- ences, ala the acclaimed television pro- gram CSI, has transmuted from civilian criminal proceedings to microscopic

FIRST-PRINCIPLES CALCULATIONS OF CHARGE STATES AND FORMATION ENERGIES OF Mg, Al, and Be TRANSMUTANTS IN 3C-SiC

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hu, Shenyang Y.; Setyawan, Wahyu; Jiang, Weilin

2014-08-28

The Vienna Ab-initio Simulation Package (VASP) is employed to calculate charge states and the formation energies of Mg, Al and Be transmutants at different lattice sites in 3C-SiC. The results provide important information on the dependence of the most stable charge state and formation energy of Mg, Al, Be and vacancies on electron potentials.

Electric Spark Discharges in Water. Low-energy Nuclear Transmutations and Light Leptonic Magnetic Monopoles in an Extended Standard Model

NASA Astrophysics Data System (ADS)

Stumpf, Harald

2017-08-01

Light leptonic magnetic monopoles were predicted by Lochak [G. Lochak, Intern. J. Theor. Phys. 24, 1019 (1985).]. Experimental indications based on nuclear transmutations were announced by Urutskoiev et al. [L. I. Urutskoiev, V. I. Liksonov, V. G. Tsinoev, Ann. Fond. L. de Broglie 27, Nr.4, 791 (2002).] and Urutskoev [L. J. Urutskoev, Ann. Fond. L. de Broglie 29, 1149 (2004).]. A theoretical interpretation of these transmutations is proposed under the assumption that light leptonic magnetic monopoles are created during spark discharges in water. The latter should be excited neutrinos according to Lochak. This hypothesis enforces the introduction of an extended Standard Model described in previous papers. The most important results of this study are (i) that multiple proton captures are responsible for the variety of transmutations and that leptonic magnetic monopoles are involved in these processes (ii) that electromagnetic duality can be established for bound states of leptonic monopoles although massive monopoles are in general unstable (iii) that criteria for the emission of leptonic magnetic monopoles and for their catalytic effect on weak decays are set up and elaborated. The study can be considered as a contribution to the efforts of Urutskoiev and Lochak to understand the reasons for accidents in power plants.

Study of proton- and deuteron-induced spallation reactions on the long-lived fission product 93Zr at 105 MeV/nucleon in inverse kinematics

NASA Astrophysics Data System (ADS)

Kawase, Shoichiro; Nakano, Keita; Watanabe, Yukinobu; Wang, He; Otsu, Hideaki; Sakurai, Hiroyoshi; Ahn, Deuk Soon; Aikawa, Masayuki; Ando, Takashi; Araki, Shouhei; Chen, Sidong; Chiga, Nobuyuki; Doornenbal, Pieter; Fukuda, Naoki; Isobe, Tadaaki; Kawakami, Shunsuke; Kin, Tadahiro; Kondo, Yosuke; Koyama, Shunpei; Kubono, Shigeru; Maeda, Yukie; Makinaga, Ayano; Matsushita, Masafumi; Matsuzaki, Teiichiro; Michimasa, Shin'ichiro; Momiyama, Satoru; Nagamine, Shunsuke; Nakamura, Takashi; Niikura, Megumi; Ozaki, Tomoyuki; Saito, Atsumi; Saito, Takeshi; Shiga, Yoshiaki; Shikata, Mizuki; Shimizu, Yohei; Shimoura, Susumu; Sumikama, Toshiyuki; Söderström, Pär-Anders; Suzuki, Hiroshi; Takeda, Hiroyuki; Takeuchi, Satoshi; Taniuchi, Ryo; Togano, Yasuhiro; Tsubota, Jun'ichi; Uesaka, Meiko; Watanabe, Yasushi; Wimmer, Kathrin; Yamamoto, Tatsuya; Yoshida, Koichi

2017-09-01

Spallation reactions for the long-lived fission product ^{93}Zr have been studied in order to provide basic data necessary for nuclear waste transmutation. Isotopic-production cross sections via proton- and deuteron-induced spallation reactions on ^{93}Zr at 105 MeV/nucleon were measured in inverse kinematics at the RIKEN Radioactive Isotope Beam Factory. Remarkable jumps in isotopic production originating from the neutron magic number N=50 were observed in Zr and Y isotopes. The experimental results were compared to the PHITS calculations considering both the intranuclear cascade and evaporation processes, and the calculations greatly overestimated the measured production yield, corresponding to few-nucleon-removal reactions. The present data suggest that the spallation reaction is a potential candidate for the treatment of ^{93}Zr in spent nuclear fuel.

Method of detecting leakage of reactor core components of liquid metal cooled fast reactors

DOEpatents

Holt, Fred E.; Cash, Robert J.; Schenter, Robert E.

1977-01-01

A method of detecting the failure of a sealed non-fueled core component of a liquid-metal cooled fast reactor having an inert cover gas. A gas mixture is incorporated in the component which includes Xenon-124; under neutron irradiation, Xenon-124 is converted to radioactive Xenon-125. The cover gas is scanned by a radiation detector. The occurrence of 188 Kev gamma radiation and/or other identifying gamma radiation-energy level indicates the presence of Xenon-125 and therefore leakage of a component. Similarly, Xe-126, which transmutes to Xe-127 and Kr-84, which produces Kr-85.sup.m can be used for detection of leakage. Different components are charged with mixtures including different ratios of isotopes other than Xenon-124. On detection of the identifying radiation, the cover gas is subjected to mass spectroscopic analysis to locate the leaking component.

Advancing the scientific basis of trivalent actinide-lanthanide separations

DOE Office of Scientific and Technical Information (OSTI.GOV)

Nash, K.L.

For advanced fuel cycles designed to support transmutation of transplutonium actinides, several options have been demonstrated for process-scale aqueous separations for U, Np, Pu management and for partitioning of trivalent actinides and fission product lanthanides away from other fission products. The more difficult mutual separation of Am/Cm from La-Tb remains the subject of considerable fundamental and applied research. The chemical separations literature teaches that the most productive alternatives to pursue are those based on ligand donor atoms less electronegative than O, specifically N- and S-containing complexants and chloride ion (Cl{sup -}). These 'soft-donor' atoms have exhibited usable selectivity in theirmore » bonding interactions with trivalent actinides relative to lanthanides. In this report, selected features of soft donor reagent design, characterization and application development will be discussed. The roles of thiocyanate, aminopoly-carboxylic acids and lactate in separation processes are detailed. (authors)« less

Advanced Space Nuclear Reactors from Fiction to Reality

NASA Astrophysics Data System (ADS)

Popa-Simil, L.

The advanced nuclear power sources are used in a large variety of science fiction movies and novels, but their practical development is, still, in its early conceptual stages, some of the ideas being confirmed by collateral experiments. The novel reactor concept uses the direct conversion of nuclear energy into electricity, has electronic control of reactivity, being surrounded by a transmutation blanket and very thin shielding being small and light that at its very limit may be suitable to power an autonomously flying car. It also provides an improved fuel cycle producing minimal negative impact to environment. The key elements started to lose the fiction attributes, becoming viable actual concepts and goals for the developments to come, and on the possibility to achieve these objectives started to become more real because the theory shows that using the novel nano-technologies this novel reactor might be achievable in less than a century.

Vortex transmutation.

PubMed

Ferrando, Albert; Zacarés, Mario; García-March, Miguel-Angel; Monsoriu, Juan A; de Córdoba, Pedro Fernández

2005-09-16

Using group theory arguments and numerical simulations, we demonstrate the possibility of changing the vorticity or topological charge of an individual vortex by means of the action of a system possessing a discrete rotational symmetry of finite order. We establish on theoretical grounds a "transmutation pass" determining the conditions for this phenomenon to occur and numerically analyze it in the context of two-dimensional optical lattices. An analogous approach is applicable to the problems of Bose-Einstein condensates in periodic potentials.

NEUTRON FLUX INTENSITY DETECTION

DOEpatents

Russell, J.T.

1964-04-21

A method of measuring the instantaneous intensity of neutron flux in the core of a nuclear reactor is described. A target gas capable of being transmuted by neutron bombardment to a product having a resonance absorption line nt a particular microwave frequency is passed through the core of the reactor. Frequency-modulated microwave energy is passed through the target gas and the attenuation of the energy due to the formation of the transmuted product is measured. (AEC)

Hydrogen bond disruption in DNA base pairs from (14)C transmutation.

PubMed

Sassi, Michel; Carter, Damien J; Uberuaga, Blas P; Stanek, Christopher R; Mancera, Ricardo L; Marks, Nigel A

2014-09-04

Recent ab initio molecular dynamics simulations have shown that radioactive carbon does not normally fragment DNA bases when it decays. Motivated by this finding, density functional theory and Bader analysis have been used to quantify the effect of C → N transmutation on hydrogen bonding in DNA base pairs. We find that (14)C decay has the potential to significantly alter hydrogen bonds in a variety of ways including direct proton shuttling (thymine and cytosine), thermally activated proton shuttling (guanine), and hydrogen bond breaking (cytosine). Transmutation substantially modifies both the absolute and relative strengths of the hydrogen bonding pattern, and in two instances (adenine and cytosine), the density at the critical point indicates development of mild covalent character. Since hydrogen bonding is an important component of Watson-Crick pairing, these (14)C-induced modifications, while infrequent, may trigger errors in DNA transcription and replication.

The Changes of Energy Interactions between Nucleus Function and Mitochondria Functions Causing Transmutation of Chronic Inflammation into Cancer Metabolism.

PubMed

Ponizovskiy, Michail R

2016-01-01

Interactions between nucleus and mitochondria functions induce the mechanism of maintenance stability of cellular internal energy according to the first law of thermodynamics in able-bodied cells and changes the mechanisms of maintenance stability of cellular internal energy creating a transition stationary state of ablebodied cells into quasi-stationary pathologic states of acute inflammation transiting then into chronic inflammation and then transmuting into cancer metabolism. The mechanisms' influences of intruding etiologic pathologic agents (microbe, virus, etc.) lead to these changes of energy interactions between nucleus and mitochondria functions causing general acute inflammation, then passing into local chronic inflammation, and reversing into cancer metabolism transmutation. Interactions between biochemical processes and biophysical processes of cellular capacitors' operations create a supplementary mechanism of maintenance stability of cellular internal energy in the norm and in pathology. Discussion of some scientific works eliminates doubts of the authors of these works.

CURE: Clean use of reactor energy

DOE Office of Scientific and Technical Information (OSTI.GOV)

NONE

1990-05-01

This paper presents the results of a joint Westinghouse Hanford Company (Westinghouse Hanford)-Pacific Northwest Laboratory (PNL) study that considered the feasibility of treating radioactive waste before disposal to reduce the inventory of long-lived radionuclides, making the waste more suitable for geologic disposal. The treatment considered here is one in which waste would be chemically separated so that long-lived radionuclides can be treated using specific processes appropriate for the nuclide. The technical feasibility of enhancing repository performance by this type of treatment is considered in this report. A joint Westinghouse Hanford-PNL study group developed a concept called the Clean Use ofmore » Reactor Energy (CURE), and evaluated the potential of current technology to reduce the long-lived radionuclide content in waste from the nuclear power industry. The CURE process consists of three components: chemical separation of elements that have significant quantities of long-lived radioisotopes in the waste, exposure in a neutron flux to transmute the radioisotopes to stable nuclides, and packaging of radionuclides that cannot be transmuted easily for storage or geologic disposal. 76 refs., 32 figs., 24 tabs.« less

The DD Cold Fusion-Transmutation Connection

NASA Astrophysics Data System (ADS)

Chubb, Talbot A.

2005-12-01

LENR theory must explain dd fusion, alpha-addition transmutations, radiationless nuclear reactions, and three-body nuclear particle reactions. Reaction without radiation requires many-body D Bloch+ periodicity in both location and internal structure dependencies. Electron scattering leads to mixed quantum states. The radiationless dd fusion reaction is 2-D Bloch+ -> {}4 He Bloch2+. Overlap between {}4 He Bloch2+ and surface Cs leads to alpha absorption. In the Iwamura et al. studies active deuterium is created by scattering at diffusion barriers.

Impact of the High Flux Isotope Reactor HEU to LEU Fuel Conversion on Cold Source Nuclear Heat Generation Rates

DOE Office of Scientific and Technical Information (OSTI.GOV)

Chandler, David

2014-03-01

Under the sponsorship of the US Department of Energy National Nuclear Security Administration, staff members at the Oak Ridge National Laboratory have been conducting studies to determine whether the High Flux Isotope Reactor (HFIR) can be converted from high enriched uranium (HEU) fuel to low enriched uranium (LEU) fuel. As part of these ongoing studies, an assessment of the impact that the HEU to LEU fuel conversion has on the nuclear heat generation rates in regions of the HFIR cold source system and its moderator vessel was performed and is documented in this report. Silicon production rates in the coldmore » source aluminum regions and few-group neutron fluxes in the cold source moderator were also estimated. Neutronics calculations were performed with the Monte Carlo N-Particle code to determine the nuclear heat generation rates in regions of the HFIR cold source and its vessel for the HEU core operating at a full reactor power (FP) of 85 MW(t) and the reference LEU core operating at an FP of 100 MW(t). Calculations were performed with beginning-of-cycle (BOC) and end-of-cycle (EOC) conditions to bound typical irradiation conditions. Average specific BOC heat generation rates of 12.76 and 12.92 W/g, respectively, were calculated for the hemispherical region of the cold source liquid hydrogen (LH2) for the HEU and LEU cores, and EOC heat generation rates of 13.25 and 12.86 W/g, respectively, were calculated for the HEU and LEU cores. Thus, the greatest heat generation rates were calculated for the EOC HEU core, and it is concluded that the conversion from HEU to LEU fuel and the resulting increase of FP from 85 MW to 100 MW will not impact the ability of the heat removal equipment to remove the heat deposited in the cold source system. Silicon production rates in the cold source aluminum regions are estimated to be about 12.0% greater at BOC and 2.7% greater at EOC for the LEU core in comparison to the HEU core. Silicon is aluminum s major transmutation product and affects mechanical properties of aluminum including density, neutron irradiation hardening, swelling, and loss of ductility. Because slightly greater quantities of silicon will be produced in the cold source moderator vessel for the LEU core, these effects will be slightly greater for the LEU core than for the HEU core. Three-group (thermal, epithermal, and fast) neutron flux results tallied in the cold source LH2 hemisphere show greater values for the LEU core under both BOC and EOC conditions. The thermal neutron flux in the LH2 hemisphere for the LEU core is about 12.4% greater at BOC and 2.7% greater at EOC than for the HEU core. Therefore, cold neutron scattering will not be adversely affected and the 4 12 neutrons conveyed to the cold neutron guide hall for research applications will be enhanced.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wieselquist, William A.

SCALE’s general depletion, activation, and spent fuel source terms analysis capabilities are enabled through a family of modules related to the main ORIGEN depletion/irradiation/decay solver. The nuclide tracking in ORIGEN is based on the principle of explicitly modeling all available nuclides and transitions in the current fundamental nuclear data for decay and neutron-induced transmutation and relies on fundamental cross section and decay data in ENDF/B VII. Cross section data for materials and reaction processes not available in ENDF/B-VII are obtained from the JEFF-3.0/A special purpose European activation library containing 774 materials and 23 reaction channels with 12,617 neutron-induced reactions belowmore » 20 MeV. Resonance cross section corrections in the resolved and unresolved range are performed using a continuous-energy treatment by data modules in SCALE. All nuclear decay data, fission product yields, and gamma-ray emission data are developed from ENDF/B-VII.1 evaluations. Decay data include all ground and metastable state nuclides with half-lives greater than 1 millisecond. Using these data sources, ORIGEN currently tracks 174 actinides, 1149 fission products, and 974 activation products. The purpose of this chapter is to describe the stand-alone capabilities and underlying methodology of ORIGEN—as opposed to the integrated depletion capability it provides in all coupled neutron transport/depletion sequences in SCALE, as described in other chapters.« less

Development of NTD Ge Sensors for Superconducting Bolometer

NASA Astrophysics Data System (ADS)

Garai, A.; Mathimalar, S.; Singh, V.; Dokania, N.; Nanal, V.; Pillay, R. G.; Ramakrishnan, S.; Shrivastava, A.; Jagadeesan, K. C.; Thakare, S. V.

2016-08-01

Neutron transmutation-doped (NTD) Ge sensors have been prepared by irradiating device-grade Ge with thermal neutrons at Dhruva reactor, BARC, Mumbai. These sensors are intended to be used for the study of neutrinoless double beta decay in ^{124}Sn with a superconducting Tin bolometer. Resistance measurements are performed on NTD Ge sensors in the temperature range 100-350 mK. The observed temperature dependence is found to be consistent with the variable-range hopping mechanism.

Spallation reaction study for the long-lived fission product 107Pd

NASA Astrophysics Data System (ADS)

Wang, He; Otsu, Hideaki; Sakurai, Hiroyoshi; Ahn, DeukSoon; Aikawa, Masayuki; Ando, Takashi; Araki, Shouhei; Chen, Sidong; Nobuyuki, Chiga; Doornenbal, Pieter; Fukuda, Naoki; Isobe, Tadaaki; Kawakami, Shunsuke; Kawase, Shoichiro; Kin, Tadahiro; Kondo, Yosuke; Koyama, Shunpei; Kubono, Shigeru; Maeda, Yukie; Makinaga, Ayano; Matsushita, Masafumi; Matsuzaki, Teiichiro; Michimasa, Shin'ichiro; Momiyama, Satoru; Nagamine, Shunsuke; Nakamura, Takashi; Nakano, Keita; Niikura, Megumi; Ozaki, Tomoyuki; Saito, Atsumi; Saito, Takeshi; Shiga, Yoshiaki; Shikata, Mizuki; Shimizu, Yohei; Shimoura, Susumu; Sumikama, Toshiyuki; Söderström, Pär-Anders; Suzuki, Hiroshi; Takeda, Hiroyuki; Takeuchi, Satoshi; Taniuchi, Ryo; Togano, Yasuhiro; Tsubota, Junichi; Uesaka, Meiko; Watanabe, Yasushi; Watanabe, Yukinobu; Wimmer, Kathrin; Yamamoto, Tatsuya; Yoshida, Koichi

2017-02-01

Spallation reactions for the long-lived fission product 107Pd have been studied for the purpose of nuclear waste transmutation. The cross sections on the proton- and deuteron-induced spallation were obtained at 196 and 118 MeV/nucleon in inverse kinematics at the RIKEN Radioactive Isotope Beam Factory. Both the target and energy dependences of cross sections have been investigated systematically. It was found that the proton-induced cross sections at 196 MeV/nucleon are close to those for deuteron obtained at 118 MeV/nucleon for the light-mass products. The experimental data are compared with the SPACS semi-empirical parameterization and the PHITS calculations including both the intranuclear cascade and evaporation processes. Our data give a design goal of proton/deuteron flux for the transmutation of 107Pd using the spallation reaction. In addition, it is found that the spallation reaction at 118 MeV/nucleon may have an advantage over the 107Pd transmutation because of the low production of other long-lived radioactive isotopes.

Fission-suppressed fusion breeder on the thorium cycle and nonproliferation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Moir, R. W.

2012-06-19

Fusion reactors could be designed to breed fissile material while suppressing fissioning thereby enhancing safety. The produced fuel could be used to startup and makeup fuel for fission reactors. Each fusion reaction can produce typically 0.6 fissile atoms and release about 1.6 times the 14 MeV neutron's energy in the blanket in the fission-suppressed design. This production rate is 2660 kg/1000 MW of fusion power for a year. The revenues would be doubled from such a plant by selling fuel at a price of 60/g and electricity at $0.05/kWh for Q=P{sub fusion}/P{sub input}=4. Fusion reactors could be designed to destroymore » fission wastes by transmutation and fissioning but this is not a natural use of fusion whereas it is a designed use of fission reactors. Fusion could supply makeup fuel to fission reactors that were dedicated to fissioning wastes with some of their neutrons. The design for safety and heat removal and other items is already accomplished with fission reactors. Whereas fusion reactors have geometry that compromises safety with a complex and thin wall separating the fusion zone from the blanket zone where wastes could be destroyed. Nonproliferation can be enhanced by mixing {sup 233}U with {sup 238}U. Also nonproliferation is enhanced in typical fission-suppressed designs by generating up to 0.05 {sup 232}U atoms for each {sup 233}U atom produced from thorium, about twice the IAEA standards of 'reduced protection' or 'self protection.' With 2.4%{sup 232}U, high explosive material is predicted to degrade owing to ionizing radiation after a little over 1/2 year and the heat rate is 77 W just after separation and climbs to over 600 W ten years later. The fissile material can be used to fuel most any fission reactor but is especially appropriate for molten salt reactors (MSR) also called liquid fluoride thorium reactors (LFTR) because of the molten fuel does not need hands on fabrication and handling.« less

Thermal Conductivity of Metallic Uranium

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hin, Celine

This project has developed a modeling and simulation approaches to predict the thermal conductivity of metallic fuels and their alloys. We focus on two methods. The first method has been developed by the team at the University of Wisconsin Madison. They developed a practical and general modeling approach for thermal conductivity of metals and metal alloys that integrates ab-initio and semi-empirical physics-based models to maximize the strengths of both techniques. The second method has been developed by the team at Virginia Tech. This approach consists of a determining the thermal conductivity using only ab-initio methods without any fitting parameters. Bothmore » methods were complementary. The models incorporated both phonon and electron contributions. Good agreement with experimental data over a wide temperature range were found. The models also provided insight into the different physical factors that govern the thermal conductivity under different temperatures. The models were general enough to incorporate more complex effects like additional alloying species, defects, transmutation products and noble gas bubbles to predict the behavior of complex metallic alloys like U-alloy fuel systems under burnup. 3 Introduction Thermal conductivity is an important thermal physical property affecting the performance and efficiency of metallic fuels [1]. Some experimental measurement of thermal conductivity and its correlation with composition and temperature from empirical fitting are available for U, Zr and their alloys with Pu and other minor actinides. However, as reviewed in by Kim, Cho and Sohn [2], due to the difficulty in doing experiments on actinide materials, thermal conductivities of metallic fuels have only been measured at limited alloy compositions and temperatures, some of them even being negative and unphysical. Furthermore, the correlations developed so far are empirical in nature and may not be accurate when used for prediction at conditions far from those used in the original fitting. Moreover, as fuels burn up in the reactor and fission products are built up, thermal conductivity is also significantly changed [3]. Unfortunately, fundamental understanding of the effect of fission products is also currently lacking. In this project, we probe thermal conductivity of metallic fuels with ab initio calculations, a theoretical tool with the potential to yield better accuracy and predictive power than empirical fitting. This work will both complement experimental data by determining thermal conductivity in wider composition and temperature ranges than is available experimentally, and also develop mechanistic understanding to guide better design of metallic fuels in the future. So far, we focused on α-U perfect crystal, the ground-state phase of U metal. We focus on two methods. The first method has been developed by the team at the University of Wisconsin Madison. They developed a practical and general modeling approach for thermal conductivity of metals and metal alloys that integrates ab-initio and semi-empirical physics-based models to maximize the strengths of both techniques. The second method has been developed by the team at Virginia Tech. This approach consists of a determining the thermal conductivity using only ab-initio methods without any fitting parameters. Both methods were complementary and very helpful to understand the physics behind the thermal conductivity in metallic uranium and other materials with similar characteristics. In Section I, the combined model developed at UWM is explained. In Section II, the ab-initio method developed at VT is described along with the uranium pseudo-potential and its validation. Section III is devoted to the work done by Jianguo Yu at INL. Finally, we will present the performance of the project in terms of milestones, publications, and presentations.« less

II. Inhibited Diffusion Driven Surface Transmutations

NASA Astrophysics Data System (ADS)

Chubb, Talbot A.

2006-02-01

This paper is the second of a set of three papers dealing with the role of coherent partitioning as a common element in Low Energy Nuclear Reactions (LENR), by which is meant cold-fusion related processes. This paper discusses the first step in a sequence of four steps that seem to be necessary to explain Iwamura 2-α-addition surface transmutations. Three concepts are examined: salt-metal interface states, sequential tunneling that transitions D+ ions from localized interstitial to Bloch form, and the general applicability of 2-dimensional vs. 3-dimensional symmetry hosting networks.

Transmutation of skyrmions to half-solitons driven by the nonlinear optical spin Hall effect.

PubMed

Flayac, H; Solnyshkov, D D; Shelykh, I A; Malpuech, G

2013-01-04

We show that the spin domains, generated in the linear optical spin Hall effect by the analog of spin-orbit interaction for exciton polaritons, are associated with the formation of a Skyrmion lattice. In the nonlinear regime, the spin anisotropy of the polariton-polariton interactions results in a spatial compression of the domains and in a transmutation of the Skyrmions into oblique half-solitons. This phase transition is associated with both the focusing of the spin currents and the emergence of a strongly anisotropic emission pattern.

Thermodynamic and Kinetic Properties of Intrinsic Defects and Mg Transmutants in 3C-SiC Determined by Density Functional Theory

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hu, Shenyang Y.; Setyawan, Wahyu; Van Ginhoven, Renee M.

2014-02-20

Density functional theory (DFT) is used to calculate the thermodynamic and kinetic properties of transmutant Mg in 3C-SiC due to high-energy neutron irradiation associated with the fusion nuclear environment. The formation and binding energies of intrinsic defects, Mg-related defects, and clusters in 3C-SiC are systematically calculated. The minimum energy paths and activation energies during point defect migration and small cluster evolution are studied using a generalized solid-state elastic band (G-SSNEB) method with DFT energy calculations. Stable defect structures and possible defect migration mechanisms are identified. The evolution of binding energies during Mg2Si formation demonstrates that the formation of Mg2Si needsmore » to overcome a critical nucleus size and nucleation barrier. It is also found that a compressive stress field exists around the Mg2Si nucleus. These data are important inputs in meso- and macro-scale modeling and experiments to understand and predict the impact of Mg on phase stability, microstructure evolution, and performance of SiC and SiC-based materials during long-term neutron exposures.« less

Materials for DEMO and reactor applications—boundary conditions and new concepts

NASA Astrophysics Data System (ADS)

Coenen, J. W.; Antusch, S.; Aumann, M.; Biel, W.; Du, J.; Engels, J.; Heuer, S.; Houben, A.; Hoeschen, T.; Jasper, B.; Koch, F.; Linke, J.; Litnovsky, A.; Mao, Y.; Neu, R.; Pintsuk, G.; Riesch, J.; Rasinski, M.; Reiser, J.; Rieth, M.; Terra, A.; Unterberg, B.; Weber, Th; Wegener, T.; You, J.-H.; Linsmeier, Ch

2016-02-01

DEMO is the name for the first stage prototype fusion reactor considered to be the next step after ITER towards realizing fusion. For the realization of fusion energy especially, materials questions pose a significant challenge already today. Heat, particle and neutron loads are a significant problem to material lifetime when extrapolating to DEMO. For many of the issues faced, advanced materials solutions are under discussion or already under development. In particular, components such as the first wall and the divertor of the reactor can benefit from introducing new approaches such as composites or new alloys into the discussion. Cracking, oxidation as well as fuel management are driving issues when deciding for new materials. Here {{{W}}}{{f}}/{{W}} composites as well as strengthened CuCrZr components together with oxidation resilient tungsten alloys allow the step towards a fusion reactor. In addition, neutron induced effects such as transmutation, embrittlement and after-heat and activation are essential. Therefore, when designing a component an approach taking into account all aspects is required.

New Insight into Nuclear Reactions in Solids

NASA Astrophysics Data System (ADS)

Miley, George H.

2003-04-01

Earlier work by the author disclosed evidence for nuclear transmutations in multi-layer thin-film Ni/Pd electrodes loaded to a high ratio of hydrogen/film metal using an electrolytic technique [1]. Non-natural isotopes abundances were found for select products. A distinctive characteristic of this and similar experiments by others is a product yield curve vs. mass with four high yield peaks distributed between low and high masses. Attempts to explain this observation have evolved around the original swimming electron layer (SEL) theory [2]. In addition, CR-39 track detector measurements have revealed low-level emission of 1.6 MeV protons and 16 MeV alpha particles from the front face of the thin film electrodes during runs [3]. Most recently Mitsubishi Corp. researchers have reported a real-time transmutation measurement using built-in XPS diagnostics where a surface layer of Sr-88 was transmuted into Mo-96 over a 200 hour run period during the diffusion of deuterium through a multi-layer thin-film Pd/CaO substrate [4]. Likewise in a companion experiment, Cs-133 was transmuted into Pr-141. These products exhibit a large deviation from natural isotopic abundance, and the characteristic signature is a mass change of 8 and charge change of 4. These various phenomena along with a preliminary theory involving SEL and orbital mixing will be presented. The objective is to provide a unified understanding of both types of experiments presented in Refs. 1 and 3. [1] G.H. Miley and J. A. Patterson, "Nuclear Transmutations in Thin-Film Nickel Coatings Undergoing Electrolysis," J. New Energy, 1, 3, 5-30 (1996). [2] H. Hora, et al., "Screening in Cold Fusion Derived from D D Reactions," Physics Ltrs. A, 175, 138-143, (1993). [3] A. Lipson, et al., "In-situ long - range alpha particles and X-ray detection in Pd thin film-cathodes during electrolysis in, Li2SO4/H2O, Bult. APS, 47, 1,Pt. II, 1219, Indianapolis, (2002). [4] Y. Iwamura, T. Itoh, et al., "Low energy nuclear reaction induced by D gas permeation through multilayer film," Japanese J. Physics, 41, pt. 1, 7A, 4642, (2002).

Transmutation approximations for the application of hybrid Monte Carlo/deterministic neutron transport to shutdown dose rate analysis

DOE Office of Scientific and Technical Information (OSTI.GOV)

Biondo, Elliott D.; Wilson, Paul P. H.

In fusion energy systems (FES) neutrons born from burning plasma activate system components. The photon dose rate after shutdown from resulting radionuclides must be quantified. This shutdown dose rate (SDR) is calculated by coupling neutron transport, activation analysis, and photon transport. The size, complexity, and attenuating configuration of FES motivate the use of hybrid Monte Carlo (MC)/deterministic neutron transport. The Multi-Step Consistent Adjoint Driven Importance Sampling (MS-CADIS) method can be used to optimize MC neutron transport for coupled multiphysics problems, including SDR analysis, using deterministic estimates of adjoint flux distributions. When used for SDR analysis, MS-CADIS requires the formulation ofmore » an adjoint neutron source that approximates the transmutation process. In this work, transmutation approximations are used to derive a solution for this adjoint neutron source. It is shown that these approximations are reasonably met for typical FES neutron spectra and materials over a range of irradiation scenarios. When these approximations are met, the Groupwise Transmutation (GT)-CADIS method, proposed here, can be used effectively. GT-CADIS is an implementation of the MS-CADIS method for SDR analysis that uses a series of single-energy-group irradiations to calculate the adjoint neutron source. For a simple SDR problem, GT-CADIS provides speedups of 200 100 relative to global variance reduction with the Forward-Weighted (FW)-CADIS method and 9 ± 5 • 104 relative to analog. As a result, this work shows that GT-CADIS is broadly applicable to FES problems and will significantly reduce the computational resources necessary for SDR analysis.« less

Transmutation approximations for the application of hybrid Monte Carlo/deterministic neutron transport to shutdown dose rate analysis

DOE PAGES

Biondo, Elliott D.; Wilson, Paul P. H.

2017-05-08

In fusion energy systems (FES) neutrons born from burning plasma activate system components. The photon dose rate after shutdown from resulting radionuclides must be quantified. This shutdown dose rate (SDR) is calculated by coupling neutron transport, activation analysis, and photon transport. The size, complexity, and attenuating configuration of FES motivate the use of hybrid Monte Carlo (MC)/deterministic neutron transport. The Multi-Step Consistent Adjoint Driven Importance Sampling (MS-CADIS) method can be used to optimize MC neutron transport for coupled multiphysics problems, including SDR analysis, using deterministic estimates of adjoint flux distributions. When used for SDR analysis, MS-CADIS requires the formulation ofmore » an adjoint neutron source that approximates the transmutation process. In this work, transmutation approximations are used to derive a solution for this adjoint neutron source. It is shown that these approximations are reasonably met for typical FES neutron spectra and materials over a range of irradiation scenarios. When these approximations are met, the Groupwise Transmutation (GT)-CADIS method, proposed here, can be used effectively. GT-CADIS is an implementation of the MS-CADIS method for SDR analysis that uses a series of single-energy-group irradiations to calculate the adjoint neutron source. For a simple SDR problem, GT-CADIS provides speedups of 200 100 relative to global variance reduction with the Forward-Weighted (FW)-CADIS method and 9 ± 5 • 104 relative to analog. As a result, this work shows that GT-CADIS is broadly applicable to FES problems and will significantly reduce the computational resources necessary for SDR analysis.« less

A flowsheet concept for an Am/Ln separation based on Am{sup VI} solvent extraction

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mincher, B.J.; Law, J.D.

2013-07-01

The separation of Am from the lanthanides and curium is a key step in proposed advanced fuel cycle scenarios. The partitioning and transmutation of Am is desirable to minimize the long-term radiotoxicity of material interred in a future high-level waste repository. However, a separation amenable to process scale-up remains elusive. Higher oxidation states of americium have recently been used to demonstrate solvent extraction-based separations using conventional fuel cycle ligands. Here, the successful partitioning of Am{sup VI} from the bulk of lanthanides and curium using diamyl-amyl-phosphonate (DAAP) extraction is reported. Due to the instability of Am{sup VI} in the organic phasemore » it was readily selectively stripped to a new acidic aqueous phase to provide separation from co-extracted Ce{sup IV}. The use of NaBiO{sub 3} as an oxidant to separate Am from the lanthanides and Cm by solvent extraction has been successfully demonstrated on the bench scale. Based on these results, flowsheet concepts can be designed that result in 96 % Am recovery in the presence of a few percent of the remaining Cm and the lanthanides in two extraction contacts. Preliminary results also indicate that the DAAP extractant is robust toward γ- irradiation under realistic conditions of acidity and dissolved oxygen concentration.« less

User Guide for VISION 3.4.7 (Verifiable Fuel Cycle Simulation) Model

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jacob J. Jacobson; Robert F. Jeffers; Gretchen E. Matthern

2011-07-01

The purpose of this document is to provide a guide for using the current version of the Verifiable Fuel Cycle Simulation (VISION) model. This is a complex model with many parameters and options; the user is strongly encouraged to read this user guide before attempting to run the model. This model is an R&D work in progress and may contain errors and omissions. It is based upon numerous assumptions. This model is intended to assist in evaluating 'what if' scenarios and in comparing fuel, reactor, and fuel processing alternatives at a systems level. The model is not intended as amore » tool for process flow and design modeling of specific facilities nor for tracking individual units of fuel or other material through the system. The model is intended to examine the interactions among the components of a fuel system as a function of time varying system parameters; this model represents a dynamic rather than steady-state approximation of the nuclear fuel system. VISION models the nuclear cycle at the system level, not individual facilities, e.g., 'reactor types' not individual reactors and 'separation types' not individual separation plants. Natural uranium can be enriched, which produces enriched uranium, which goes into fuel fabrication, and depleted uranium (DU), which goes into storage. Fuel is transformed (transmuted) in reactors and then goes into a storage buffer. Used fuel can be pulled from storage into either separation or disposal. If sent to separations, fuel is transformed (partitioned) into fuel products, recovered uranium, and various categories of waste. Recycled material is stored until used by its assigned reactor type. VISION is comprised of several Microsoft Excel input files, a Powersim Studio core, and several Microsoft Excel output files. All must be co-located in the same folder on a PC to function. You must use Powersim Studio 8 or better. We have tested VISION with the Studio 8 Expert, Executive, and Education versions. The Expert and Education versions work with the number of reactor types of 3 or less. For more reactor types, the Executive version is currently required. The input files are Excel2003 format (xls). The output files are macro-enabled Excel2007 format (xlsm). VISION 3.4 was designed with more flexibility than previous versions, which were structured for only three reactor types - LWRs that can use only uranium oxide (UOX) fuel, LWRs that can use multiple fuel types (LWR MF), and fast reactors. One could not have, for example, two types of fast reactors concurrently. The new version allows 10 reactor types and any user-defined uranium-plutonium fuel is allowed. (Thorium-based fuels can be input but several features of the model would not work.) The user identifies (by year) the primary fuel to be used for each reactor type. The user can identify for each primary fuel a contingent fuel to use if the primary fuel is not available, e.g., a reactor designated as using mixed oxide fuel (MOX) would have UOX as the contingent fuel. Another example is that a fast reactor using recycled transuranic (TRU) material can be designated as either having or not having appropriately enriched uranium oxide as a contingent fuel. Because of the need to study evolution in recycling and separation strategies, the user can now select the recycling strategy and separation technology, by year.« less

Correlations, soliton modes, and non-Hermitian linear mode transmutation in the one-dimensional noisy Burgers equation.

PubMed

Fogedby, Hans C

2003-08-01

Using the previously developed canonical phase space approach applied to the noisy Burgers equation in one dimension, we discuss in detail the growth morphology in terms of nonlinear soliton modes and superimposed linear modes. We moreover analyze the non-Hermitian character of the linear mode spectrum and the associated dynamical pinning, and mode transmutation from diffusive to propagating behavior induced by the solitons. We discuss the anomalous diffusion of growth modes, switching and pathways, correlations in the multisoliton sector, and in detail the correlations and scaling properties in the two-soliton sector.

Dimensional Transmutation by Monopole Condensation in QCD

NASA Astrophysics Data System (ADS)

Cho, Y. M.

2015-01-01

The dimensional transmutation by the monopole condensation in QCD is reviewed. Using Abelian projection of the gauge potential which projects out the monopole potential gauge independently, we we show that there are two types of gluons: the color neutral binding gluons which plays the role of the confining agent and the colored valence gluons which become confined prisoners. With this we calculate the one-loop QCD effective potential and show the monopole condensation becomes the true vacuum of QCD. We propose to test the existence of two types of gluons experimentally by re-analyzing the existing gluon jets data.

``Recycling'' Nuclear Power Plant Waste: Technical Difficulties and Proliferation Concerns

NASA Astrophysics Data System (ADS)

Lyman, Edwin

2007-04-01

One of the most vexing problems associated with nuclear energy is the inability to find a technically and politically viable solution for the disposal of long-lived radioactive waste. The U.S. plan to develop a geologic repository for spent nuclear fuel at Yucca Mountain in Nevada is in jeopardy, as a result of managerial incompetence, political opposition and regulatory standards that may be impossible to meet. As a result, there is growing interest in technologies that are claimed to have the potential to drastically reduce the amount of waste that would require geologic burial and the length of time that the waste would require containment. A scenario for such a vision was presented in the December 2005 Scientific American. While details differ, these technologies share a common approach: they require chemical processing of spent fuel to extract plutonium and other long-lived actinide elements, which would then be ``recycled'' into fresh fuel for advanced reactors and ``transmuted'' into shorter-lived fission products. Such a scheme is the basis for the ``Global Nuclear Energy Partnership,'' a major program unveiled by the Department of Energy (DOE) in early 2006. This concept is not new, but has been studied for decades. Major obstacles include fundamental safety issues, engineering feasibility and cost. Perhaps the most important consideration in the post-9/11 era is that these technologies involve the separation of plutonium and other nuclear weapon-usable materials from highly radioactive fission products, providing opportunities for terrorists seeking to obtain nuclear weapons. While DOE claims that it will only utilize processes that do not produce ``separated plutonium,'' it has offered no evidence that such technologies would effectively deter theft. It is doubtful that DOE's scheme can be implemented without an unacceptable increase in the risk of nuclear terrorism.

Studies on Materials for Heavy-Liquid-Metal-Cooled Reactors in Japan

DOE Office of Scientific and Technical Information (OSTI.GOV)

Minoru Takahashi; Masayuki Igashira; Toru Obara

2002-07-01

Recent studies on materials for the development of lead-bismuth (Pb-Bi)-cooled fast reactors (FR) and accelerator-driven sub-critical systems (ADS) in Japan are reported. The measurement of the neutron cross section of Bi to produce {sup 210}Po, the removal experiment of Po contamination and steel corrosion test in Pb-Bi flow were performed in Tokyo Institute of Technology. A target material corrosion test was performed in the project of Transmutation Experimental Facility for ADS in Japan Atomic Energy Research Institute (JAERI). Steel corrosion test was started in Mitsui Engineering and Shipbuilding Co., LTD (MES). The feasibility study for FR cycle performed in Japanmore » Nuclear Cycle Institute (JNC) are described. (authors)« less

Testing of 100 mK bolometers for space applications

NASA Technical Reports Server (NTRS)

Murray, A. G.; Ade, P. A. R.; Bhatia, R. S.; Griffin, M. J.; Maffei, B.; Nartallo, R.; Beeman, J. W.; Bock, J.; Lange, A.; DelCastillo, H.

1996-01-01

Electrical and optical performance data are presented for a prototype 100 mK spider-web bolometer operating under very low photon backgrounds. These data are compared with the bolometer theory and are used to estimate the expected sensitivity of such a detector used for low background space astronomy. The results demonstrate that the sensitivity and speed of response requirements of the bolometer instruments proposed for these missions can be met by 100 mK spider-web bolometers using neutron transmutation doped germanium as the temperature sensitive element.

High-temperature annealing of proton irradiated beryllium – A dilatometry-based study

DOE PAGES

Simos, Nikolaos; Elbakhshwan, Mohamed; Zhong, Zhong; ...

2016-04-07

S—200 F grade beryllium has been irradiated with 160 MeV protons up to 1.2 10 20 cm –2 peak fluence and irradiation temperatures in the range of 100–200 °C. To address the effect of proton irradiation on dimensional stability, an important parameter in its consideration in fusion reactor applications, and to simulate high temperature irradiation conditions, multi-stage annealing using high precision dilatometry to temperatures up to 740 °C were conducted in air. X-ray diffraction studies were also performed to compliment the macroscopic thermal study and offer a microscopic view of the irradiation effects on the crystal lattice. The primary objectivemore » was to qualify the competing dimensional change processes occurring at elevated temperatures namely manufacturing defect annealing, lattice parameter recovery, transmutation 4He and 3H diffusion and swelling and oxidation kinetics. Further, quantification of the effect of irradiation dose and annealing temperature and duration on dimensional changes is sought. Here, the study revealed the presence of manufacturing porosity in the beryllium grade, the oxidation acceleration effect of irradiation including the discontinuous character of oxidation advancement, the effect of annealing duration on the recovery of lattice parameters recovery and the triggering temperature for transmutation gas diffusion leading to swelling.« less

Transmutation of uranium and thorium in the particle field of the Quinta sub-critical assembly

NASA Astrophysics Data System (ADS)

Hashemi-Nezhad, S. R.; Asquith, N. L.; Voronko, V. A.; Sotnikov, V. V.; Zhadan, Alina; Zhuk, I. V.; Potapenko, A.; Husak, Krystsina; Chilap, V.; Adam, J.; Baldin, A.; Berlev, A.; Furman, W.; Kadykov, M.; Khushvaktov, J.; Kudashkin, I.; Mar'in, I.; Paraipan, M.; Pronskih, V.; Solnyshkin, A.; Tyutyunnikov, S.

2018-03-01

The fission rates of natural uranium and thorium were measured in the particle field of Quinta, a 512 kg natural uranium target-blanket sub-critical assembly. The Quinta assembly was irradiated with deuterons of energy 4 GeV from the Nuclotron accelerator of the Joint Institute for Nuclear Research (JINR), Dubna, Russia. Fission rates of uranium and thorium were measured using Gamma spectroscopy and fission track techniques. The production rate of 239Np was also measured. The obtained experimental results were compared with Monte Carlo predictions using the MCNPX 2.7 code employing the physics and fission-evaporation models of INCL4-ABLA, CEM03.03 and LAQGSM03.03. Some of the neutronic characteristics of the Quinta are compared with the "Energy plus Transmutation (EpT)" subcritical assembly, which is composed of a lead target and natU blanket. This comparison clearly demonstrates the importance of target material, neutron moderator and reflector types on the performance of a spallation neutron driven subcritical system. As the dimensions of the Quinta are very close to those of an optimal multi-rod-uranium target, the experimental and Monte Carlo calculation results presented in this paper provide insights on the particle field within a uranium target as well as in Accelerator Driven Systems in general.

Neutron transmutation doping of silicon in the SAFARI-1 research reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Louw, P.A.; Robertson, D.G.; Strydom, W.J.

1994-12-31

The SAFARI-1 research reactor has operated with an exemplary safety record since commissioning in 1965. As part of a commercialisation effort a silicon irradiation facility (SILIRAD) has been installed in the poolside region of SAFARI-1 for Neutron Transmutation Doping (NTD) of silicon. Commissioning of the facility took place in the last quarter of 1992 with a series of trial irradiations which were performed in close collaboration with Wacker Chemitronic of Germany. A methodology for the determination of irradiation times necessary to achieve the target resistivities was verified on the basis of the results from the trial irradiations. All production activitiesmore » are controlled by quality assurance procedures. To date some hundred and twelve silicon ingots (103 mm diameter) have been successfully irradiated on a commercial contract basis. The observed axial and radial variations in the resistivity profile of the ingots are very small compared to the profiles associated with conventionally doped silicon and small tolerances on target resistivities are attained. In this paper an overview of the design and characterisation of SILIRAD is given and the methods applied that ensure a quality product are described. Results obtained from trial and production irradiations are presented and the envisaged future modifications to SILIRAD discussed.« less

TiO2 films with rich bulk oxygen vacancies prepared by electrospinning for dye-sensitized solar cells

NASA Astrophysics Data System (ADS)

Li, Xiaodong; Gao, Caitian; Wang, Jiangtao; Lu, Bingan; Chen, Wanjun; Song, Jie; Zhang, Shanshan; Zhang, Zhenxing; Pan, Xiaojun; Xie, Erqing

2012-09-01

Highly transparent nanocrystalline TiO2 films have been fabricated by electrospinning (ES) technique based on a transmutation process from as-spun nanofibers with an appropriate amount of tri-ethanolamine (TEOA) added to the precursor. A possible evolution mechanism of the transparent nanocrystalline TiO2 films is proposed. It is found that the films prepared via transmutation from electrospun nanofibers possess rich bulk oxygen vacancies (BOVs, PL band at 621-640 nm) by using photoluminescence (PL) spectroscopy. Contrastively, the dominant peak in PL spectrum of the spin-coated film is the emission from surface oxygen vacancies (SOVs, PL band at 537-555 nm). The electrospun TiO2 films with rich BOVs induce large open-circuit voltage (Voc) and fill factor (FF) improvements in dye-sensitized solar cells (DSCs), and thus a large improvement of energy conversion efficiency (η). In addition, these performance advantages are maintained for a double-layer cell with a doctor-bladed ˜7 μm top layer (P25 nanometer TiO2, Degussa) and an electrospun ˜3 μm bottom layer. The double-layer cell yields a high η of 6.01%, which has increased by 14% as compared with that obtained from a 10 μm thick P25 film.

Muonic alchemy: Transmuting elements with the inclusion of negative muons

NASA Astrophysics Data System (ADS)

Moncada, Félix; Cruz, Daniel; Reyes, Andrés

2012-06-01

In this Letter we present a theoretical study of atoms in which one electron has been replaced by a negative muon. We have treated these muonic systems with the Any Particle Molecular Orbital (APMO) method. A comparison between the electronic and muonic radial distributions revealed that muons are much more localized than electrons. Therefore, the muonic cloud is screening effectively one positive charge of the nucleus. Our results have revealed that by replacing an electron in an atom by a muon there is a transmutation of the electronic properties of that atom to those of the element with atomic number Z - 1.

Statistical Transmutation in Floquet Driven Optical Lattices.

PubMed

Sedrakyan, Tigran A; Galitski, Victor M; Kamenev, Alex

2015-11-06

We show that interacting bosons in a periodically driven two dimensional (2D) optical lattice may effectively exhibit fermionic statistics. The phenomenon is similar to the celebrated Tonks-Girardeau regime in 1D. The Floquet band of a driven lattice develops the moat shape, i.e., a minimum along a closed contour in the Brillouin zone. Such degeneracy of the kinetic energy favors fermionic quasiparticles. The statistical transmutation is achieved by the Chern-Simons flux attachment similar to the fractional quantum Hall case. We show that the velocity distribution of the released bosons is a sensitive probe of the fermionic nature of their stationary Floquet state.

Detection of endogenous lithium in neuropsychiatric disorders--a model for biological transmutation.

PubMed

Kurup, Ravi Kumar; Kurup, Parameswara Achutha

2002-01-01

The human hypothalamus produces an endogenous membrane Na(+)-K(+) ATPase inhibitor, digoxin. A digoxin induced model of cellular/neuronal quantal state and perception has been described by the authors. Biological transmutation has been described in microbial systems in the quantal state. The study focuses on the plasma levels of digoxin, RBC membrane Na(+)-K(+) ATPase activity, plasma levels of magnesium and lithium in neuropsychiatric and systemic disorders. Inhibition of RBC membrane Na(+)-K(+) ATPase activity was observed in most cases along with an increase in the levels of serum digoxin and lithium and a decrease in the level of serum Mg(++). The generation of endogenous lithium would obviously occur due to biological transmutation from magnesium. Digoxin and lithium together can produce added membrane Na(+)-K(+) ATPase inhibition. The role of membrane Na(+)-K(+) ATPase inhibition in the pathogenesis of neuropsychiatric and systemic disorders is discussed. The inhibition of membrane Na(+)-K(+) ATPase can contribute to an increase in intracellular calcium and a decrease in magnesium, which can result in a defective neurotransmitter transport mechanism, mitochondrial dysfunction and apoptosis, defective golgi body function and protein processing dysfunction, immune dysfunction and oncogenesis. Copyright 2002 John Wiley & Sons, Ltd.

The Aspiring Adept: Robert Boyle and His Alchemical Quest (Lawrence M. Principe)

NASA Astrophysics Data System (ADS)

Kovac, Jeffrey

1999-10-01

Robert Boyle is widely regarded as the Father of Modern Chemistry, who broke once and for all from the irrational, misguided alchemy that preceded him. One of the goals of this carefully researched and argued new book by Lawrence M. Principe, Assistant Professor in the Department of Chemistry and the Institute for the History of Science, Medicine, and Technology at The Johns Hopkins University, is to refute the two errors in this characterization of Boyle and to understand his life, thought, and work in the intellectual and social context of his time. This book is not for the casual reader; it is a detailed scholarly treatise in the history of science, but it provides a fresh and interesting perspective on Boyle and on the development of chemistry in the 17th century. Boyle is usually characterized as a modern scientist and his most famous book, The Skeptical Chymist, as a critique of traditional alchemy. Principe demonstrates that this characterization is based on a selective and sometimes incorrect reading of Boyle's works. Like Newton, Boyle was deeply involved in traditional transmutational alchemy, reading the works of other alchemists, performing experiments, and even witnessing transmutations. Alchemy, however, was not a monolith and Boyle adhered to what Principe tentatively identifies as a uniquely English school of supernatural alchemy. According to Principe, The Skeptical Chymist was mainly a criticism of the Paracelsians interested in chemical medicine rather than a defense of what we would now regard as modern chemistry. To further support his characterization of Boyle and to better reveal Boyle's involvement in alchemyparticularly the transmutation of base metals to gold, termed chrysopoeiaPrincipe has reconstructed from some 20 fragments one of Boyle's alchemical manuscripts, his Dialogue on the Transmutation of Metals. The full text of this lost work is included as Appendix 1. Two other primary sources, Interview Accounts of Transmutations and Prefaces to Boyle's Other Chrysopoetic Writings, and Robert Boyle's Dialogue on the Converse with Angels Aided by the Philosophers Stone, are also printed, as Appendices 2 and 3. Both Robert Boyle and 17th-century chemistry emerge from this book as exceedingly complex. The development of chemistry cannot be regarded as either straightforward or linear. As Boyle's work exemplifies, ideas from traditional alchemy were important in the development of chemistry. For example, Boyle's corpuscularian hypothesis is partly derived from the alchemical corpuscularian traditionthe minima of Geber. Alchemy was part of the intellectual context in which both Boyle and Newton were raised and it played an important part in their thinking. For Boyle, alchemy was also closely linked to Christianity. He regarded the philosophers' stone as a powerful weapon against the growing atheism of his time. The possessor of the stone could summon angels and other spirits, thus providing support for crucial theological truths. Religion was as important a motivation as both natural philosophy and the potential for the development of potent medicines for Boyle's study of alchemy. This book is a rich source of information on Robert Boyle, alchemy, and the development of 17th century chemistry. It is not an easy book, however; it requires close attention and some background in the history of science. For those interested in the development of modern science, this is a valuable addition to the growing collection of excellent new books.

Brief overview of the long-lived radionuclide separation processes developed in france in connection with the spin program

NASA Astrophysics Data System (ADS)

Madic, Charles; Bourges, Jacques; Dozol, Jean-François

1995-09-01

To reduce the long-term potential hazards associated with the management of nuclear wastes generated by nuclear fuel reprocessing, one alternative is the transmutation of long-lived radionuclides into short-lived radionuclides by nuclear means (P & T strategy). In this context, according to the law passed by the French Parliament on 30 December 1991, the CEA launched the SPIN program for the design of long-lived radionuclide separation and nuclear incineration processes. The research in progress to define separation processes focused mainly on the minor actinides (neptunium, americium and curium) and some fission products, like cesium and technetium. To separate these long-lived radionuclides, two strategies were developed. The first involves research on new operating conditions for improving the PUREX fuel reprocessing technology. This approach concerns the elements neptunium and technetium (iodine and zirconium can also be considered). The second strategy involves the design of new processes; DIAMEX for the co-extraction of minor actinides from the high-level liquid waste leaving the PUREX process, An(III)/Ln(III) separation using tripyridyltriazine derivatives or picolinamide extracting agents; SESAME for the selective separation of americium after its oxidation to Am(IV) or Am(VI) in the presence of a heteropolytungstate ligand, and Cs extraction using a new class of extracting agents, calixarenes, which exhibit exceptional Cs separation properties, especially in the presence of sodium ion. This lecture focuses on the latest achievements in these research areas.

Interim status report on lead-cooled fast reactor (LFR) research and development.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tzanos, C. P.; Sienicki, J. J.; Moisseytsev, A.

2008-03-31

This report discusses the status of Lead-Cooled Fast Reactor (LFR) research and development carried out during the first half of FY 2008 under the U.S. Department of Energy Generation IV Nuclear Energy Systems Initiative. Lead-Cooled Fast Reactor research and development has recently been transferred from Generation IV to the Reactor Campaign of the Global Nuclear Energy Partnership (GNEP). Another status report shall be issued at the end of FY 2008 covering all of the LFR activities carried out in FY 2008 for both Generation IV and GNEP. The focus of research and development in FY 2008 is an initial investigationmore » of a concept for a LFR Advanced Recycling Reactor (ARR) Technology Pilot Plant (TPP)/demonstration test reactor (demo) incorporating features and operating conditions of the European Lead-cooled SYstem (ELSY) {approx} 600 MWe lead (Pb)-cooled LFR preconceptual design for the transmutation of waste and central station power generation, and which would enable irradiation testing of advanced fuels and structural materials. Initial scoping core concept development analyses have been carried out for a 100 MWt core composed of sixteen open-lattice 20 by 20 fuel assemblies largely similar to those of the ELSY preconceptual fuel assembly design incorporating fuel pins with mixed oxide (MOX) fuel, central control rods in each fuel assembly, and cooled with Pb coolant. For a cycle length of three years, the core is calculated to have a conversion ratio of 0.79, an average discharge burnup of 108 MWd/kg of heavy metal, and a burnup reactivity swing of about 13 dollars. With a control rod in each fuel assembly, the reactivity worth of an individual rod would need to be significantly greater than one dollar which is undesirable for postulated rod withdrawal reactivity insertion events. A peak neutron fast flux of 2.0 x 10{sup 15} (n/cm{sup 2}-s) is calculated. For comparison, the 400 MWt Fast Flux Test Facility (FFTF) achieved a peak neutron fast flux of 7.2 x 10{sup 15} (n/cm{sup 2}-s) and the initially 563 MWt PHENIX reactor attained 2.0 x 10{sup 15} (n/cm{sup 2}-s) before one of three intermediate cooling loops was shut down due to concerns about potential steam generator tube failures. The calculations do not assume a test assembly location for advanced fuels and materials irradiation in place of a fuel assembly (e.g., at the center of the core); the calculations have not examined whether it would be feasible to replace the central assembly by a test assembly location. However, having only fifteen driver assemblies implies a significant effect due to perturbations introduced by the test assembly. The peak neutron fast flux is low compared with the fast fluxes previously achieved in FFTF and PHENIX. Furthermore, the peak neutron fluence is only about half of the limiting value (4 x 10{sup 23} n/cm{sup 2}) typically used for ferritic steels. The results thus suggest that a larger power level (e.g., 400 MWt) and a larger core would be better for a TPP based upon the ELSY fuel assembly design and which can also perform irradiation testing of advanced fuels and materials. In particular, a core having a higher power level and larger dimensions would achieve a suitable average discharge burnup, peak fast flux, peak fluence, and would support the inclusion of one or more test assembly locations. Participation in the Generation IV International Forum Provisional System Steering Committee for the LFR is being maintained throughout FY 2008. Results from the analysis of samples previously exposed to flowing lead-bismuth eutectic (LBE) in the DELTA loop are summarized and a model for the oxidation/corrosion kinetics of steels in heavy liquid metal coolants was applied to systematically compare the calculated long-term (i.e., following several years of growth) oxide layer thicknesses of several steels.« less

Mixtures of Charged Bosons Confined in Harmonic Traps and Bose-Einstein Condensation Mechanism for Low-Energy Nuclear Reactions and Transmutation Processes in Condensed Matters

NASA Astrophysics Data System (ADS)

Kim, Yeong E.; Zubarev, Alexander L.

2006-02-01

A mixture of two different species of positively charged bosons in harmonic traps is considered in the mean-field approximation. It is shown that depending on the ratio of parameters, the two components may coexist in same regions of space, in spite of the Coulomb repulsion between the two species. Application of this result is discussed for the generalization of the Bose-Einstein condensation mechanism for low-energy nuclear reaction (LENR) and transmutation processes in condensed matters. For the case of deutron-lithium (d + Li) LENR, the result indicates that (d + 6Li) reactions may dominate over (d + d) reactions in LENR experiments.

Unifying Theory of Low-Energy Nuclear Reaction and Transmutation Processes in Deuterated/hydrogenated Metals, Acoustic Cavitation, Glow Discharge, and Deuteron Beam Experiments

NASA Astrophysics Data System (ADS)

Kim, Yeong E.; Zubarev, Alexander L.

The most basic theoretical challenge for understanding low-energy nuclear reaction (LENR) and transmutation reaction (LETR) in condensed matters is to find mechanisms by which the large Coulomb barrier between fusing nuclei can be overcome. A unifying theory of LENR and LETR has been developed to provide possible mechanisms for the LENR and LETR processes in matters based on high-density nano-scale and micro-scale quantum plasmas. It is shown that recently developed theoretical models based on Bose-Einstein Fusion (BEF) mechanism and Quantum Plasma Nuclear Fusion (QPNF) mechanism are applicable to the results of many different types of LENR and LETR experiments.

Robert Boyle, Transmutation, and the History of Chemistry before Lavoisier: A Response to Kuhn.

PubMed

Newman, William R

2014-01-01

In an influential article of 1952, Thomas Kuhn argued that Robert Boyle had little or no influence on the subsequent development of chemistry. This essay challenges Kuhn's view on two fronts. First, it shows that Johann Joachim Becher developed his hierarchical matter theory under the influence of Boyle and then transmitted it to the founder of the phlogiston theory, G. E. Stahl. Second, this essay argues that transmutational matter theories were not necessarily opposed to the existence of stable chemical species, pace Kuhn. Boyle's corpuscular theory descended largely from the tradition of "chymical atomism," which often advocated both chrysopoeia and the reality of robust chemical substances.

Silver (Ag) Transport Mechanisms in TRISO Coated Particles: A Critical Review

DOE Office of Scientific and Technical Information (OSTI.GOV)

IJ van Rooyen; ML Dunzik-Gougar; PM van Rooyen

2014-05-01

Transport of 110mAg in the intact SiC layer of TRISO coated particles has been studied for approximately 30 years without arriving at a satisfactory explanation of the transport mechanism. In this paper the possible mechanisms postulated in previous experimental studies, both in-reactor and out-of reactor research environment studies are critically reviewed and of particular interest are relevance to very high temperature gas reactor operating and accident conditions. Among the factors thought to influence Ag transport are grain boundary stoichiometry, SiC grain size and shape, the presence of free silicon, nano-cracks, thermal decomposition, palladium attack, transmutation products, layer thinning and coatedmore » particle shape. Additionally new insight to nature and location of fission products has been gained via recent post irradiation electron microscopy examination of TRISO coated particles from the DOE’s fuel development program. The combined effect of critical review and new analyses indicates a direction for investigating possible the Ag transport mechanism including the confidence level with which these mechanisms may be experimentally verified.« less

Silver (Ag) Transport Mechanisms in TRISO coated particles: A Critical Review

DOE Office of Scientific and Technical Information (OSTI.GOV)

I J van Rooyen; J H Neethling; J A A Engelbrecht

2012-10-01

Transport of 110mAg in the intact SiC layer of TRISO coated particles has been studied for approximately 30 years without arriving at a satisfactory explanation of the transport mechanism. In this paper the possible mechanisms postulated in previous experimental studies, both in-reactor and out-of reactor research environment studies are critically reviewed and of particular interest are relevance to very high temperature gas reactor operating and accident conditions. Among the factors thought to influence Ag transport are grain boundary stoichiometry, SiC grain size and shape, the presence of free silicon, nano-cracks, thermal decomposition, palladium attack, transmutation products, layer thinning and coatedmore » particle shape. Additionally new insight to nature and location of fission products has been gained via recent post irradiation electron microscopy examination of TRISO coated particles from the DOE’s fuel development program. The combined effect of critical review and new analyses indicates a direction for investigating possible the Ag transport mechanism including the confidence level with which these mechanisms may be experimentally verified.« less

The measurement programme at the neutron time-of-flight facility n_TOF at CERN

NASA Astrophysics Data System (ADS)

Gunsing, F.; Aberle, O.; Andrzejewski, J.; Audouin, L.; Bécares, V.; Bacak, M.; Balibrea-Correa, J.; Barbagallo, M.; Barros, S.; Bečvář, F.; Beinrucker, C.; Belloni, F.; Berthoumieux, E.; Billowes, J.; Bosnar, D.; Brown, A.; Brugger, M.; Caamaño, M.; Calviño, F.; Calviani, M.; Cano-Ott, D.; Cardella, R.; Casanovas, A.; Castelluccio, D. M.; Cerutti, F.; Chen, Y. H.; Chiaveri, E.; Colonna, N.; Cortés-Giraldo, M. A.; Cortés, G.; Cosentino, L.; Damone, L. A.; Deo, K.; Diakaki, M.; Domingo-Pardo, C.; Dressler, R.; Dupont, E.; Durán, I.; Fernández-Domínguez, B.; Ferrari, A.; Ferreira, P.; Finocchiaro, P.; Frost, R. J. W.; Furman, V.; Ganesan, S.; García, A. R.; Gawlik, A.; Gheorghe, I.; Gilardoni, S.; Glodariu, T.; Gonçalves, I. F.; González, E.; Goverdovski, A.; Griesmayer, E.; Guerrero, C.; Göbel, K.; Harada, H.; Heftrich, T.; Heinitz, S.; Hernández-Prieto, A.; Heyse, J.; Jenkins, D. G.; Jericha, E.; Käppeler, F.; Kadi, Y.; Kalamara, A.; Katabuchi, T.; Kavrigin, P.; Ketlerov, V.; Khryachkov, V.; Kimura, A.; Kivel, N.; Kokkoris, M.; Krtička, M.; Kurtulgil, D.; Leal-Cidoncha, E.; Lederer, C.; Leeb, H.; Lerendegui, J.; Licata, M.; Meo, S. Lo; Lonsdale, S. J.; Losito, R.; Macina, D.; Marganiec, J.; Martínez, T.; Masi, A.; Massimi, C.; Mastinu, P.; Mastromarco, M.; Matteucci, F.; Maugeri, E. A.; Mazzone, A.; Mendoza, E.; Mengoni, A.; Milazzo, P. M.; Mingrone, F.; Mirea, M.; Montesano, S.; Musumarra, A.; Nolte, R.; Negret, A.; Oprea, A.; Palomo-Pinto, F. R.; Paradela, C.; Patronis, N.; Pavlik, A.; Perkowski, J.; Porras, I.; Praena, J.; Quesada, J. M.; Radeck, D.; Rajeev, K.; Rauscher, T.; Reifarth, R.; Riego-Perez, A.; Robles, M.; Rout, P.; Rubbia, C.; Ryan, J. A.; Sabaté-Gilarte, M.; Saxena, A.; Schillebeeckx, P.; Schmidt, S.; Schumann, D.; Sedyshev, P.; Smith, A. G.; Sosnin, N. V.; Stamatopoulos, A.; Suryanarayana, S. V.; Tagliente, G.; Tain, J. L.; Tarifeño-Saldivia, A.; Tarrío, D.; Tassan-Got, L.; Tsinganis, A.; Valenta, S.; Vannini, G.; Variale, V.; Vaz, P.; Ventura, A.; Vlachoudis, V.; Vlastou, R.; Wallner, A.; Warren, S.; Weigand, M.; Weiss, C.; Wolf, C.; Woods, P. J.; Wright, T.; Žugec, P.

2017-09-01

Neutron-induced reaction cross sections are important for a wide variety of research fields ranging from the study of nuclear level densities, nucleosynthesis to applications of nuclear technology like design, and criticality and safety assessment of existing and future nuclear reactors, radiation dosimetry, medical applications, nuclear waste transmutation, accelerator-driven systems and fuel cycle investigations. Simulations and calculations of nuclear technology applications largely rely on evaluated nuclear data libraries. The evaluations in these libraries are based both on experimental data and theoretical models. CERN's neutron time-of-flight facility n_TOF has produced a considerable amount of experimental data since it has become fully operational with the start of its scientific measurement programme in 2001. While for a long period a single measurement station (EAR1) located at 185 m from the neutron production target was available, the construction of a second beam line at 20 m (EAR2) in 2014 has substantially increased the measurement capabilities of the facility. An outline of the experimental nuclear data activities at n_TOF will be presented.

Multiple rod-cone and cone-rod photoreceptor transmutations in snakes: evidence from visual opsin gene expression.

PubMed

Simões, Bruno F; Sampaio, Filipa L; Loew, Ellis R; Sanders, Kate L; Fisher, Robert N; Hart, Nathan S; Hunt, David M; Partridge, Julian C; Gower, David J

2016-01-27

In 1934, Gordon Walls forwarded his radical theory of retinal photoreceptor 'transmutation'. This proposed that rods and cones used for scotopic and photopic vision, respectively, were not fixed but could evolve into each other via a series of morphologically distinguishable intermediates. Walls' prime evidence came from series of diurnal and nocturnal geckos and snakes that appeared to have pure-cone or pure-rod retinas (in forms that Walls believed evolved from ancestors with the reverse complement) or which possessed intermediate photoreceptor cells. Walls was limited in testing his theory because the precise identity of visual pigments present in photoreceptors was then unknown. Subsequent molecular research has hitherto neglected this topic but presents new opportunities. We identify three visual opsin genes, rh1, sws1 and lws, in retinal mRNA of an ecologically and taxonomically diverse sample of snakes central to Walls' theory. We conclude that photoreceptors with superficially rod- or cone-like morphology are not limited to containing scotopic or photopic opsins, respectively. Walls' theory is essentially correct, and more research is needed to identify the patterns, processes and functional implications of transmutation. Future research will help to clarify the fundamental properties and physiology of photoreceptors adapted to function in different light levels. © 2016 The Author(s).

Transmutation effects on long-term Cs retention in phyllosilicate minerals from first principles.

PubMed

Sassi, Michel; Okumura, Masahiko; Machida, Masahiko; Rosso, Kevin M

2017-10-11

The accidental release and incorporation of radiocesium into soil minerals represents a massive environmental, technical and social challenge. Accurately forecasting the evolving distribution and fate of long- and medium-lived isotopes such as 137 Cs and 134 Cs over decadal time scales is essential. The cesium cation has long been modeled as a strongly and selectively sorbed species into clay mineral interlayers; however, because of the time scales involved by the radioisotopes half-lives, the effects of radioactive decay on Cs retention have been unknown. We report density functional theory (DFT) simulations of transmutation effects of radiocesium on long-term Cs retention in phlogopite. The calculations show that the progressive appearance of daughter product Ba 2+ is accompanied by a proportional increase in thermodynamic driving force to preferentially discharge remaining Cs, both radioactive and stable, back into aqueous solution. Based on thermodynamic analysis, the findings indicate that radiocesium transmutation provides a mean to weaken the binding of Cs in phyllosilicate minerals, therefore potentially involving a premature re-release of Cs back into the environment. In the case where radiogenic Ba 2+ ions accumulate in the mineral, collateral effects would ultimately be an increase in the overall interlayer binding energy and a lower resorption capacity.

Irradiation tests of ITER candidate Hall sensors using two types of neutron spectra.

PubMed

Ďuran, I; Bolshakova, I; Viererbl, L; Sentkerestiová, J; Holyaka, R; Lahodová, Z; Bém, P

2010-10-01

We report on irradiation tests of InSb based Hall sensors at two irradiation facilities with two distinct types of neutron spectra. One was a fission reactor neutron spectrum with a significant presence of thermal neutrons, while another one was purely fast neutron field. Total neutron fluence of the order of 10(16) cm(-2) was accumulated in both cases, leading to significant drop of Hall sensor sensitivity in case of fission reactor spectrum, while stable performance was observed at purely fast neutron spectrum. This finding suggests that performance of this particular type of Hall sensors is governed dominantly by transmutation. Additionally, it further stresses the need to test ITER candidate Hall sensors under neutron flux with ITER relevant spectrum.

Design of Lead-Free Inorganic Halide Perovskites for Solar Cells via Cation-Transmutation.

PubMed

Zhao, Xin-Gang; Yang, Ji-Hui; Fu, Yuhao; Yang, Dongwen; Xu, Qiaoling; Yu, Liping; Wei, Su-Huai; Zhang, Lijun

2017-02-22

Hybrid organic-inorganic halide perovskites with the prototype material of CH 3 NH 3 PbI 3 have recently attracted intense interest as low-cost and high-performance photovoltaic absorbers. Despite the high power conversion efficiency exceeding 20% achieved by their solar cells, two key issues-the poor device stabilities associated with their intrinsic material instability and the toxicity due to water-soluble Pb 2+ -need to be resolved before large-scale commercialization. Here, we address these issues by exploiting the strategy of cation-transmutation to design stable inorganic Pb-free halide perovskites for solar cells. The idea is to convert two divalent Pb 2+ ions into one monovalent M + and one trivalent M 3+ ions, forming a rich class of quaternary halides in double-perovskite structure. We find through first-principles calculations this class of materials have good phase stability against decomposition and wide-range tunable optoelectronic properties. With photovoltaic-functionality-directed materials screening, we identify 11 optimal materials with intrinsic thermodynamic stability, suitable band gaps, small carrier effective masses, and low excitons binding energies as promising candidates to replace Pb-based photovoltaic absorbers in perovskite solar cells. The chemical trends of phase stabilities and electronic properties are also established for this class of materials, offering useful guidance for the development of perovskite solar cells fabricated with them.

Improving High-Temperature Measurements in Nuclear Reactors with Mo/Nb Thermocouples

NASA Astrophysics Data System (ADS)

Villard, J.-F.; Fourrez, S.; Fourmentel, D.; Legrand, A.

2008-10-01

Many irradiation experiments performed in research reactors are used to assess the effects of nuclear radiations on material or fuel sample properties, and are therefore a crucial stage in most qualification and innovation studies regarding nuclear technologies. However, monitoring these experiments requires accurate and reliable instrumentation. Among all measurement systems implemented in irradiation devices, temperature—and more particularly high-temperature (above 1000°C)—is a major parameter for future experiments related, for example, to the Generation IV International Forum (GIF) Program or the International Thermonuclear Experimental Reactor (ITER) Project. In this context, the French Commissariat à l’Energie Atomique (CEA) develops and qualifies innovative in-pile instrumentation for its irradiation experiments in current and future research reactors. Logically, a significant part of these research and development programs concerns the improvement of in-pile high-temperature measurements. This article describes the development and qualification of innovative high-temperature thermocouples specifically designed for in-pile applications. This key study has been achieved with technical contributions from the Thermocoax Company. This new kind of thermocouple is based on molybdenum and niobium thermoelements, which remain nearly unchanged by thermal neutron flux even under harsh nuclear environments, whereas typical high-temperature thermocouples such as Type C or Type S are altered by significant drifts caused by material transmutations under the same conditions. This improvement has a significant impact on the temperature measurement capabilities for future irradiation experiments. Details of the successive stages of this development are given, including the results of prototype qualification tests and the manufacturing process.

Measurements and analysis of leakage neutron spectra from multiple-slab sample assemblies comprising W,U,C, and CH2 with D-T neutron irradiation.

PubMed

Luo, F; Han, R; Chen, Z; Nie, Y; Sun, Q; Shi, F; Zhang, S; Tian, G; Song, L; Ruan, X; Ye, M Y

2018-07-01

The accelerator driven subcritical system (ADS) is regarded as a safe and clean nuclear power system, which can be used for the transmutation of nuclear waste and the breeding of nuclear fuel. In this study, in order to validate nuclear data and the neutron transportation performance of the materials related to ADS, we measured the leakage neutron spectra from multiple-slab sample assemblies using 14.8 MeV D-T neutrons. Two types of assemblies comprising A-1 (W+U+C+CH 2 ) and A-2 (U+C+CH 2 ) were both built up gradually starting with the first wall. The measured spectra were compared with those calculated using the Monte Carlo code neutron transport coed (MCNP)-4C. A comparison of the results showed that the experimental leakage neutron spectra for both A-1 or A-2 were reproduced well by the three evaluated nuclear data libraries with discrepancies of less than 15% (A-1) and 12% (A-2), except when below 3 MeV. For 2-cm and 5-cm uranium samples, the CENDL-3.1 calculations exhibited large discrepancies in the energy range of 2-8 MeV and above 13 MeV. Thus, the CENDL-3.1 library for uranium should be reevaluated, especially around this energy range. It was significant that the leakage neuron spectra changed clearly when the latest material layer was added during the building of assemblies A-1 and A-2. Copyright © 2018 Elsevier Ltd. All rights reserved.

Structural response of Nd-stabilized zirconia and its composite under extreme conditions of swift heavy ion irradiation

NASA Astrophysics Data System (ADS)

Nandi, Chiranjit; Grover, V.; Kulriya, P. K.; Poswal, A. K.; Prakash, Amrit; Khan, K. B.; Avasthi, D. K.; Tyagi, A. K.

2018-02-01

Inert matrix fuel concept for minor actinide transmutation proposes stabilized zirconia as the major component for inert matrix. The present study explores Nd-stabilized zirconia (Zr0.8Nd0.2O1.9; Nd as surrogate for Am) and its composites for radiation tolerance against fission fragments. The introduction of MgO in the composite with stabilised zirconia is performed from the point of view to enhance the thermal conductivity. The radiation damage is also compared with Nd-stabilized zirconia co-doped with Y3+ (Zr0.8Nd0.1Y0.1O1.9) in order to mimic doping of minor actinides in Y3+ containing stabilized zirconia (Nd as surrogate for Am). The compositions were synthesized by gel combustion followed by high temperature sintering and characterised by XRD, SEM and EDS. Irradiation was carried out by 120 MeV Au ions at various fluences and irradiation induced structural changes were probed by in-situ X-ray diffraction (XRD). XRD demonstrated the retention of crystallinity for all the three samples but the extent of the damage was found to be highly dependent on the nominal composition. It was observed that introduction of Y3+ along with Nd3+ to stabilize cubic zirconia imparted poorer radiation stability. On the other hand, formation of a CERCER composite of MgO with Nd-stabilised zirconia enhanced its behaviour against swift heavy ion irradiation. Investigating these compositions by XANES spectroscopy post irradiation did not show any change in local electronic structure of constituent ions.

EASY-II Renaissance: n, p, d, α, γ-induced Inventory Code System

NASA Astrophysics Data System (ADS)

Sublet, J.-Ch.; Eastwood, J. W.; Morgan, J. G.

2014-04-01

The European Activation SYstem has been re-engineered and re-written in modern programming languages so as to answer today's and tomorrow's needs in terms of activation, transmutation, depletion, decay and processing of radioactive materials. The new FISPACT-II inventory code development project has allowed us to embed many more features in terms of energy range: up to GeV; incident particles: alpha, gamma, proton, deuteron and neutron; and neutron physics: self-shielding effects, temperature dependence and covariance, so as to cover all anticipated application needs: nuclear fission and fusion, accelerator physics, isotope production, stockpile and fuel cycle stewardship, materials characterization and life, and storage cycle management. In parallel, the maturity of modern, truly general purpose libraries encompassing thousands of target isotopes such as TENDL-2012, the evolution of the ENDF-6 format and the capabilities of the latest generation of processing codes PREPRO, NJOY and CALENDF have allowed the activation code to be fed with more robust, complete and appropriate data: cross sections with covariance, probability tables in the resonance ranges, kerma, dpa, gas and radionuclide production and 24 decay types. All such data for the five most important incident particles (n, p, d, α, γ), are placed in evaluated data files up to an incident energy of 200 MeV. The resulting code system, EASY-II is designed as a functional replacement for the previous European Activation System, EASY-2010. It includes many new features and enhancements, but also benefits already from the feedback from extensive validation and verification activities performed with its predecessor.

Detection of spallation neutrons and protons using the (nat)Cd activation technique in transmutation experiments at Dubna.

PubMed

Manolopoulou, M; Stoulos, S; Fragopoulou, M; Brandt, R; Westmeier, W; Krivopustov, M; Sosnin, A; Zamani, M

2006-07-01

Various spallation sources have been used to transmute long-lived radioactive waste, mostly making use of the wide energy neutron fluence. In addition to neutrons, a large number of protons and gamma rays are also emitted from these sources. In this paper (nat)Cd is proved to be a useful activation detector for determining both thermal-epithermal neutron as well as secondary proton fluences. The fluences measured with (nat)Cd compared with other experimental data and calculations of DCM-DEM code were found to be in reasonable agreement. An accumulation of thermal-epithermal neutrons around the center of the target (i.e. after approx. 10 cm) and of secondary protons towards the end of the target is observed.

Transmutation of 129I and 237Np using spallation neutrons produced by 1.5, 3.7 and 7.4 GeV protons

NASA Astrophysics Data System (ADS)

Wan, J.-S.; Schmidt, Th.; Langrock, E.-J.; Vater, P.; Brandt, R.; Adam, J.; Bradnova, V.; Bamblevski, V. P.; Gelovani, L.; Gridnev, T. D.; Kalinnikov, V. G.; Krivopustov, M. I.; Kulakov, B. A.; Sosnin, A. N.; Perelygin, V. P.; Pronskikh, V. S.; Stegailov, V. I.; Tsoupko-Sitnikov, V. M.; Modolo, G.; Odoj, R.; Phlippen, P.-W.; Zamani-Valassiadou, M.; Adloff, J. C.; Debeauvais, M.; Hashemi-Nezhad, S. R.; Guo, S.-L.; Li, L.; Wang, Y.-L.; Dwivedi, K. K.; Zhuk, I. V.; Boulyga, S. F.; Lomonossova, E. M.; Kievitskaja, A. F.; Rakhno, I. L.; Chigrinov, S. E.; Wilson, W. B.

2001-05-01

Small samples of 129I and 237Np, two long-lived radwaste nuclides, were exposed to spallation neutron fluences from relatively small metal targets of lead and uranium, that were surrounded with a 6 cm thick paraffin moderator, and irradiated with 1.5, 3.7 and 7.4 GeV protons. The (n,γ) transmutation rates were determined for these nuclides. Conventional radiochemical La- and U-sensors and a variety of solid-state nuclear track detectors were irradiated simultaneously with secondary neutrons. Compared with results from calculations with well-known cascade codes (LAHET from Los Alamos and DCM/CEM from Dubna), the observed secondary neutron fluences are larger.

Developments and Tendencies in Fission Reactor Concepts

NASA Astrophysics Data System (ADS)

Adamov, E. O.; Fuji-Ie, Y.

This chapter describes, in two parts, new-generation nuclear energy systems that are required to be in harmony with nature and to make full use of nuclear resources. The issues of transmutation and containment of radioactive waste will also be addressed. After a short introduction to the first part, Sect. 58.1.2 will detail the requirements these systems must satisfy on the basic premise of peaceful use of nuclear energy. The expected designs themselves are described in Sect. 58.1.3. The subsequent sections discuss various types of advanced reactor systems. Section 58.1.4 deals with the light water reactor (LWR) whose performance is still expected to improve, which would extend its application in the future. The supercritical-water-cooled reactor (SCWR) will also be shortly discussed. Section 58.1.5 is mainly on the high temperature gas-cooled reactor (HTGR), which offers efficient and multipurpose use of nuclear energy. The gas-cooled fast reactor (GFR) is also included. Section 58.1.6 focuses on the sodium-cooled fast reactor (SFR) as a promising concept for advanced nuclear reactors, which may help both to achieve expansion of energy sources and environmental protection thus contributing to the sustainable development of mankind. The molten-salt reactor (MSR) is shortly described in Sect. 58.1.7. The second part of the chapter deals with reactor systems of a new generation, which are now found at the research and development (R&D) stage and in the medium term of 20-30 years can shape up as reliable, economically efficient, and environmentally friendly energy sources. They are viewed as technologies of cardinal importance, capable of resolving the problems of fuel resources, minimizing the quantities of generated radioactive waste and the environmental impacts, and strengthening the regime of nonproliferation of the materials suitable for nuclear weapons production. Particular attention has been given to naturally safe fast reactors with a closed fuel cycle (CFC) - as an advanced and promising reactor system that offers solutions to the above problems. The difference (not confrontation) between the approaches to nuclear power development based on the principles of “inherent safety” and “natural safety” is demonstrated.

High energy-density liquid rocket fuel performance

NASA Technical Reports Server (NTRS)

Rapp, Douglas C.

1990-01-01

A fuel performance database of liquid hydrocarbons and aluminum-hydrocarbon fuels was compiled using engine parametrics from the Space Transportation Engine Program as a baseline. Propellant performance parameters are introduced. General hydrocarbon fuel performance trends are discussed with respect to hydrogen-to-carbon ratio and heat of formation. Aluminum-hydrocarbon fuel performance is discussed with respect to aluminum metal loading. Hydrocarbon and aluminum-hydrocarbon fuel performance is presented with respect to fuel density, specific impulse, and propellant density specific impulse.

Proposal for New Experimental Tests of the Bose-Einstein Condensation Mechanism for Low-Energy Nuclear Reaction and Transmutation Processes in Deuterium Loaded - and Nano-Scale Cavities

NASA Astrophysics Data System (ADS)

Kim, Yeong E.; Koltick, David S.; Reifenberger, Ronald G.; Zubarev, Alexander L.

2006-02-01

Most of experimental results of low-energy nuclear reaction (LENR) reported so far cannot be reproduced on demand. There have been persistent experimental results indicating that the LENR and transmutation processes in condensed matters (LENRTPCM) are surface phenomena rather than bulk phenomena. Recently proposed Bose-Einstein condensation (BEC) mechanism may provide a suitable theoretical description of the surface phenomena. New experiments are proposed and described for testing the BEC mechanism for LENR and transmutation processes in micro- and nano-scale traps. (1) We propose the use of micro- or nano-porous conducting materials as a cathode in electrolysis experiments with heavy water with or without Li in order to stabilize the active surface spots and to enhance the effect for the purpose of improving the reproducibility of excess heat generation and nuclear emission. (2) We propose new experimental tests of the BEC mechanism by measuring the pressure and temperature dependence of LENR events using deuterium gas and these deuterated metals with or without Li. If the LENRTPCM are surface phenomena, the proposed use of micro-/nano-scale porous materials is expected to enhance and scale up the LENRTPCM effects by many order of magnitude, and thus may lead to better reproductivity and theoretical understanding of the phenomena.

An innovative way of thinking nuclear waste management - Neutron physics of a reactor directly operating on SNF.

PubMed

Merk, Bruno; Litskevich, Dzianis; Bankhead, Mark; Taylor, Richard J

2017-01-01

A solution for the nuclear waste problem is the key challenge for an extensive use of nuclear reactors as a major carbon free, sustainable, and applied highly reliable energy source. Partitioning and Transmutation (P&T) promises a solution for improved waste management. Current strategies rely on systems designed in the 60's for the massive production of plutonium. We propose an innovative strategic development plan based on invention and innovation described with the concept of developments in s-curves identifying the current boundary conditions, and the evolvable objectives. This leads to the ultimate, universal vision for energy production characterized by minimal use of resources and production of waste, while being economically affordable and safe, secure and reliable in operation. This vision is transformed into a mission for a disruptive development of the future nuclear energy system operated by burning of existing spent nuclear fuel (SNF) without prior reprocessing. This highly innovative approach fulfils the sustainability goals and creates new options for P&T. A proof on the feasibility from neutronic point of view is given demonstrating sufficient breeding of fissile material from the inserted SNF. The system does neither require new resources nor produce additional waste, thus it provides a highly sustainable option for a future nuclear system fulfilling the requests of P&T as side effect. In addition, this nuclear system provides enhanced resistance against misuse of Pu and a significantly reduced fuel cycle. However, the new system requires a demand driven rethinking of the separation process to be efficient.

An innovative way of thinking nuclear waste management – Neutron physics of a reactor directly operating on SNF

PubMed Central

Litskevich, Dzianis; Bankhead, Mark; Taylor, Richard J.

2017-01-01

A solution for the nuclear waste problem is the key challenge for an extensive use of nuclear reactors as a major carbon free, sustainable, and applied highly reliable energy source. Partitioning and Transmutation (P&T) promises a solution for improved waste management. Current strategies rely on systems designed in the 60’s for the massive production of plutonium. We propose an innovative strategic development plan based on invention and innovation described with the concept of developments in s-curves identifying the current boundary conditions, and the evolvable objectives. This leads to the ultimate, universal vision for energy production characterized by minimal use of resources and production of waste, while being economically affordable and safe, secure and reliable in operation. This vision is transformed into a mission for a disruptive development of the future nuclear energy system operated by burning of existing spent nuclear fuel (SNF) without prior reprocessing. This highly innovative approach fulfils the sustainability goals and creates new options for P&T. A proof on the feasibility from neutronic point of view is given demonstrating sufficient breeding of fissile material from the inserted SNF. The system does neither require new resources nor produce additional waste, thus it provides a highly sustainable option for a future nuclear system fulfilling the requests of P&T as side effect. In addition, this nuclear system provides enhanced resistance against misuse of Pu and a significantly reduced fuel cycle. However, the new system requires a demand driven rethinking of the separation process to be efficient. PMID:28749952

NTD-GE Based Microcalorimeter Performance

NASA Technical Reports Server (NTRS)

Bandler, Simon; Silver, Eric; Schnopper, Herbert; Murray, Stephen; Barbera, Marco; Madden, Norm; Landis, Don; Beeman, Jeff; Haller, Eugene; Tucker, Greg

2000-01-01

Our group has been developing x-ray microcalorimeters consisting of neutron transmutation doped (NTD) germanium thermistors attached to superconducting tin absorbers. We discuss the performance of single pixel x-ray detectors, and describe an array technology. In this paper we describe the read-out circuit that allows us to measure fast signals in our detectors as this will be important in understanding the primary cause of resolution broadening. We describe briefly a multiplexing scheme that allows a number of different calorimeters to be read out using a single JFET. We list the possible causes of broadening and give a description of the experiment which best demonstrates the cause of the primary broadening source. We mention our strategy for finding a suitable solution to this problem and describe briefly a technology for building arrays of these calorimeters.

From teosinte to maize: the catastrophic sexual transmutation.

PubMed

Iltis, H H

1983-11-25

An alternative to the theory that the ear of maize (Zea mays ssp. mays) evolved from a slender female ear of a Mexican annual teosinte holds that it was derived from the central spike of a male teosinte inflorescence (tassel) which terminates the primary lateral branches. This alternative hypothesis is more consistent with morphology and explains the anomalous lack of significant genetic and biochemical differences between these taxa. Maize, the only cereal with unisexual inflorescences, evolved through a sudden epigenetic sexual transmutation involving condensation of primary branches, which brought their tassels into the zone of female expression, leading to strong apical dominance and a catastrophic shift in nutrient allocation. Initially, this quantum change may have involved no new mutations, but rather genetic assimilation under human selection of an abnormality, perhaps environmentally triggered.

Detecting neutrons by forward recoil protons at the Energy & Transmutation facility: Detector development and calibration with 14.1-MeV neutrons

NASA Astrophysics Data System (ADS)

Afanasev, S.; Vishnevskiy, A.; Vishnevskiy, D.; Rogachev, A.; Tyutyunnikov, S.

2017-05-01

As part of the Energy & Transmutation project, we are developing a detector for neutrons with energies in the 10-100 MeV range emitted from the target irradiated by a charged-particle beam. The neutron is detected by measuring the time-of-flight and total kinetic energy of the forward-going recoil proton [1] knocked out at a small angle from a thin layer of plastic scintillator, which has to be selected against an intense background created by γ quanta, scattered neutrons, and charged particles. On the other hand, neutron energy has to be measured over the full range with no extra tuning of the detector operation regime. Initial measurements with a source of 14.1-MeV neutrons are reported.

Comparative performance evaluation of transform coding in image pre-processing

NASA Astrophysics Data System (ADS)

Menon, Vignesh V.; NB, Harikrishnan; Narayanan, Gayathri; CK, Niveditha

2017-07-01

We are in the midst of a communication transmute which drives the development as largely as dissemination of pioneering communication systems with ever-increasing fidelity and resolution. Distinguishable researches have been appreciative in image processing techniques crazed by a growing thirst for faster and easier encoding, storage and transmission of visual information. In this paper, the researchers intend to throw light on many techniques which could be worn at the transmitter-end in order to ease the transmission and reconstruction of the images. The researchers investigate the performance of different image transform coding schemes used in pre-processing, their comparison, and effectiveness, the necessary and sufficient conditions, properties and complexity in implementation. Whimsical by prior advancements in image processing techniques, the researchers compare various contemporary image pre-processing frameworks- Compressed Sensing, Singular Value Decomposition, Integer Wavelet Transform on performance. The paper exposes the potential of Integer Wavelet transform to be an efficient pre-processing scheme.

Clustering of transmutation elements tantalum, rhenium and osmium in tungsten in a fusion environment

NASA Astrophysics Data System (ADS)

You, Yu-Wei; Kong, Xiang-Shan; Wu, Xuebang; Liu, C. S.; Fang, Q. F.; Chen, J. L.; Luo, G.-N.

2017-08-01

The formation of transmutation solute-rich precipitates has been reported to seriously degrade the mechanical properties of tungsten in a fusion environment. However, the underlying mechanisms controlling the formation of the precipitates are still unknown. In this study, first-principles calculations are therefore performed to systemically determine the stable structures and binding energies of solute clusters in tungsten consisting of tantalum, rhenium and osmium atoms as well as irradiation-induced vacancies. These clusters are known to act as precursors for the formation of precipitates. We find that osmium can easily segregate to form clusters even in defect-free tungsten alloys, whereas extremely high tantalum and rhenium concentrations are required for the formation of clusters. Vacancies greatly facilitate the clustering of rhenium and osmium, while tantalum is an exception. The binding energies of vacancy-osmium clusters are found to be much higher than those of vacancy-tantalum and vacancy-rhenium clusters. Osmium is observed to strongly promote the formation of vacancy-rhenium clusters, while tantalum can suppress the formation of vacancy-rhenium and vacancy-osmium clusters. The local strain and electronic structure are analyzed to reveal the underlying mechanisms governing the cluster formation. Employing the law of mass action, we predict the evolution of the relative concentration of vacancy-rhenium clusters. This work presents a microscopic picture describing the nucleation and growth of solute clusters in tungsten alloys in a fusion reactor environment, and thereby explains recent experimental phenomena.

Polycrystalline silicon semiconducting material by nuclear transmutation doping

DOEpatents

Cleland, John W.; Westbrook, Russell D.; Wood, Richard F.; Young, Rosa T.

1978-01-01

A NTD semiconductor material comprising polycrystalline silicon having a mean grain size less than 1000 microns and containing phosphorus dispersed uniformly throughout the silicon rather than at the grain boundaries.

Alternate-Fueled Combustor-Sector Performance: Part A: Combustor Performance Part B: Combustor Emissions

NASA Technical Reports Server (NTRS)

Shouse, D. T.; Neuroth, C.; Henricks, R. C.; Lynch, A.; Frayne, C.; Stutrud, J. S.; Corporan, E.; Hankins, T.

2010-01-01

Alternate aviation fuels for military or commercial use are required to satisfy MIL-DTL-83133F(2008) or ASTM D 7566 (2010) standards, respectively, and are classified as drop-in fuel replacements. To satisfy legacy issues, blends to 50% alternate fuel with petroleum fuels are certified individually on the basis of feedstock. Adherence to alternate fuels and fuel blends requires smart fueling systems or advanced fuel-flexible systems, including combustors and engines without significant sacrifice in performance or emissions requirements. This paper provides preliminary performance (Part A) and emissions and particulates (Part B) combustor sector data for synthetic-parafinic-kerosene- (SPK-) type fuel and blends with JP-8+100 relative to JP-8+100 as baseline fueling.

Quantitative Evaluation of Management Courses: Part 1

ERIC Educational Resources Information Center

Cunningham, Cyril

1973-01-01

The author describes how he developed a method of evaluating and comparing management courses of different types and lengths by applying an ordinal system of relative values using a process of transmutation. (MS)

Spallation reaction study for fission products in nuclear waste: Cross section measurements for 137Cs and 90Sr on proton and deuteron

NASA Astrophysics Data System (ADS)

Wang, H.; Otsu, H.; Sakurai, H.; Ahn, D. S.; Aikawa, M.; Doornenbal, P.; Fukuda, N.; Isobe, T.; Kawakami, S.; Koyama, S.; Kubo, T.; Kubono, S.; Lorusso, G.; Maeda, Y.; Makinaga, A.; Momiyama, S.; Nakano, K.; Niikura, M.; Shiga, Y.; Söderström, P.-A.; Suzuki, H.; Takeda, H.; Takeuchi, S.; Taniuchi, R.; Watanabe, Ya.; Watanabe, Yu.; Yamasaki, H.; Yoshida, K.

2016-03-01

We have studied spallation reactions for the fission products 137Cs and 90Sr for the purpose of nuclear waste transmutation. The spallation cross sections on the proton and deuteron were obtained in inverse kinematics for the first time using secondary beams of 137Cs and 90Sr at 185 MeV/nucleon at the RIKEN Radioactive Isotope Beam Factory. The target dependence has been investigated systematically, and the cross-section differences between the proton and deuteron are found to be larger for lighter spallation products. The experimental data are compared with the PHITS calculation, which includes cascade and evaporation processes. Our results suggest that both proton- and deuteron-induced spallation reactions are promising mechanisms for the transmutation of radioactive fission products.

The Existence of a Designer Al=Al Double Bond in the LiAl2 H4- Cluster Formed by Electronic Transmutation.

PubMed

Lundell, Katie A; Zhang, Xinxing; Boldyrev, Alexander I; Bowen, Kit H

2017-12-22

The Al=Al double bond is elusive in chemistry. Herein we report the results obtained via combined photoelectron spectroscopy and ab initio studies of the LiAl 2 H 4 - cluster that confirm the formation of a conventional Al=Al double bond. Comprehensive searches for the most stable structures of the LiAl 2 H 4 - cluster have shown that the global minimum isomer I possesses a geometric structure which resembles that of Si 2 H 4 , demonstrating a successful example of the transmutation of Al atoms into Si atoms by electron donation. Theoretical simulations of the photoelectron spectrum discovered the coexistence of two isomers in the ion beam, including the one with the Al=Al double bond. © 2017 Wiley-VCH Verlag GmbH & Co. KGaA, Weinheim.

Practising alchemy: the transmutation of evidence into best health care.

PubMed

Goodyear-Smith, Felicity

2011-04-01

Alchemy was the synthesis or transmutation of all elements in perfect balance to obtain the philosopher's stone, the key to health. Just as alchemists sought this, so health practitioners always seek the best possible practice for optimal health outcomes for our patients. Best practice requires full knowledge--a little information can be dangerous. We need to serve our apprenticeship before we master our profession. Our profession is about improving health care. While the journey may start at medical school, the learning never ceases. It is not only about practising medicine, it is about the development of the practitioner. Professional practice requires systematic thinking combined with capacity to deal morally and creatively in areas of complexity and uncertainty appropriate to a specific context. It requires exemplary communication skills to interact with patients to facilitate collaborative decision making resulting in best practice. The synthesis of scientific and contextual evidence is a concept which applies to all disciplines where theoretical knowledge needs to be transferred to action to inform best practice. Decisions need to be made which take into account a complex array of factors, such as social and legal issues and resource constraints. Therefore, journey towards best practice involves transmutation of these three elements: scientific knowledge, the context in which it is applied and phronesis, the practical wisdom of the practitioner. All science has its limitations and we can never know all possible contextual information. Hence, like the philosopher's stone, best practice is a goal to which we aspire but never quite attain.

Effect of broad properties fuel on injector performance in a reverse flow combustor

NASA Technical Reports Server (NTRS)

Raddlebaugh, S. M.; Norgren, C. T.

1983-01-01

The effect of fuel type on the performance of various fuel injectors was investigated in a reverse flow combustor. Combustor performance and emissions are documented for simplex pressure atomizing, spill flow, and airblast fuel injectors using a broad properties fuel and compared with performance using Jet A fuel. Test conditions simulated a range of flight conditions including sea level take off, low and high altitude cruise, as well as a parametric evaluation of the effect of increased combustor loading. The baseline simplex injector produced higher emission levels with corresponding lower combustion efficiency with the broad properties fuel. There was little or not loss in performance by the two advanced concept injectors with the broad properties fuel. The airblast injector proved to be especially insensitive to fuel type.

Infrared bolometers with silicon nitride micromesh absorbers

NASA Technical Reports Server (NTRS)

Bock, J. J.; Turner, A. D.; DelCastillo, H. M.; Beeman, J. W.; Lange, A. E.; Mauskopf, P. D.

1996-01-01

Sensitive far infrared and millimeter wave bolometers fabricated from a freestanding membrane of low stress silicon nitride are reported. The absorber, consisting of a metallized silicon nitride micromesh thermally isolated by radial legs of silicon nitride, is placed in an integrating cavity to efficiently couple to single mode or multiple mode infrared radiation. This structure provides low heat capacity, low thermal conduction and minimal cross section to energetic particles. A neutron transmutation doped Ge thermister is bump bonded to the center of the device and read out with evaporated Cr-Au leads. The limiting performance of the micromesh absorber is discussed and the recent results obtained from a 300 mK cold stage are summarized.

(n,xn) cross section measurements for Y-89 foils used as detectors for high energy neutron measurements in the deeply subcritical assembly "QUINTA"

NASA Astrophysics Data System (ADS)

Bielewicz, Marcin; Kilim, Stanisław; Strugalska-Gola, Elżbieta; Szuta, Marcin; Wojciechowski, Andrzej; Tyutyunnikov, Sergey; Prokofiev, Alexander; Passoth, Elke

2017-09-01

Study of the deep subcritical systems (QUINTA) using relativistic beams is performed within the project "Energy and Transmutation of Radioactive Wastes" (E&T - RAW). The experiment assembly was irradiated by deuteron/proton beam (Dubna NUCLOTRON). We calculated the neutron energy spectrum inside the whole assembly by using threshold energy (n,xn) reactions in yttrium (Y-89) foils. There are almost no experimental cross section data for those reactions. New Y-89(n,xn) cross section measurements were carried out at The Svedberg laboratory (TSL) in Uppsala, Sweden in 2015. In this paper we present preliminary results of those experiments.

Effects of mixing system and pilot fuel quality on diesel-biogas dual fuel engine performance.

PubMed

Bedoya, Iván Darío; Arrieta, Andrés Amell; Cadavid, Francisco Javier

2009-12-01

This paper describes results obtained from CI engine performance running on dual fuel mode at fixed engine speed and four loads, varying the mixing system and pilot fuel quality, associated with fuel composition and cetane number. The experiments were carried out on a power generation diesel engine at 1500 m above sea level, with simulated biogas (60% CH(4)-40% CO(2)) as primary fuel, and diesel and palm oil biodiesel as pilot fuels. Dual fuel engine performance using a naturally aspirated mixing system and diesel as pilot fuel was compared with engine performance attained with a supercharged mixing system and biodiesel as pilot fuel. For all loads evaluated, was possible to achieve full diesel substitution using biogas and biodiesel as power sources. Using the supercharged mixing system combined with biodiesel as pilot fuel, thermal efficiency and substitution of pilot fuel were increased, whereas methane and carbon monoxide emissions were reduced.

Nuclear fuels status

NASA Technical Reports Server (NTRS)

Kania, Michael

1991-01-01

A discussion on coated particle fuel performance from a modular High Temperature Gas Reactor (HTGR) is presented along with experimental results. The following topics are covered: (1) the coated particle fuel concept; (2) the functional requirements; (3) performance limiting mechanisms; (4) fuel performance; and (5) methods/techniques for characterizing performance.

Fuel Cell Electric Vehicle Performance Composite Data Products: Spring 2018

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kurtz, Jennifer M; Sprik, Samuel; Ainscough, Christopher D

This publication includes 22 composite data products (CDPs) produced in Spring 2018 for fuel cell electric vehicle performance in the categories of deployment, driving behavior, fuel economy, fueling behavior, and hydrogen performance.

Performance Evaluation of a High Bandwidth Liquid Fuel Modulation Valve for Active Combustion Control

NASA Technical Reports Server (NTRS)

Saus, Joseph R.; DeLaat, John C.; Chang, Clarence T.; Vrnak, Daniel R.

2012-01-01

At the NASA Glenn Research Center, a characterization rig was designed and constructed for the purpose of evaluating high bandwidth liquid fuel modulation devices to determine their suitability for active combustion control research. Incorporated into the rig s design are features that approximate conditions similar to those that would be encountered by a candidate device if it were installed on an actual combustion research rig. The characterized dynamic performance measures obtained through testing in the rig are planned to be accurate indicators of expected performance in an actual combustion testing environment. To evaluate how well the characterization rig predicts fuel modulator dynamic performance, characterization rig data was compared with performance data for a fuel modulator candidate when the candidate was in operation during combustion testing. Specifically, the nominal and off-nominal performance data for a magnetostrictive-actuated proportional fuel modulation valve is described. Valve performance data were collected with the characterization rig configured to emulate two different combustion rig fuel feed systems. Fuel mass flows and pressures, fuel feed line lengths, and fuel injector orifice size was approximated in the characterization rig. Valve performance data were also collected with the valve modulating the fuel into the two combustor rigs. Comparison of the predicted and actual valve performance data show that when the valve is operated near its design condition the characterization rig can appropriately predict the installed performance of the valve. Improvements to the characterization rig and accompanying modeling activities are underway to more accurately predict performance, especially for the devices under development to modulate fuel into the much smaller fuel injectors anticipated in future lean-burning low-emissions aircraft engine combustors.

PROGRESS IN THE STUDY OF ION IRRADIATION IN TUNGSTEN

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jiang, Weilin; Kruska, Karen; Henager, Charles H.

2017-02-27

The experimental study intends to generate data to validate the theoretical predictions on defect accumulation and recovery, as well as to investigate microstructural evolution and transmutant precipitation in mono- and poly-crystalline tungsten using ion implantation.

Alchemy--A History of Early Technology.

ERIC Educational Resources Information Center

Pollard, A. M.

1988-01-01

Reviews the history of alchemy including personalities and methods. Discusses the philosophy associated with various early chemists and alchemists. Attempts to show that it was not unreasonable for ancient alchemists to believe in the possibility of transmutation. (CW)

Achieving dynamic switchable filter based on a transmutable metasurface using SMA

NASA Astrophysics Data System (ADS)

Chen, Xin; Gao, Jinsong; Kang, Bonan

2017-09-01

We propose a switchable filter composed of transmutable array using shape memory alloys (SMA). It could exhibit a temperature induced morphology change spontaneously like the biological excitability, acting as a shutter that allows the incident energy to be selectively transmitted or reflected with in excess of 12dB isolation at the certain frequencies for both polarizations. Equivalent circuit models describe the operational principle qualitatively and the switching effect is underpinned by the full-wave analysis. A further physical mechanism is shown by contrasting the distributions of electric field and surface current on the surface at the same frequency for the two working modes. The experimental results consist with the theoretical simulations, indicating that the metasurface could serve as one innovative solution for manipulating the electromagnetic waves and enlighten the next generation of advanced electromagnetic materials with more freedom in the processes of design and manufacturing.

ISM band to U-NII band frequency transverter and method of frequency transversion

DOEpatents

Stepp, Jeffrey David [Grandview, MO; Hensley, Dale [Grandview, MO

2006-04-04

A frequency transverter (10) and method for enabling bi-frequency dual-directional transfer of digitally encoded data on an RF carrier by translating between a crowded or otherwise undesirable first frequency band, such as the 2.4 GHz ISM band, and a less-crowded or otherwise desirable second frequency band, such as the 5.0 GHz-6.0 GHz U-NII band. In a preferred embodiment, the transverter (10) connects between an existing data radio (11) and its existing antenna (30), and comprises a bandswitch (12); an input RF isolating device (14); a transmuter (16); a converter (18); a dual output local oscillator (20); an output RF isolating device (22); and an antenna (24) tuned to the second frequency band. The bandswitch (12) allows for bypassing the transverter (10), thereby facilitating its use with legacy systems. The transmuter (14) and converter (16) are adapted to convert to and from, respectively, the second frequency band.

ISM band to U-NII band frequency transverter and method of frequency transversion

DOEpatents

Stepp, Jeffrey David [Grandview, MO; Hensley, Dale [Grandview, MO

2006-09-12

A frequency transverter (10) and method for enabling bi-frequency dual-directional transfer of digitally encoded data on an RF carrier by translating between a crowded or otherwise undesirable first frequency band, such as the 2.4 GHz ISM band, and a less-crowded or otherwise desirable second frequency band, such as the 5.0 GHz 6.0 GHz U-NII band. In a preferred embodiment, the transverter (10) connects between an existing data radio (11) and its existing antenna (30), and comprises a bandswitch (12); an input RF isolating device (14); a transmuter (16); a converter (18); a dual output local oscillator (20); an output RF isolating device (22); and an antenna (24) tuned to the second frequency band. The bandswitch (12) allows for bypassing the transverter (10), thereby facilitating its use with legacy systems. The transmuter (14) and converter (16) are adapted to convert to and from, respectively, the second frequency band.

Transmutation of Personal Glucose Meters into Portable and Highly Sensitive Microbial Pathogen Detection Platform.

PubMed

Wang, Zhenzhen; Chen, Zhaowei; Gao, Nan; Ren, Jinsong; Qu, Xiaogang

2015-10-07

Herein, for the first time, we presented a simple and general approach by using personal glucose meters (PGM) for portable and ultrasensitive detection of microbial pathogens. Upon addition of pathogenic bacteria, glucoamylase-quaternized magnetic nanoparticles (GA-QMNPS) conjugates were disrupted by the competitive multivalent interactions between bacteria and QMNPS, resulting in the release of GA. After magnetic separation, the free GA could catalyze the hydrolysis of amylose into glucose for quantitative readout by PGM. In such way, PGM was transmuted into a bacterial detection device and extremely low detection limits down to 20 cells mL(-1) was achieved. More importantly, QMNPS could inhibit the growth of the bacteria and destroy its cellular structure, which enabled bacteria detection and inhibition simultaneously. The simplicity, portability, sensitivity and low cost of presented work make it attractive for clinical applications. © 2015 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

Electron teleportation and statistical transmutation in multiterminal Majorana islands

NASA Astrophysics Data System (ADS)

Michaeli, Karen; Landau, L. Aviad; Sela, Eran; Fu, Liang

2017-11-01

We study a topological superconductor island with spatially separated Majorana modes coupled to multiple normal-metal leads by single-electron tunneling in the Coulomb blockade regime. We show that low-temperature transport in such a Majorana island is carried by an emergent charge-e boson composed of a Majorana mode and an electronic excitation in leads. This transmutation from Fermi to Bose statistics has remarkable consequences. For noninteracting leads, the system flows to a non-Fermi-liquid fixed point, which is stable against tunnel couplings anisotropy or detuning away from the charge-degeneracy point. As a result, the system exhibits a universal conductance at zero temperature, which is a fraction of the conductance quantum, and low-temperature corrections with a universal power-law exponent. In addition, we consider Majorana islands connected to interacting one-dimensional leads, and find different stable fixed points near and far from the charge-degeneracy point.

Computational Analysis of Dynamic SPK(S8)-JP8 Fueled Combustor-Sector Performance

NASA Technical Reports Server (NTRS)

Ryder, R.; Hendricks, Roberts C.; Huber, M. L.; Shouse, D. T.

2010-01-01

Civil and military flight tests using blends of synthetic and biomass fueling with jet fuel up to 50:50 are currently considered as "drop-in" fuels. They are fully compatible with aircraft performance, emissions and fueling systems, yet the design and operations of such fueling systems and combustors must be capable of running fuels from a range of feedstock sources. This paper provides Smart Combustor or Fuel Flexible Combustor designers with computational tools, preliminary performance, emissions and particulates combustor sector data. The baseline fuel is kerosene-JP-8+100 (military) or Jet A (civil). Results for synthetic paraffinic kerosene (SPK) fuel blends show little change with respect to baseline performance, yet do show lower emissions. The evolution of a validated combustor design procedure is fundamental to the development of dynamic fueling of combustor systems for gas turbine engines that comply with multiple feedstock sources satisfying both new and legacy systems.

40 CFR 63.7515 - When must I conduct subsequent performance tests, fuel analyses, or tune-ups?

Code of Federal Regulations, 2014 CFR

2014-07-01

... performance tests, fuel analyses, or tune-ups? 63.7515 Section 63.7515 Protection of Environment ENVIRONMENTAL..., fuel analyses, or tune-ups? (a) You must conduct all applicable performance tests according to § 63... performance tests and the associated fuel analyses within 60 days after the completion of the performance...

40 CFR 63.7515 - When must I conduct subsequent performance tests, fuel analyses, or tune-ups?

Code of Federal Regulations, 2013 CFR

2013-07-01

... performance tests, fuel analyses, or tune-ups? 63.7515 Section 63.7515 Protection of Environment ENVIRONMENTAL..., fuel analyses, or tune-ups? (a) You must conduct all applicable performance tests according to § 63... performance tests and the associated fuel analyses within 60 days after the completion of the performance...

Modeling and Analysis of Actinide Diffusion Behavior in Irradiated Metal Fuel

NASA Astrophysics Data System (ADS)

Edelmann, Paul G.

There have been numerous attempts to model fast reactor fuel behavior in the last 40 years. The US currently does not have a fully reliable tool to simulate the behavior of metal fuels in fast reactors. The experimental database necessary to validate the codes is also very limited. The DOE-sponsored Advanced Fuels Campaign (AFC) has performed various experiments that are ready for analysis. Current metal fuel performance codes are either not available to the AFC or have limitations and deficiencies in predicting AFC fuel performance. A modified version of a new fuel performance code, FEAST-Metal , was employed in this investigation with useful results. This work explores the modeling and analysis of AFC metallic fuels using FEAST-Metal, particularly in the area of constituent actinide diffusion behavior. The FEAST-Metal code calculations for this work were conducted at Los Alamos National Laboratory (LANL) in support of on-going activities related to sensitivity analysis of fuel performance codes. A sensitivity analysis of FEAST-Metal was completed to identify important macroscopic parameters of interest to modeling and simulation of metallic fuel performance. A modification was made to the FEAST-Metal constituent redistribution model to enable accommodation of newer AFC metal fuel compositions with verified results. Applicability of this modified model for sodium fast reactor metal fuel design is demonstrated.

Pushing the limits of photoreception in twilight conditions: The rod-like cone retina of the deep-sea pearlsides.

PubMed

de Busserolles, Fanny; Cortesi, Fabio; Helvik, Jon Vidar; Davies, Wayne I L; Templin, Rachel M; Sullivan, Robert K P; Michell, Craig T; Mountford, Jessica K; Collin, Shaun P; Irigoien, Xabier; Kaartvedt, Stein; Marshall, Justin

2017-11-01

Most vertebrates have a duplex retina comprising two photoreceptor types, rods for dim-light (scotopic) vision and cones for bright-light (photopic) and color vision. However, deep-sea fishes are only active in dim-light conditions; hence, most species have lost their cones in favor of a simplex retina composed exclusively of rods. Although the pearlsides, Maurolicus spp., have such a pure rod retina, their behavior is at odds with this simplex visual system. Contrary to other deep-sea fishes, pearlsides are mostly active during dusk and dawn close to the surface, where light levels are intermediate (twilight or mesopic) and require the use of both rod and cone photoreceptors. This study elucidates this paradox by demonstrating that the pearlside retina does not have rod photoreceptors only; instead, it is composed almost exclusively of transmuted cone photoreceptors. These transmuted cells combine the morphological characteristics of a rod photoreceptor with a cone opsin and a cone phototransduction cascade to form a unique photoreceptor type, a rod-like cone, specifically tuned to the light conditions of the pearlsides' habitat (blue-shifted light at mesopic intensities). Combining properties of both rods and cones into a single cell type, instead of using two photoreceptor types that do not function at their full potential under mesopic conditions, is likely to be the most efficient and economical solution to optimize visual performance. These results challenge the standing paradigm of the function and evolution of the vertebrate duplex retina and emphasize the need for a more comprehensive evaluation of visual systems in general.

Pushing the limits of photoreception in twilight conditions: The rod-like cone retina of the deep-sea pearlsides

PubMed Central

de Busserolles, Fanny; Cortesi, Fabio; Helvik, Jon Vidar; Davies, Wayne I. L.; Templin, Rachel M.; Sullivan, Robert K. P.; Michell, Craig T.; Mountford, Jessica K.; Collin, Shaun P.; Irigoien, Xabier; Kaartvedt, Stein; Marshall, Justin

2017-01-01

Most vertebrates have a duplex retina comprising two photoreceptor types, rods for dim-light (scotopic) vision and cones for bright-light (photopic) and color vision. However, deep-sea fishes are only active in dim-light conditions; hence, most species have lost their cones in favor of a simplex retina composed exclusively of rods. Although the pearlsides, Maurolicus spp., have such a pure rod retina, their behavior is at odds with this simplex visual system. Contrary to other deep-sea fishes, pearlsides are mostly active during dusk and dawn close to the surface, where light levels are intermediate (twilight or mesopic) and require the use of both rod and cone photoreceptors. This study elucidates this paradox by demonstrating that the pearlside retina does not have rod photoreceptors only; instead, it is composed almost exclusively of transmuted cone photoreceptors. These transmuted cells combine the morphological characteristics of a rod photoreceptor with a cone opsin and a cone phototransduction cascade to form a unique photoreceptor type, a rod-like cone, specifically tuned to the light conditions of the pearlsides’ habitat (blue-shifted light at mesopic intensities). Combining properties of both rods and cones into a single cell type, instead of using two photoreceptor types that do not function at their full potential under mesopic conditions, is likely to be the most efficient and economical solution to optimize visual performance. These results challenge the standing paradigm of the function and evolution of the vertebrate duplex retina and emphasize the need for a more comprehensive evaluation of visual systems in general. PMID:29134201

Ionizing radiation and tritium transmutation both cause formation of 5-hydroxymethyl-2'-deoxyuridine in cellular DNA.

PubMed Central

Teebor, G W; Frenkel, K; Goldstein, M S

1984-01-01

HeLa cells grown in the presence of [methyl-3H]thymidine contained large amounts of 5-hydroxymethyl-2'-deoxyuridine (HMdU) in their DNA. When the cells were grown in [6-3H]thymidine and their DNA was labeled to the same specific activity, no HMdU was present. When such [6-3H]thymidine-labeled cells were exposed to increasing amounts of gamma-radiation, small but increasing amounts of HMdU were formed in their DNA. This indicates that HMdU can be formed in DNA by two distinct mechanisms. The first is the result of the transmutation of 3H to 3He (beta decay) in the methyl group of thymidine, leading to formation of a carbocation. This short-lived ion reacts with hydroxide ions of water, yielding the hydroxymethyl group. HMdU that is formed by this mechanism is formed at the rate of beta decay of 3H. It appears only in [methyl-3H]thymidine residues and is present in the DNA of both nonirradiated and gamma-irradiated cells. The second mechanism is the result of the radiolysis of water caused by ionizing radiation. The resultant radical species, particularly hydroxyl radicals, may react with many sites on DNA. When the methyl group of thymine is attacked by hydroxyl radicals, the hydroxymethyl group is formed. The formation of HMdU by this mechanism was detected only when [6-3H]thymidine-labeled cells were used, since transmutation of 3H in position 6 of thymine cannot yield HMdU. PMID:6582490

Ionizing radiation and tritium transmutation both cause formation of 5-hydroxymethyl-2'-deoxyuridine in cellular DNA.

PubMed

Teebor, G W; Frenkel, K; Goldstein, M S

1984-01-01

HeLa cells grown in the presence of [methyl-3H]thymidine contained large amounts of 5-hydroxymethyl-2'-deoxyuridine (HMdU) in their DNA. When the cells were grown in [6-3H]thymidine and their DNA was labeled to the same specific activity, no HMdU was present. When such [6-3H]thymidine-labeled cells were exposed to increasing amounts of gamma-radiation, small but increasing amounts of HMdU were formed in their DNA. This indicates that HMdU can be formed in DNA by two distinct mechanisms. The first is the result of the transmutation of 3H to 3He (beta decay) in the methyl group of thymidine, leading to formation of a carbocation. This short-lived ion reacts with hydroxide ions of water, yielding the hydroxymethyl group. HMdU that is formed by this mechanism is formed at the rate of beta decay of 3H. It appears only in [methyl-3H]thymidine residues and is present in the DNA of both nonirradiated and gamma-irradiated cells. The second mechanism is the result of the radiolysis of water caused by ionizing radiation. The resultant radical species, particularly hydroxyl radicals, may react with many sites on DNA. When the methyl group of thymine is attacked by hydroxyl radicals, the hydroxymethyl group is formed. The formation of HMdU by this mechanism was detected only when [6-3H]thymidine-labeled cells were used, since transmutation of 3H in position 6 of thymine cannot yield HMdU.

Gold nanoparticles production using reactor and cyclotron based methods in assessment of (196,198)Au production yields by (197)Au neutron absorption for therapeutic purposes.

PubMed

Khorshidi, Abdollah

2016-11-01

Medical nano-gold radioisotopes is produced regularly using high-flux nuclear reactors, and an accelerator-driven neutron activator can turn out higher yield of (197)Au(n,γ)(196,198)Au reactions. Here, nano-gold production via radiative/neutron capture was investigated using irradiated Tehran Research Reactor flux and also simulated proton beam of Karaj cyclotron in Iran. (197)Au nano-solution, including 20nm shaped spherical gold and water, was irradiated under Tehran reactor flux at 2.5E+13n/cm(2)/s for (196,198)Au activity and production yield estimations. Meanwhile, the yield was examined using 30MeV proton beam of Karaj cyclotron via simulated new neutron activator containing beryllium target, bismuth moderator around the target, and also PbF2 reflector enclosed the moderator region. Transmutation in (197)Au nano-solution samples were explored at 15 and 25cm distances from the target. The neutron flux behavior inside the water and bismuth moderators was investigated for nano-gold particles transmutation. The transport of fast neutrons inside bismuth material as heavy nuclei with a lesser lethargy can be contributed in enhanced nano-gold transmutation with long duration time than the water moderator in reactor-based method. Cyclotron-driven production of βeta-emitting radioisotopes for brachytherapy applications can complete the nano-gold production technology as a safer approach as compared to the reactor-based method. Copyright © 2016 Elsevier B.V. All rights reserved.

Modern Alchemy.

ERIC Educational Resources Information Center

Seaborg, Glenn T.

1983-01-01

Reviews the historical development of the periodic table, examining major changes due to understanding of radioactivity, synthetic transmutation by bombardment, differences between transuranium elements and the lanthanide series, and the transactinide elements. Discusses the continuing work on atomic synthesis and its importance in extending our…

Canadian experience in irradiation and testing of MOX fuel

NASA Astrophysics Data System (ADS)

Yatabe, S.; Floyd, M.; Dimayuga, F.

2018-04-01

Experimental irradiation and performance testing of Mixed OXide (MOX) fuel at the Canadian Nuclear Laboratories (CNL) has taken place for more than 40 years. These experiments investigated MOX fuel behaviour and compared it with UO2 behaviour to develop and verify fuel performance models. This article compares the performance of MOX of various concentrations and homogeneities, under different irradiation conditions. These results can be applied to future fuel designs. MOX fuel irradiated by CNL was found to be comparable in performance to similarly designed and operated UO2 fuel. MOX differs in behaviour from UO2 fuel in several ways. Fission-gas release, grain growth and the thickness of zirconium oxide on the inner sheath appear to be related to MOX fuel homogeneity. Columnar grains formed at the pellet centre begin to develop at lower powers in MOX than in UO2 fuel.

Alternate aircraft fuels prospects and operational implications

NASA Technical Reports Server (NTRS)

Witcofski, R. D.

1977-01-01

The paper discusses NASA studies of the potentials of coal-derived aviation fuels, specifically synthetic aviation kerosene, liquid methane, and liquid hydrogen. Topics include areas of fuel production, air terminal requirements for aircraft fueling (for liquid hydrogen only), and the performance characteristics of aircraft designed to utilize alternate fuels. Energy requirements associated with the production of each of the three selected fuels are determined, and fuel prices are estimated. Subsonic commercial air transports using liquid hydrogen fuel have been analyzed, and their performance and the performance of aircraft which use commercial aviation kerosene are compared. Environmental and safety issues are considered.

Theory and methods for measuring the effective multiplication constant in ADS

NASA Astrophysics Data System (ADS)

Rugama Saez, Yolanda

2001-10-01

In the thesis an absolute measurements technique for the subcriticality determination is presented. The ADS is a hybrid system where a subcritical system is fed by a proton accelerator. There are different proposals to define an ADS, one is to use plutonium and minor actinides from power plants waste as fuel to be transmuted into non radioactive isotopes (transmuter/burner, ATW). Another proposal is to use a Th232-U233 cycle (Energy Amplifier), being that thorium is an interesting and abundant fertile isotope. The development of accelerator driven systems (ADS) requires the development of methods to monitor and control the subcriticality of this kind of system without interfering with its normal operation mode. With this finality, we have applied noise analysis techniques that allow us to characterise the system when it is operating. The method presented in this thesis is based on the stochastic neutron and photon transport theory that can be implemented by presently available neutron/photon transport codes. In this work, first we analyse the stochastic transport theory which has been applied to define a parameter to determine the subcritical reactivity monitoring measurements. Finally we give the main limitations and recommendations for these subcritical monitoring methodology. As a result of the theoretical methodology, done in the first part of this thesis, a monitoring measurement technique has been developed and verified using two coupled Monte Carlo programs. The first one, LAHET, simulates the spallation collisions and the high energy transport and the other, MCNP-DSP, is used to estimate the counting statistics from a neutron/photon ray counter in a fissile system, as well as the transport for neutron with energies less than 20 MeV. From the coupling of both codes we developed the LAHET/MCNP-DSP code which, has the capability to simulate the total process in the ADS from the proton interaction to the signal detector processing. In these simulations, we compute the cross power spectral densities between pairs of detectors located inside the system which, is defined as the measured parameter. From the comparison of the theoretical predictions with the Monte Carlo simulations, we obtain some practical and simple methods to determine the system multiplication constant. (Abstract shortened by UMI.)

40 CFR 600.207-08 - Calculation and use of vehicle-specific 5-cycle-based fuel economy values for vehicle...

Code of Federal Regulations, 2011 CFR

2011-07-01

.... (i) Calculate the 5-cycle city and highway fuel economy values from the tests performed using gasoline or diesel test fuel. (ii)(A) Calculate the 5-cycle city and highway fuel economy values from the tests performed using alcohol or natural gas test fuel, if 5-cycle testing has been performed. Otherwise...

The Potential of the LFR and the ELSY Project

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cinotti, L; Smith, C F; Sienicki, J J

2007-03-12

This paper presents the current status of the development of the Lead-cooled Fast Reactor (LFR) in support of Generation IV (GEN IV) Nuclear Energy Systems. The approach being taken by the GIF plan is to address the research priorities of each member state in developing an integrated and coordinated research program to achieve common objectives, while avoiding duplication of effort. The integrated plan being prepared by the LFR Provisional System Steering Committee of the GIF, known as the LFR System research Plan (SRP) recognizes two principal technology tracks for pursuit of LFR technology: (1) a small, transportable system of 10-100more » MWe size that features a very long refueling interval, (2) a larger-sized system rated at about 600 MWe, intended for central station power generation and waste transmutation. This paper, in particular, describes the ongoing activities to develop the Small Secure Transportable Autonomous Reactor (SSTAR) and the European Lead-cooled SYstem (ELSY), the two research initiatives closely aligned with the overall tracks of the SRP and outlines the Proliferation-resistant Environment-friendly Accident-tolerant Continual & Economical Reactors (PEACER) conceived with particular focus on burning/transmuting of long-living TRU waste and fission fragments of concern, such as Tc and I. The current reference design for the SSTAR is a 20 MWe natural circulation pool-type reactor concept with a small shippable reactor vessel. Specific features of the lead coolant, the nitride fuel containing transuranics, the fast spectrum core, and the small size combine to promote a unique approach to achieve proliferation resistance, while also enabling fissile self-sufficiency, autonomous load following, simplicity of operation, reliability, transportability, as well as a high degree of passive safety. Conversion of the core thermal power into electricity at a high plant efficiency of 44% is accomplished utilizing a supercritical carbon dioxide Brayton cycle power converter. The ELSY reference design is a 600 MWe pool-type reactor cooled by pure lead. This concept has been under development since September 2006, and is sponsored by the Sixth Framework Programme of EURATOM. The ELSY project is being performed by a consortium consisting of twenty organizations including seventeen from Europe, two from Korea and one from the USA. ELSY aims to demonstrate the possibility of designing a competitive and safe fast critical reactor using simple engineered technical features while fully complying with the Generation IV goal of minor actinide (MA) burning capability. The use of a compact and simple primary circuit with the additional objective that all internal components be removable, are among the reactor features intended to assure competitive electric energy generation and long-term investment protection. Simplicity is expected to reduce both the capital cost and the construction time; these are also supported by the compactness of the reactor building (reduced footprint and height). The reduced footprint would be possible due to the elimination of the Intermediate Cooling System, the reduced elevation the result of the design approach of reduced-height components.« less

Performance of Transuranic-Loaded Fully Ceramic Micro-Encapsulated Fuel in LWRs Final Report, Including Void Reactivity Evaluation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Michael A. Pope; R. Sonat Sen; Brian Boer

2011-09-01

The current focus of the Deep Burn Project is on once-through burning of transuranics (TRU) in light-water reactors (LWRs). The fuel form is called Fully-Ceramic Micro-encapsulated (FCM) fuel, a concept that borrows the tri-isotropic (TRISO) fuel particle design from high-temperature reactor technology. In the Deep Burn LWR (DB-LWR) concept, these fuel particles are pressed into compacts using SiC matrix material and loaded into fuel pins for use in conventional LWRs. The TRU loading comes from the spent fuel of a conventional LWR after 5 years of cooling. Unit cell and assembly calculations have been performed using the DRAGON-4 code tomore » assess the physics attributes of TRU-only FCM fuel in an LWR lattice. Depletion calculations assuming an infinite lattice condition were performed with calculations of various reactivity coefficients performed at each step. Unit cells and assemblies containing typical UO2 and mixed oxide (MOX) fuel were analyzed in the same way to provide a baseline against which to compare the TRU-only FCM fuel. Then, assembly calculations were performed evaluating the performance of heterogeneous arrangements of TRU-only FCM fuel pins along with UO2 pins.« less

40 CFR 63.7515 - When must I conduct subsequent performance tests or fuel analyses?

Code of Federal Regulations, 2010 CFR

2010-07-01

... performance tests or fuel analyses? 63.7515 Section 63.7515 Protection of Environment ENVIRONMENTAL PROTECTION... Requirements § 63.7515 When must I conduct subsequent performance tests or fuel analyses? (a) You must conduct... previous performance test. (f) You must conduct a fuel analysis according to § 63.7521 for each type of...

40 CFR 63.7515 - When must I conduct subsequent performance tests, fuel analyses, or tune-ups?

Code of Federal Regulations, 2011 CFR

2011-07-01

... performance tests, fuel analyses, or tune-ups? 63.7515 Section 63.7515 Protection of Environment ENVIRONMENTAL... Compliance Requirements § 63.7515 When must I conduct subsequent performance tests, fuel analyses, or tune... and the associated initial fuel analyses within 90 days after the completion of the performance tests...

40 CFR 63.7515 - When must I conduct subsequent performance tests or fuel analyses?

Code of Federal Regulations, 2011 CFR

2011-07-01

... performance tests or fuel analyses? 63.7515 Section 63.7515 Protection of Environment ENVIRONMENTAL PROTECTION... Requirements § 63.7515 When must I conduct subsequent performance tests or fuel analyses? (a) You must conduct... previous performance test. (f) You must conduct a fuel analysis according to § 63.7521 for each type of...

40 CFR 63.7515 - When must I conduct subsequent performance tests, fuel analyses, or tune-ups?

Code of Federal Regulations, 2012 CFR

2012-07-01

... performance tests, fuel analyses, or tune-ups? 63.7515 Section 63.7515 Protection of Environment ENVIRONMENTAL... Compliance Requirements § 63.7515 When must I conduct subsequent performance tests, fuel analyses, or tune... and the associated initial fuel analyses within 90 days after the completion of the performance tests...

40 CFR 63.7515 - When must I conduct subsequent performance tests or fuel analyses?

Code of Federal Regulations, 2012 CFR

2012-07-01

... performance tests or fuel analyses? 63.7515 Section 63.7515 Protection of Environment ENVIRONMENTAL PROTECTION... Requirements § 63.7515 When must I conduct subsequent performance tests or fuel analyses? (a) You must conduct... previous performance test. (f) You must conduct a fuel analysis according to § 63.7521 for each type of...

High density fuel qualification for a gas turbine engine

DOE Office of Scientific and Technical Information (OSTI.GOV)

Macleod, J.D.; Orbanski, B.; Hastings, P.R.

1992-01-01

A program for the evaluation of gas turbine engine performance, carried out in the Engine Laboratory of the National Research Council of Canada, is described. Problems under consideration include performance alteration between JP-4 fuel and a high energy density fuel, called strategic military fuel (SMF); performance deterioration during the accelerated endurance test; and emission analysis. The T56 fuel control system is found to be capable of operation on the higher energy density fuel with no detrimental effects regarding control of the engine's normal operating regime. The deterioration of the engine performance during 150-hour endurance tests on SMF was very high,more » which was caused by an increase in turbine nozzle effective flow area and turbine blade untwist. The most significant performance losses during the endurance tests were on corrected output power, fuel flow, specific fuel consumption and compressor and turbine presure ratio. 9 refs.« less

FY2016 Ceramic Fuels Development Annual Highlights

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mcclellan, Kenneth James

Key challenges for the Advanced Fuels Campaign are the development of fuel technologies to enable major increases in fuel performance (safety, reliability, power and burnup) beyond current technologies, and development of characterization methods and predictive fuel performance models to enable more efficient development and licensing of advanced fuels. Ceramic fuel development activities for fiscal year 2016 fell within the areas of 1) National and International Technical Integration, 2) Advanced Accident Tolerant Ceramic Fuel Development, 3) Advanced Techniques and Reference Materials Development, and 4) Fabrication of Enriched Ceramic Fuels. High uranium density fuels were the focus of the ceramic fuels efforts.more » Accomplishments for FY16 primarily reflect the prioritization of identification and assessment of new ceramic fuels for light water reactors which have enhanced accident tolerance while also maintaining or improving normal operation performance, and exploration of advanced post irradiation examination techniques which will support more efficient testing and qualification of new fuel systems.« less

Simulating the Use of Alternative Fuels in a Turbofan Engine

NASA Technical Reports Server (NTRS)

Litt, Jonathan S.; Chin, Jeffrey Chevoor; Liu, Yuan

2013-01-01

The interest in alternative fuels for aviation has created a need to evaluate their effect on engine performance. The use of dynamic turbofan engine simulations enables the comparative modeling of the performance of these fuels on a realistic test bed in terms of dynamic response and control compared to traditional fuels. The analysis of overall engine performance and response characteristics can lead to a determination of the practicality of using specific alternative fuels in commercial aircraft. This paper describes a procedure to model the use of alternative fuels in a large commercial turbofan engine, and quantifies their effects on engine and vehicle performance. In addition, the modeling effort notionally demonstrates that engine performance may be maintained by modifying engine control system software parameters to account for the alternative fuel.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pawlowski, Alexander; Splitter, Derek A

It is well known that spark ignited engine performance and efficiency is closely coupled to fuel octane number. The present work combines historical and recent trends in spark ignition engines to build a database of engine design, performance, and fuel octane requirements over the past 80 years. The database consists of engine compression ratio, required fuel octane number, peak mean effective pressure, specific output, and combined unadjusted fuel economy for passenger vehicles and light trucks. Recent trends in engine performance, efficiency, and fuel octane number requirement were used to develop correlations of fuel octane number utilization, performance, specific output. Themore » results show that historically, engine compression ratio and specific output have been strongly coupled to fuel octane number. However, over the last 15 years the sales weighted averages of compression ratios, specific output, and fuel economy have increased, while the fuel octane number requirement has remained largely unchanged. Using the developed correlations, 10-year-out projections of engine performance, design, and fuel economy are estimated for various fuel octane numbers, both with and without turbocharging. The 10-year-out projection shows that only by keeping power neutral while using 105 RON fuel will allow the vehicle fleet to meet CAFE targets if only the engine is relied upon to decrease fuel consumption. If 98 RON fuel is used, a power neutral fleet will have to reduce vehicle weight by 5%.« less

40 CFR 600.206-93 - Calculation and use of fuel economy values for gasoline-fueled, diesel-fueled, electric, alcohol...

Code of Federal Regulations, 2011 CFR

2011-07-01

... tests performed using gasoline or diesel test fuel. (ii) Calculate the city, highway, and combined fuel economy values from the tests performed using alcohol or natural gas test fuel. (b) If only one equivalent... 40 Protection of Environment 30 2011-07-01 2011-07-01 false Calculation and use of fuel economy...

Fuel saver based on electromagnetic induction for automotive engine

NASA Astrophysics Data System (ADS)

Siregar, Houtman P.; Sibarani, Maradu

2007-12-01

In the considered research is designed and analyzed the performance of the fuel saver which is based on electromagnetic induction for automotive diesel engine. The fuel saver which is based on permanent magnet has sold in market and its performance has tested. In comparison to the former fuel saver, in the proposed work is produced fuel saver which is based on electromagnetic induction. The considered research is the continuation of my former work. Performance of the produced fuel saver which is installed in the fuel line of internal combustion engine rig is compared to the performance of the standard internal combustion engine rig Speed of the engine, wire diameter of coil, and number of coil which is coiled in the winding of the the fuel saver are chosen as the testing variables. The considered research has succeeded to design the fuel saver which is based on electromagnetic induction for saving the automotive fuel consumption. Results of the research show that the addition of the fuel saver which is based on electromagnetic induction to the flow of the diesel fuel can significantly save the automative fuel consumption. In addition the designed fuel saver can reduce the opacity of the emission gas.

Utilization of TRISO Fuel with LWR Spent Fuel in Fusion-Fission Hybrid Reactor System

NASA Astrophysics Data System (ADS)

Acır, Adem; Altunok, Taner

2010-10-01

HTRs use a high performance particulate TRISO fuel with ceramic multi-layer coatings due to the high burn up capability and very neutronic performance. TRISO fuel because of capable of high burn up and very neutronic performance is conducted in a D-T fusion driven hybrid reactor. In this study, TRISO fuels particles are imbedded body-centered cubic (BCC) in a graphite matrix with a volume fraction of 68%. The neutronic effect of TRISO coated LWR spent fuel in the fuel rod used hybrid reactor on the fuel performance has been investigated for Flibe, Flinabe and Li20Sn80 coolants. The reactor operation time with the different first neutron wall loads is 24 months. Neutron transport calculations are evaluated by using XSDRNPM/SCALE 5 codes with 238 group cross section library. The effect of TRISO coated LWR spent fuel in the fuel rod used hybrid reactor on tritium breeding (TBR), energy multiplication (M), fissile fuel breeding, average burn up values are comparatively investigated. It is shown that the high burn up can be achieved with TRISO fuel in the hybrid reactor.

Insights.

ERIC Educational Resources Information Center

Bogner, Donna, Ed.

1988-01-01

Describes two methods to teach radioactive decay to secondary students with wide ranging abilities. Activities are designed to follow classroom discussions of atomic structure, transmutation, half life, and nuclear decay. Includes "The Tasmanian Empire: A Radioactive Dating Activity" and an exercise to teach concepts of half life without…

Family Environmental and Genetic Influences on Children's Future Chemical Dependency.

ERIC Educational Resources Information Center

Kumpfer, Karol L.; DeMarsh, Joseph

1985-01-01

Discusses the following in relation to their predictability to future drug abuse in youth: (1) susceptibility of children of chemically dependent parents; (2) genetic transmutation; (3) family structure and management; (4) socialization; and (5) cognitive family characteristics. (Author/LHW)

IEA-R1 Nuclear Research Reactor: 58 Years of Operating Experience and Utilization for Research, Teaching and Radioisotopes Production

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cardenas, Jose Patricio Nahuel; Filho, Tufic Madi; Saxena, Rajendra

IEA-R1 research reactor at the Instituto de Pesquisas Energeticas e Nucleares (Nuclear and Energy Research Institute) IPEN, Sao Paulo, Brazil is the largest power research reactor in Brazil, with a maximum power rating of 5 MWth. It is being used for basic and applied research in the nuclear and neutron related sciences, for the production of radioisotopes for medical and industrial applications, and for providing services of neutron activation analysis, real time neutron radiography, and neutron transmutation doping of silicon. IEA-R1 is a swimming pool reactor, with light water as the coolant and moderator, and graphite and beryllium as reflectors.more » The reactor was commissioned on September 16, 1957 and achieved its first criticality. It is currently operating at 4.5 MWth with a 60-hour cycle per week. In the early sixties, IPEN produced {sup 131}I, {sup 32}P, {sup 198}Au, {sup 24}Na, {sup 35}S, {sup 51}Cr and labeled compounds for medical use. During the past several years, a concerted effort has been made in order to upgrade the reactor power to 5 MWth through refurbishment and modernization programs. One of the reasons for this decision was to produce {sup 99}Mo at IPEN. The reactor cycle will be gradually increased to 120 hours per week continuous operation. It is anticipated that these programs will assure the safe and sustainable operation of the IEA-R1 reactor for several more years, to produce important primary radioisotopes {sup 99}Mo, {sup 125}I, {sup 131}I, {sup 153}Sm and {sup 192}Ir. Currently, all aspects of dealing with fuel element fabrication, fuel transportation, isotope processing, and spent fuel storage are handled by IPEN at the site. The reactor modernization program is slated for completion by 2015. This paper describes 58 years of operating experience and utilization of the IEA-R1 research reactor for research, teaching and radioisotopes production. (authors)« less

Effect of Axially Staged Fuel Introduction on Performance of One-quarter Sector of Annular Turbojet Combustor

NASA Technical Reports Server (NTRS)

Zettle, Eugene V; Mark, Herman

1953-01-01

The design principle of injecting liquid fuel at more than one axial station in an annual turbojet combustor was investigated. Fuel was injected into the combustor as much as 5 inches downstream of the primary fuel injectors. Many fuel-injection configurations were examined and the performance results are presented for 11 configurations that best demonstrate the trends in performance obtained. The performance investigations were made at a constant combustor-inlet pressure of 15 inches of mercury absolute and at air flows up to 70 percent higher than values typical of current design practice. At these higher air flows, staging the fuel introduction improved the combustion efficiency considerably over that obtained in the combustor when no fuel staging was employed. At air flows currently encountered in turbojet engines, fuel staging was of minor value. Radial temperature distribution seemed relatively unaffected by the location of fuel-injection stations.

TRISO Fuel Performance: Modeling, Integration into Mainstream Design Studies, and Application to a Thorium-fueled Fusion-Fission Hybrid Blanket

NASA Astrophysics Data System (ADS)

Powers, Jeffrey J.

2011-12-01

This study focused on creating a new tristructural isotropic (TRISO) coated particle fuel performance model and demonstrating the integration of this model into an existing system of neutronics and heat transfer codes, creating a user-friendly option for including fuel performance analysis within system design optimization and system-level trade-off studies. The end product enables both a deeper understanding and better overall system performance of nuclear energy systems limited or greatly impacted by TRISO fuel performance. A thorium-fueled hybrid fusion-fission Laser Inertial Fusion Energy (LIFE) blanket design was used for illustrating the application of this new capability and demonstrated both the importance of integrating fuel performance calculations into mainstream design studies and the impact that this new integrated analysis had on system-level design decisions. A new TRISO fuel performance model named TRIUNE was developed and verified and validated during this work with a novel methodology established for simulating the actual lifetime of a TRISO particle during repeated passes through a pebble bed. In addition, integrated self-consistent calculations were performed for neutronics depletion analysis, heat transfer calculations, and then fuel performance modeling for a full parametric study that encompassed over 80 different design options that went through all three phases of analysis. Lastly, side studies were performed that included a comparison of thorium and depleted uranium (DU) LIFE blankets as well as some uncertainty quantification work to help guide future experimental work by assessing what material properties in TRISO fuel performance modeling are most in need of improvement. A recommended thorium-fueled hybrid LIFE engine design was identified with an initial fuel load of 20MT of thorium, 15% TRISO packing within the graphite fuel pebbles, and a 20cm neutron multiplier layer with beryllium pebbles in flibe molten salt coolant. It operated at a system power level of 2000 MWth, took about 3.5 years to reach full plateau power, and was capable of an End of Plateau burnup of 38.7 %FIMA if considering just the neutronic constraints in the system design; however, fuel performance constraints led to a maximum credible burnup of 12.1 %FIMA due to a combination of internal gas pressure and irradiation effects on the TRISO materials (especially PyC) leading to SiC pressure vessel failures. The optimal neutron spectrum for the thorium-fueled blanket options evaluated seemed to favor a hard spectrum (low but non-zero neutron multiplier thicknesses and high TRISO packing fractions) in terms of neutronic performance but the fuel performance constraints demonstrated that a significantly softer spectrum would be needed to decrease the rate of accumulation of fast neutron fluence in order to improve the maximum credible burnup the system could achieve.

High-temperature fatigue life of type 316 stainless steel containing irradiation induced helium

NASA Astrophysics Data System (ADS)

Grossbeck, M. L.; Liu, K. C.

Specimens of 20%-cold-worked AISI type 316 stainless steel were irradiated in the High Flux Isotope Reactor (HFIR) at 550°C to a maximum damage level of 15 dpa and a transmutation produced helium level of 820 at. ppm. Fully reversed strain controlled fatigue tests were performed in a vacuum at 550°C. No significant effect of the irradiation on low-cycle fatigue life was observed; however, the strain range of the 10 7 cycle endurance limit decreased from 0.35 to 0.30%. The relation between total strain range and number of cycles to failure was found to be ΔEt = 0.02 Nf-0.12+ Nf-0.6 for N f < 10 7 cycles.

US-RERTR Advanced Fuel Development Plans : 1999.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Meyer, M. K.

1998-10-22

Twelve fuel alloys were included in the very-high-density RERTR-1 and RERTR-2 microplate irradiation experiments. Experience gained during fabrication and results from the post-irradiation examination of these fuels has allowed us to narrow the focus of our fuel development efforts in preparation for the next set of irradiation experiments. Specific technical problems in both the areas of fuel fabrication and irradiation performance remain to be addressed. Examples of these are powder fabrication, fuel phase gamma stability versus density, and fuel-matrix interaction. In order to more efficiently address metal alloy fuel performance issues, work will continue on establishing a theoretical basis formore » alloy stability and metal alloy dispersion fuel irradiation performance. Plans to address these fuel development issues in the coming year will be presented.« less

Condensed Matter Nuclear Science

NASA Astrophysics Data System (ADS)

Biberian, Jean-Paul

2006-02-01

1. General. A tribute to gene Mallove - the "Genie" reactor / K. Wallace and R. Stringham. An update of LENR for ICCF-11 (short course, 10/31/04) / E. Storms. New physical effects in metal deuterides / P. L. Hagelstein ... [et al.]. Reproducibility, controllability, and optimization of LENR experiments / D. J. Nagel -- 2. Experiments. Electrochemistry. Evidence of electromagnetic radiation from Ni-H systems / S. Focardi ... [et al.]. Superwave reality / I. Dardik. Excess heat in electrolysis experiments at energetics technologies / I. Dardik ... [et al.]. "Excess heat" during electrolysis in platinum/K[symbol]CO[symbol]/nickel light water system / J. Tian ... [et al.]. Innovative procedure for the, in situ, measurement of the resistive thermal coefficient of H(D)/Pd during electrolysis; cross-comparison of new elements detected in the Th-Hg-Pd-D(H) electrolytic cells / F. Celani ... [et al.]. Emergence of a high-temperature superconductivity in hydrogen cycled Pd compounds as an evidence for superstoihiometric H/D sites / A. Lipson ... [et al.]. Plasma electrolysis. Calorimetry of energy-efficient glow discharge - apparatus design and calibration / T. B. Benson and T. O. Passell. Generation of heat and products during plasma electrolysis / T. Mizuno ... [et al.]. Glow discharge. Excess heat production in Pd/D during periodic pulse discharge current in various conditions / A. B. Karabut. Beam experiments. Accelerator experiments and theoretical models for the electron screening effect in metallic environments / A. Huke, K. Czerski, and P. Heide. Evidence for a target-material dependence of the neutron-proton branching ratio in d+d reactions for deuteron energies below 20keV / A. Huke ... [et al.]. Experiments on condensed matter nuclear events in Kobe University / T. Minari ... [et al.]. Electron screening constraints for the cold fusion / K. Czerski, P. Heide, and A. Huke. Cavitation. Low mass 1.6 MHz sonofusion reactor / R. Stringham. Particle detection. Research into characteristics of X-ray emission laser beams from solidstate cathode medium of high-current glow discharge / A. B. Karabut. Charged particles from Ti and Pd foils / L. Kowalski ... [et al.]. Cr-39 track detectors in cold fusion experiments: review and perspectives / A. S. Roussetski. Energetic particle shower in the vapor from electrolysis / R. A. Oriani and J. C. Fisher. Nuclear reactions produced in an operating electrolysis cell / R. A. Oriani and J. C. Fisher. Evidence of microscopic ball lightning in cold fusion experiments / E. H. Lewis. Neutron emission from D[symbol] gas in magnetic fields under low temperature / T. Mizuno ... [et al.]. Energetic charged particle emission from hydrogen-loaded Pd and Ti cathodes and its enhancement by He-4 implantation / A. G. Lipson ... [et al.]. H-D permeation. Observation of nuclear transmutation reactions induced by D[symbol] gas permeation through Pd complexes / Y. Iwamura ... [et al.]. Deuterium (hydrogen) flux permeating through palladium and condensed matter nuclear science / Q. M. Wei ... [et al.]. Triggering. Precursors and the fusion reactions in polarized Pd/D-D[symbol]O system: effect of an external electric field / S. Szpak, P. A. Mosier-Boss, and F. E. Gordon. Calorimetric and neutron diagnostics of liquids during laser irradiation / Yu. N. Bazhutov ... [et al.]. Anomalous neutron capture and plastic deformation of Cu and Pd cathodes during electrolysis in a weak thermalized neutron field: evidence of nuclei-lattice exchange / A. G. Lipson and G. H. Miley. H-D loading. An overview of experimental studies on H/Pd over-loading with thin Pd wires and different electrolytic solutions / A. Spallone ... [et al.] -- 3. Transmutations. Photon and particle emission, heat production, and surface transformation in Ni-H system / E. Campari ... [et al.]. Surface analysis of hydrogen-loaded nickel alloys / E. Campari ... [et al.]. Low-energy nuclear reactions and the leptonic monopole / G. Lochak and L. Urutskoev. Results of analysis of Ti foil after glow discharge with deuterium / I. B. Savvatimova and D. V. Gavritenkov. Enhancement mechanisms of low-energy nuclear reactions / F. A. Gareev, I. E. Zhidkova, and Y. L. Ratis. Co-deposition of palladium with hydrogen isotopes / J. Dash and A. Ambadkar. Variation of the concentration of isotopes copper and zinc in human plasmas of patients affected by cancer / A. Triassi. Transmutation of metal at low energy in a confined plasma in water / D. Cirillo and V. Iorio. The conditions and realization of self-similar Coulomb collapse of condensed target and low-energy laboratory nucleosynthesis / S. V. Adamenko and V. I. Vysotskii. The spatial structure of water and the problem of controlled low-energy nuclear reactions in water matrix / V. I. Vysotskii and A. A. Kornilova. Experiments on controlled decontamination of water mixture of longlived active isotopes in biological cells / V. I. Vysotskii. Assessment of the biological effects of "strange" radiation / E. A. Pryakhin ... [et al.]. Possible nuclear transmutation of nitrogen in the earth's atmosphere / M. Fukuhara. Evidences on the occurrence of LENR-type processes in alchemical transmutations / J. Pérez-Pariente. History of the discovery of transmutation at Texas A&M University / J. O.-M. Bockris -- 4. Theory. Quantum electrodynamics. Concerning the modeling of systems in terms of quantum electro dynamics: the special case of "cold fusion" / M. Abyaneh ... [et al.]. Screening. Theoretical model of the probability of fusion between deuterons within deformed lattices with microcracks at room temperature / F. Fulvio. Resonant tunnelling. Effective interaction potential in the deuterium plasma and multiple resonance scattering / T. Toimela. Multiple scattering theory and condensed matter nuclear science - "super-absorption" in a crystal latice / X. Z. Li ... [et al.]. Ion band states. Framework for understanding LENR processes, using conventional condensed matter physics / S. R. Chubb. I. Bloch ions / T. A. Chubb. II. Inhibited diffusion driven surface transmutations / T. A. Chubb. III. Bloch nuclides, Iwamura transmutations, and Oriani showers / T. A. Chubb. Bose-Einstein condensate. Theoretical study of nuclear reactions induced by Bose-Einstein condensation in Pd / K.-I. Tsuchiya and H. Okumura. Proposal for new experimental tests of the Bose-Einstein condensation mechanism for low-energy nuclear reaction and transmutation processes in deuterium loaded micro- and nano-scale cavities / Y. E. Kim ... [et al.]. Mixtures of charged bosons confined in harmonic traps and Bose-Einstein condensation mechanism for low-energy nuclear reactions and transmutation processes in condensed matters / Y. E. Kim and A. L. Zubarev. Alternative interpretation of low-energy nuclear reaction processes with deuterated metals based on the Bose-Einstein condensation mechanism / Y. E. Kim and T. O. Passell. Multi-body fusion. [symbol]He/[symbol]He production ratios by tetrahedral symmetric condensation / A. Takahashi. Phonon coupling. Phonon-exchange models: some new results / P. L. Hagelstein. Neutron clusters. Cold fusion phenomenon and solid state nuclear physics / H. Kozima. Neutrinos, magnetic monopoles. Neutrino-driven nuclear reactions of cold fusion and transmutation / V. Filimonov. Light monopoles theory: an overview of their effects in physics, chemistry, biology, and nuclear science (weak interactions) / G. Lochak. Electrons clusters and magnetic monopoles / M. Rambaut. Others. Effects of atomic electrons on nuclear stability and radioactive decay / D. V. Filippov, L. I. Urutskoev, and A. A. Rukhadze. Search for erzion nuclear catalysis chains from cosmic ray erzions stopping in organic scintillator / Yu. N. Bazhutov and E. V. Pletnikov. Low-energy nuclear reactions resulting as picometer interactions with similarity to K-shell electron capture / H. Hora ... [et al.] -- 5. Other topics. On the possible magnetic mechanism of shortening the runaway of RBMK-1000 reactor at Chernobyl Nuclear Power Plant / D. V. Filippov ... [et al.]. Cold fusion in the context of a scientific revolution in physics: history and economic ramifications / E. Lewis. The nucleovoltaic cell / D. D. Moon. Introducing the book "Cold Fusion and the Future" / J. Rothwell. Recent cold fusion claims: are they valid? / L. Kowalski. History of attempts to publish a paper / L. Kowalski.

Analysis of the Reactor Physics of Low-Enrichment Fuel for the INL Advanced Test Reactor in support of RERTR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mark DeHart; William Skerjanc; Sean Morrell

2012-06-01

Analysis of the performance of the ATR with a LEU fuel design shows promise in terms of a core design that will yield the same neutron sources in target locations. A proposed integral cladding burnable absorber design appears to meet power profile requirements that will satisfy power distributions for safety limits. Performance of this fuel design is ongoing; the current work is the initial evaluation of the core performance of this fuel design with increasing burnup. Results show that LEU fuel may have a longer lifetime that HEU fuel however, such limits may be set by mechanical performance of themore » fuel rather that available reactivity. Changes seen in the radial fuel power distribution with burnup in LEU fuel will require further study to ascertain the impact on neutron fluxes in target locations. Source terms for discharged fuel have also been studied. By its very nature, LEU fuel produces much more plutonium than is present in HEU fuel at discharge. However, the effect of the plutonium inventory appears to have little affect on radiotoxicity or decay heat in the fuel.« less

Investigation of polymer electrolyte membrane fuel cell internal behaviour during long term operation and its use in prognostics

NASA Astrophysics Data System (ADS)

Mao, Lei; Jackson, Lisa; Jackson, Tom

2017-09-01

This paper investigates the polymer electrolyte membrane (PEM) fuel cell internal behaviour variation at different operating condition, with characterization test data taken at predefined inspection times, and uses the determined internal behaviour evolution to predict the future PEM fuel cell performance. For this purpose, a PEM fuel cell behaviour model is used, which can be related to various fuel cell losses. By matching the model to the collected polarization curves from the PEM fuel cell system, the variation of fuel cell internal behaviour can be obtained through the determined model parameters. From the results, the source of PEM fuel cell degradation during its lifetime at different conditions can be better understood. Moreover, with determined fuel cell internal behaviour, the future fuel cell performance can be obtained by predicting the future model parameters. By comparing with prognostic results using adaptive neuro fuzzy inference system (ANFIS), the proposed prognostic analysis can provide better predictions for PEM fuel cell performance at dynamic condition, and with the understanding of variation in PEM fuel cell internal behaviour, mitigation strategies can be designed to extend the fuel cell performance.

40 CFR 600.209-12 - Calculation of vehicle-specific 5-cycle fuel economy and CO2 emission values for a model type.

Code of Federal Regulations, 2012 CFR

2012-07-01

... city and highway fuel economy and CO2 emission values from the tests performed using gasoline or diesel test fuel. (ii) If 5-cycle testing was performed on the alcohol or natural gas test fuel, calculate the city and highway fuel economy and CO2 emission values from the tests performed using alcohol or natural...

40 CFR 600.209-12 - Calculation of vehicle-specific 5-cycle fuel economy and CO2 emission values for a model type.

Code of Federal Regulations, 2014 CFR

2014-07-01

... city and highway fuel economy and CO2 emission values from the tests performed using gasoline or diesel test fuel. (ii) If 5-cycle testing was performed on the alcohol or natural gas test fuel, calculate the city and highway fuel economy and CO2 emission values from the tests performed using alcohol or natural...

40 CFR 600.209-12 - Calculation of vehicle-specific 5-cycle fuel economy and CO2 emission values for a model type.

Code of Federal Regulations, 2013 CFR

2013-07-01

... city and highway fuel economy and CO2 emission values from the tests performed using gasoline or diesel test fuel. (ii) If 5-cycle testing was performed on the alcohol or natural gas test fuel, calculate the city and highway fuel economy and CO2 emission values from the tests performed using alcohol or natural...

Emissions Prediction and Measurement for Liquid-Fueled TVC Combustor with and without Water Injection

NASA Technical Reports Server (NTRS)

Brankovic, A.; Ryder, R. C., Jr.; Hendricks, R. C.; Liu, N.-S.; Shouse, D. T.; Roquemore, W. M.

2005-01-01

An investigation is performed to evaluate the performance of a computational fluid dynamics (CFD) tool for the prediction of the reacting flow in a liquid-fueled combustor that uses water injection for control of pollutant emissions. The experiment consists of a multisector, liquid-fueled combustor rig operated at different inlet pressures and temperatures, and over a range of fuel/air and water/fuel ratios. Fuel can be injected directly into the main combustion airstream and into the cavities. Test rig performance is characterized by combustor exit quantities such as temperature and emissions measurements using rakes and overall pressure drop from upstream plenum to combustor exit. Visualization of the flame is performed using gray scale and color still photographs and high-frame-rate videos. CFD simulations are performed utilizing a methodology that includes computer-aided design (CAD) solid modeling of the geometry, parallel processing over networked computers, and graphical and quantitative post-processing. Physical models include liquid fuel droplet dynamics and evaporation, with combustion modeled using a hybrid finite-rate chemistry model developed for Jet-A fuel. CFD and experimental results are compared for cases with cavity-only fueling, while numerical studies of cavity and main fueling was also performed. Predicted and measured trends in combustor exit temperature, CO and NOx are in general agreement at the different water/fuel loading rates, although quantitative differences exist between the predictions and measurements.

Utilization of alternative fuels in diesel engines

NASA Technical Reports Server (NTRS)

Lestz, S. A.

1984-01-01

Performance and emission data are collected for various candidate alternate fuels and compare these data to that for a certified petroleum based number two Diesel fuel oil. Results for methanol, ethanol, four vegetable oils, two shale derived oils, and two coal derived oils are reported. Alcohol fumigation does not appear to be a practical method for utilizing low combustion quality fuels in a Diesel engine. Alcohol fumigation enhances the bioactivity of the emitted exhaust particles. While it is possible to inject many synthetic fuels using the engine stock injection system, wholly acceptable performance is only obtained from a fuel whose specifications closely approach those of a finished petroleum based Diesel oil. This is illustrated by the contrast between the poor performance of the unupgraded coal derived fuel blends and the very good performance of the fully refined shale derived fuel.

Systems Analysis Of Advanced Coal-Based Power Plants

NASA Technical Reports Server (NTRS)

Ferrall, Joseph F.; Jennings, Charles N.; Pappano, Alfred W.

1988-01-01

Report presents appraisal of integrated coal-gasification/fuel-cell power plants. Based on study comparing fuel-cell technologies with each other and with coal-based alternatives and recommends most promising ones for research and development. Evaluates capital cost, cost of electricity, fuel consumption, and conformance with environmental standards. Analyzes sensitivity of cost of electricity to changes in fuel cost, to economic assumptions, and to level of technology. Recommends further evaluation of integrated coal-gasification/fuel-cell integrated coal-gasification/combined-cycle, and pulverized-coal-fired plants. Concludes with appendixes detailing plant-performance models, subsystem-performance parameters, performance goals, cost bases, plant-cost data sheets, and plant sensitivity to fuel-cell performance.

Alternate-Fueled Combustor-Sector Performance. Parts A and B; (A) Combustor Performance; (B) Combustor Emissions

NASA Technical Reports Server (NTRS)

Shouse, D. T.; Hendricks, R. C.; Lynch, A.; Frayne, C. W.; Stutrud, J. S.; Corporan, E.; Hankins, T.

2012-01-01

Alternate aviation fuels for military or commercial use are required to satisfy MIL-DTL-83133F(2008) or ASTM D 7566 (2010) standards, respectively, and are classified as "drop-in" fuel replacements. To satisfy legacy issues, blends to 50% alternate fuel with petroleum fuels are certified individually on the basis of processing and assumed to be feedstock agnostic. Adherence to alternate fuels and fuel blends requires "smart fueling systems" or advanced fuel-flexible systems, including combustors and engines, without significant sacrifice in performance or emissions requirements. This paper provides preliminary performance (Part A) and emissions and particulates (Part B) combustor sector data. The data are for nominal inlet conditions at 225 psia and 800 F (1.551 MPa and 700 K), for synthetic-paraffinic-kerosene- (SPK-) type (Fisher-Tropsch (FT)) fuel and blends with JP-8+100 relative to JP-8+100 as baseline fueling. Assessments are made of the change in combustor efficiency, wall temperatures, emissions, and luminosity with SPK of 0%, 50%, and 100% fueling composition at 3% combustor pressure drop. The performance results (Part A) indicate no quantifiable differences in combustor efficiency, a general trend to lower liner and higher core flow temperatures with increased FT fuel blends. In general, emissions data (Part B) show little differences, but with percent increase in FT-SPK-type fueling, particulate emissions and wall temperatures are less than with baseline JP-8. High-speed photography illustrates both luminosity and combustor dynamic flame characteristics.

Separation of Long-Lived Fission Products Tc-99 and I-129 from Synthetic Effluents by Crown Ethers

DOE Office of Scientific and Technical Information (OSTI.GOV)

Paviet-Hartmann, P.; Hartmann, T.

2006-07-01

To minimize significantly the radio-toxic inventory of nuclear geological repositories to come as well as to reduce the potential of radionuclides migration and to minimize long-term exposure, the concept of partitioning and transmutation (P/T) of nuclear waste is currently discussed. Transmutation offers the possibility to convert radio-toxic radionuclides with long half-lives into radionuclides of shorter half-lives, less toxic isotopes, or even into stable isotopes. Besides the most prominent isotopes of neptunium, plutonium, americium, and curium, the long-lived fission products Tc-99 and I-129 (half-lives of 2.13 x 10{sup 5} years, and 1.57 x 10{sup 7} years, respectively) are promising candidates formore » transmutation in order to prevent their migration from a nuclear repository. Partitioning and transmutation of the most radio-toxic radionuclides will not only minimize the nuclear waste load but most importantly will significantly reduce the long-term radio-toxic hazard of nuclear waste repositories to come. Prior to the deployment of partitioning and transmutation, selective extraction techniques are required to separate the radionuclides of concern. Since the discovery of crown ethers by C. Pedersen, various applications of crown ethers have drawn much attention. Although liquid-liquid extraction of alkali and alkali earth metals by crown ethers has been extensively studied, little data is available on the extraction of Tc-99 and I-129 by crown ethers. The methods developed herein for the specific extraction of Tc-99 and I-129 provide recommendations in support of their selectively extraction from liquid radioactive waste streams, mainly ILW. We report data on the solvent extraction of Tc-99 and I-129 from synthetic effluents by six crown ethers of varying cavity dimensions and derivatization. To satisfy the needs of new extractant systems we are demonstrating that crown ether (CE) based systems have the potential to serve as selective extractants for the separation of these long lived radionuclides from high level nuclear waste (HLW), intermediate level nuclear waste (ILW), and low level nuclear waste (LLW) streams. The experimental results show that dibenzo-18-crown-6 (DB 18C6) is highly selective towards Tc-99, and dicyclohexano-18-crown-6 (DC18C6) is highly selective towards I-129. The nature of the diluent was examined and was shown to be the most influential variable in controlling the extraction coefficients of Tc-99 and I-129. Therefore the addition of polar diluent acetone to non-polar diluent toluene enhanced the distribution coefficient of Tc-99 (DTc) was by a factor of 30. For I-129, the best extraction yield was obtained after introducing tetrachloroethane. Through the process, by a single extraction step, 85 % to 95 % of Tc-99 was extracted from synthetic effluents, while 84 % to 88 % of I-129 was extracted from different acidic media. The extraction by crown ether is a fairly rapid process and the total preparation time of the chemical separation takes about 20 minutes for a batch of eight samples. (authors)« less

76 FR 3587 - Standards of Performance for Fossil-Fuel-Fired, Electric Utility, Industrial-Commercial...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-01-20

... Standards of Performance for Fossil-Fuel-Fired, Electric Utility, Industrial-Commercial-Institutional, and... Fossil fuel-fired electric utility steam generating units. Federal Government 22112 Fossil fuel-fired... 22112 Fossil fuel-fired electric utility steam generating units owned by municipalities. 921150 Fossil...

Impact of future fuel properties on aircraft engines and fuel systems

NASA Technical Reports Server (NTRS)

Rudey, R. A.; Grobman, J. S.

1978-01-01

From current projections of the availability of high-quality petroleum crude oils, it is becoming increasingly apparent that the specifications for hydrocarbon jet fuels may have to be modified. The problems that are most likely to be encountered as a result of these modifications relate to engine performance, component durability and maintenance, and aircraft fuel-system performance. The effect on engine performance will be associated with changes in specific fuel consumption, ignition at relight limits, at exhaust emissions. Durability and maintenance will be affected by increases in combustor liner temperatures, carbon deposition, gum formation in fuel nozzles, and erosion and corrosion of turbine blades and vanes. Aircraft fuel-system performance will be affected by increased deposits in fuel-system heat exchangers and changes in the pumpability and flowability of the fuel. The severity of the potential problems is described in terms of the fuel characteristics most likely to change in the future. Recent data that evaluate the ability of current-technology aircraft to accept fuel specification changes are presented, and selected technological advances that can reduce the severity of the problems are described and discussed.

High-level radioactive waste management alternatives

DOE Office of Scientific and Technical Information (OSTI.GOV)

None

1974-05-01

A summary of a comprehensive overview study of potential alternatives for long-term management of high-level radioactive waste is presented. The concepts studied included disposal in geologic formations, disposal in seabeds, disposal in ice caps, disposal into space, and elimination by transmutation. (TFD)

Clinical Investigation Program Report

DTIC Science & Technology

1992-10-01

alchemists claim of transmutation of metals by asserting the fundamental differen- ii ces of metals. His medical masterwork was the Canon which remained an...1992. Blount BW: Sexually transmitted disease update. Am Acad Fam Phy, Washington, DC, Oct 1991. Blount BW: A comparison of Family Practice content

Christian Realism’s Response to International Terrorism

DTIC Science & Technology

2002-04-01

defense which cannot be transmuted into instruments of aggression. The frustrations of the average man, who can never realise the power and the...historical existence, tensions have their root in natural, geographic, economic, racial, national and sexual conditions. But since it interprets

AGR-1 Compact 1-3-1 Post-Irradiation Examination Results

DOE Office of Scientific and Technical Information (OSTI.GOV)

Demkowicz, Paul Andrew

The Advanced Gas Reactor (AGR) Fuel Development and Qualification Program was established to perform the requisite research and development on tristructural isotropic (TRISO) coated particle fuel to support deployment of a high-temperature gas-cooled reactor (HTGR). The work continues as part of the Advanced Reactor Technologies (ART) TRISO Fuel program. The overarching program goal is to provide a baseline fuel qualification data set to support licensing and operation of an HTGR. To achieve these goals, the program includes the elements of fuel fabrication, irradiation, post-irradiation examination (PIE) and safety testing, fuel performance modeling, and fission product transport (INL 2015). A seriesmore » of fuel irradiation experiments is being planned and conducted in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). These experiments will provide data on fuel performance under irradiation, support fuel process development, qualify the fuel for normal operating conditions, provide irradiated fuel for safety testing, and support the development of fuel performance and fission product transport models. The first of these irradiation tests, designated AGR-1, began in the ATR in December 2006 and ended in November 2009. This experiment was conducted primarily to act as a shakedown test of the multicapsule test train design and provide early data on fuel performance for use in fuel fabrication process development. It also provided samples for post-irradiation safety testing, where fission product retention of the fuel at high temperatures will be experimentally measured. The capsule design and details of the AGR-1 experiment have been presented previously (Grover, Petti, and Maki 2010, Maki 2009).« less

AGR-1 Compact 5-3-1 Post-Irradiation Examination Results

DOE Office of Scientific and Technical Information (OSTI.GOV)

Demkowicz, Paul; Harp, Jason; Winston, Phil

The Advanced Gas Reactor (AGR) Fuel Development and Qualification Program was established to perform the requisite research and development on tristructural isotropic (TRISO) coated particle fuel to support deployment of a high-temperature gas-cooled reactor (HTGR). The work continues as part of the Advanced Reactor Technologies (ART) TRISO Fuel program. The overarching program goal is to provide a baseline fuel qualification data set to support licensing and operation of an HTGR. To achieve these goals, the program includes the elements of fuel fabrication, irradiation, post-irradiation examination (PIE) and safety testing, fuel performance, and fission product transport (INL 2015). A series ofmore » fuel irradiation experiments is being planned and conducted in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). These experiments will provide data on fuel performance under irradiation, support fuel process development, qualify the fuel for normal operating conditions, provide irradiated fuel for safety testing, and support the development of fuel performance and fission product transport models. The first of these irradiation tests, designated AGR-1, began in the ATR in December 2006 and ended in November 2009. This experiment was conducted primarily to act as a shakedown test of the multicapsule test train design and provide early data on fuel performance for use in fuel fabrication process development. It also provided samples for post-irradiation safety testing, where fission product retention of the fuel at high temperatures will be experimentally measured. The capsule design and details of the AGR-1 experiment have been presented previously.« less

Light Water Breeder Reactor fuel rod design and performance characteristics (LWBR Development Program)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Campbell, W.R.; Giovengo, J.F.

1987-10-01

Light Water Breeder Reactor (LWBR) fuel rods were designed to provide a reliable fuel system utilizing thorium/uranium-233 mixed-oxide fuel while simultaneously minimizing structural material to enhance fuel breeding. The fuel system was designed to be capable of operating successfully under both load follow and base load conditions. The breeding objective required thin-walled, low hafnium content Zircaloy cladding, tightly spaced fuel rods with a minimum number of support grid levels, and movable fuel rod bundles to supplant control rods. Specific fuel rod design considerations and their effects on performance capability are described. Successful completion of power operations to over 160 percentmore » of design lifetime including over 200 daily load follow cycles has proven the performance capability of the fuel system. 68 refs., 19 figs., 44 tabs.« less

Low contaminant formic acid fuel for direct liquid fuel cell

DOEpatents

Masel, Richard I [Champaign, IL; Zhu, Yimin [Urbana, IL; Kahn, Zakia [Palatine, IL; Man, Malcolm [Vancouver, CA

2009-11-17

A low contaminant formic acid fuel is especially suited toward use in a direct organic liquid fuel cell. A fuel of the invention provides high power output that is maintained for a substantial time and the fuel is substantially non-flammable. Specific contaminants and contaminant levels have been identified as being deleterious to the performance of a formic acid fuel in a fuel cell, and embodiments of the invention provide low contaminant fuels that have improved performance compared to known commercial bulk grade and commercial purified grade formic acid fuels. Preferred embodiment fuels (and fuel cells containing such fuels) including low levels of a combination of key contaminants, including acetic acid, methyl formate, and methanol.

40 CFR 600.206-08 - Calculation and use of FTP-based and HFET-based fuel economy values for vehicle configurations.

Code of Federal Regulations, 2013 CFR

2013-07-01

... fuel economy values from the tests performed using gasoline or diesel test fuel. (ii) Calculate the city, highway, and combined fuel economy values from the tests performed using alcohol or natural gas test fuel. (b) If only one equivalent petroleum-based fuel economy value exists for an electric...

40 CFR 600.206-08 - Calculation and use of FTP-based and HFET-based fuel economy values for vehicle configurations.

Code of Federal Regulations, 2011 CFR

2011-07-01

..., highway, and combined fuel economy values from the tests performed using gasoline or diesel test fuel. (ii) Calculate the city, highway, and combined fuel economy values from the tests performed using alcohol or natural gas test fuel. (b) If only one equivalent petroleum-based fuel economy value exists for an...

40 CFR 600.206-08 - Calculation and use of FTP-based and HFET-based fuel economy values for vehicle configurations.

Code of Federal Regulations, 2012 CFR

2012-07-01

... fuel economy values from the tests performed using gasoline or diesel test fuel. (ii) Calculate the city, highway, and combined fuel economy values from the tests performed using alcohol or natural gas test fuel. (b) If only one equivalent petroleum-based fuel economy value exists for an electric...

Pre-irradiation testing and analysis to support the LWRS Hybrid SiC-CMC-Zircaloy-04 unfueled rodlet irradiation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Isabella J van Rooyen

2012-09-01

Nuclear fuel performance is a significant driver of nuclear power plant operational performance, safety, economics and waste disposal requirements. The Advanced Light Water Reactor (LWR) Nuclear Fuel Development Pathway focuses on improving the scientific knowledge basis to enable the development of high-performance, high burn-up fuels with improved safety and cladding integrity and improved nuclear fuel cycle economics. To achieve significant improvements, fundamental changes are required in the areas of nuclear fuel composition, cladding integrity, and fuel/cladding interaction.

Pre-irradiation testing and analysis to support the LWRS Hybrid SiC-CMC-Zircaloy-04 unfueled rodlet irradiation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Isabella J van Rooyen

2013-01-01

Nuclear fuel performance is a significant driver of nuclear power plant operational performance, safety, economics and waste disposal requirements. The Advanced Light Water Reactor (LWR) Nuclear Fuel Development Pathway focuses on improving the scientific knowledge basis to enable the development of high-performance, high burn-up fuels with improved safety and cladding integrity and improved nuclear fuel cycle economics. To achieve significant improvements, fundamental changes are required in the areas of nuclear fuel composition, cladding integrity, and fuel/cladding interaction.

Selection of optimal sensors for predicting performance of polymer electrolyte membrane fuel cell

NASA Astrophysics Data System (ADS)

Mao, Lei; Jackson, Lisa

2016-10-01

In this paper, sensor selection algorithms are investigated based on a sensitivity analysis, and the capability of optimal sensors in predicting PEM fuel cell performance is also studied using test data. The fuel cell model is developed for generating the sensitivity matrix relating sensor measurements and fuel cell health parameters. From the sensitivity matrix, two sensor selection approaches, including the largest gap method, and exhaustive brute force searching technique, are applied to find the optimal sensors providing reliable predictions. Based on the results, a sensor selection approach considering both sensor sensitivity and noise resistance is proposed to find the optimal sensor set with minimum size. Furthermore, the performance of the optimal sensor set is studied to predict fuel cell performance using test data from a PEM fuel cell system. Results demonstrate that with optimal sensors, the performance of PEM fuel cell can be predicted with good quality.

Low-temperature irradiation behavior of uranium-molybdenum alloy dispersion fuel

NASA Astrophysics Data System (ADS)

Meyer, M. K.; Hofman, G. L.; Hayes, S. L.; Clark, C. R.; Wiencek, T. C.; Snelgrove, J. L.; Strain, R. V.; Kim, K.-H.

2002-08-01

Irradiation tests have been conducted to evaluate the performance of a series of high-density uranium-molybdenum (U-Mo) alloy, aluminum matrix dispersion fuels. Fuel plates incorporating alloys with molybdenum content in the range of 4-10 wt% were tested. Two irradiation test vehicles were used to irradiate low-enrichment fuels to approximately 40 and 70 at.% 235U burnup in the advanced test reactor at fuel temperatures of approximately 65 °C. The fuel particles used to fabricate dispersion specimens for most of the test were produced by generating filings from a cast rod. In general, fuels with molybdenum contents of 6 wt% or more showed stable in-reactor fission gas behavior, exhibiting a distribution of small, stable gas bubbles. Fuel particle swelling was moderate and decreased with increasing alloy content. Fuel particles with a molybdenum content of 4 wt% performed poorly, exhibiting extensive fuel-matrix interaction and the growth of relatively large fission gas bubbles. Fuel particles with 4 or 6 wt% molybdenum reacted more rapidly with the aluminum matrix than those with higher-alloy content. Fuel particles produced by an atomization process were also included in the test to determine the effect of fuel particle morphology and microstructure on fuel performance for the U-10Mo composition. Both of the U-10Mo fuel particle types exhibited good irradiation performance, but showed visible differences in fission gas bubble nucleation and growth behavior.

TRISO Fuel Performance: Modeling, Integration into Mainstream Design Studies, and Application to a Thorium-fueled Fusion-Fission Hybrid Blanket

DOE Office of Scientific and Technical Information (OSTI.GOV)

Powers, Jeffrey James

2011-11-30

This study focused on creating a new tristructural isotropic (TRISO) coated particle fuel performance model and demonstrating the integration of this model into an existing system of neutronics and heat transfer codes, creating a user-friendly option for including fuel performance analysis within system design optimization and system-level trade-off studies. The end product enables both a deeper understanding and better overall system performance of nuclear energy systems limited or greatly impacted by TRISO fuel performance. A thorium-fueled hybrid fusion-fission Laser Inertial Fusion Energy (LIFE) blanket design was used for illustrating the application of this new capability and demonstrated both the importancemore » of integrating fuel performance calculations into mainstream design studies and the impact that this new integrated analysis had on system-level design decisions. A new TRISO fuel performance model named TRIUNE was developed and verified and validated during this work with a novel methodology established for simulating the actual lifetime of a TRISO particle during repeated passes through a pebble bed. In addition, integrated self-consistent calculations were performed for neutronics depletion analysis, heat transfer calculations, and then fuel performance modeling for a full parametric study that encompassed over 80 different design options that went through all three phases of analysis. Lastly, side studies were performed that included a comparison of thorium and depleted uranium (DU) LIFE blankets as well as some uncertainty quantification work to help guide future experimental work by assessing what material properties in TRISO fuel performance modeling are most in need of improvement. A recommended thorium-fueled hybrid LIFE engine design was identified with an initial fuel load of 20MT of thorium, 15% TRISO packing within the graphite fuel pebbles, and a 20cm neutron multiplier layer with beryllium pebbles in flibe molten salt coolant. It operated at a system power level of 2000 MW th, took about 3.5 years to reach full plateau power, and was capable of an End of Plateau burnup of 38.7 %FIMA if considering just the neutronic constraints in the system design; however, fuel performance constraints led to a maximum credible burnup of 12.1 %FIMA due to a combination of internal gas pressure and irradiation effects on the TRISO materials (especially PyC) leading to SiC pressure vessel failures. The optimal neutron spectrum for the thorium-fueled blanket options evaluated seemed to favor a hard spectrum (low but non-zero neutron multiplier thicknesses and high TRISO packing fractions) in terms of neutronic performance but the fuel performance constraints demonstrated that a significantly softer spectrum would be needed to decrease the rate of accumulation of fast neutron fluence in order to improve the maximum credible burnup the system could achieve.« less

Irradiation performance of HTGR recycle fissile fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Homan, F.J.; Long, E.L. Jr.

1976-08-01

The irradiation performance of candidate HTGR recycle fissile fuel under accelerated testing conditions is reviewed. Failure modes for coated-particle fuels are described, and the performance of candidate recycle fissile fuels is discussed in terms of these failure modes. The bases on which UO/sub 2/ and (Th,U)O/sub 2/ were rejected as candidate recycle fissile fuels are outlined, along with the bases on which the weak-acid resin (WAR)-derived fissile fuel was selected as the reference recycle kernel. Comparisons are made relative to the irradiation behavior of WAR-derived fuels of varying stoichiometry and conclusions are drawn about the optimum stoichiometry and the rangemore » of acceptable values. Plans for future testing in support of specification development, confirmation of the results of accelerated testing by real-time experiments, and improvement in fuel performance and reliability are described.« less

High Performance Fuel Cell and Electrolyzer Membrane Electrode Assemblies (MEAs) for Space Energy Storage Systems

NASA Technical Reports Server (NTRS)

Valdez, Thomas I.; Billings, Keith J.; Kisor, Adam; Bennett, William R.; Jakupca, Ian J.; Burke, Kenneth; Hoberecht, Mark A.

2012-01-01

Regenerative fuel cells provide a pathway to energy storage system development that are game changers for NASA missions. The fuel cell/ electrolysis MEA performance requirements 0.92 V/ 1.44 V at 200 mA/cm2 can be met. Fuel Cell MEAs have been incorporated into advanced NFT stacks. Electrolyzer stack development in progress. Fuel Cell MEA performance is a strong function of membrane selection, membrane selection will be driven by durability requirements. Electrolyzer MEA performance is catalysts driven, catalyst selection will be driven by durability requirements. Round Trip Efficiency, based on a cell performance, is approximately 65%.

Spray visualization of alternative fuels at hot ambient conditions

NASA Astrophysics Data System (ADS)

Kannaiyan, Kumaran; Sadr, Reza

2017-11-01

Gas-to-Liquid (GTL) has gained significant interest as drop-in alternative jet fuel owing to its cleaner combustion characteristics. The physical and evaporation properties of GTL fuels are different from those of the conventional jet fuels. Those differences will have an effect on the spray, and in turn, the combustion performance. In this study, the non-reacting near nozzle spray dynamics such as spray cone angle, liquid sheet breakup and liquid velocity of GTL fuel will be investigated and compared with those of the conventional jet fuel. This work is a follow up of the preliminary study performed at atmospheric ambient conditions where differences were observed in the near nozzle spray characteristics between the fuels. Whereas, in this study the spray visualization will be performed in a hot and inert environment to account for the difference in evaporation characteristics of the fuels. The spray visualization images will be captured using the shadowgraph technique. A rigorous statistical analysis of the images will be performed to compare the spray dynamics between the fuels.

NREL Fuels and Engines R&D Revs Up Vehicle Efficiency, Performance (Text

Science.gov Websites

Version) | News | NREL Fuels and Engines R&D Revs Up Vehicle Efficiency, Performance (Text Version) NREL Fuels and Engines R&D Revs Up Vehicle Efficiency, Performance (Text Version) NREL's combustion to the evolution of how fuels interact with engine and vehicle design. This is a text version of

Effects of chemical equilibrium on turbine engine performance for various fuels and combustor temperatures

NASA Technical Reports Server (NTRS)

Tran, Donald H.; Snyder, Christopher A.

1992-01-01

A study was performed to quantify the differences in turbine engine performance with and without the chemical dissociation effects for various fuel types over a range of combustor temperatures. Both turbojet and turbofan engines were studied with hydrocarbon fuels and cryogenic, nonhydrocarbon fuels. Results of the study indicate that accuracy of engine performance decreases when nonhydrocarbon fuels are used, especially at high temperatures where chemical dissociation becomes more significant. For instance, the deviation in net thrust for liquid hydrogen fuel can become as high as 20 percent at 4160 R. This study reveals that computer central processing unit (CPU) time increases significantly when dissociation effects are included in the cycle analysis.

Multidisciplinary Simulation of Graphite-Composite and Cermet Fuel Elements for NTP Point of Departure Designs

NASA Technical Reports Server (NTRS)

Stewart, Mark E.; Schnitzler, Bruce G.

2015-01-01

This paper compares the expected performance of two Nuclear Thermal Propulsion fuel types. High fidelity, fluid/thermal/structural + neutronic simulations help predict the performance of graphite-composite and cermet fuel types from point of departure engine designs from the Nuclear Thermal Propulsion project. Materials and nuclear reactivity issues are reviewed for each fuel type. Thermal/structural simulations predict thermal stresses in the fuel and thermal expansion mis-match stresses in the coatings. Fluid/thermal/structural/neutronic simulations provide predictions for full fuel elements. Although NTP engines will utilize many existing chemical engine components and technologies, nuclear fuel elements are a less developed engine component and introduce design uncertainty. Consequently, these fuel element simulations provide important insights into NTP engine performance.

U.S. Department of Energy-Funded Performance Validation of Fuel Cell Material Handling Equipment (Presentation)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kurtz, J.; Sprik, S.; Ramsden, T.

2013-11-01

This webinar presentation to the UK Hydrogen and Fuel Cell Association summarizes how the U.S. Department of Energy is enabling early fuel cell markets; describes objectives of the National Fuel Cell Technology Evaluation Center; and presents performance status of fuel cell material handling equipment.

40 CFR 60.43 - Standard for sulfur dioxide (SO2).

Code of Federal Regulations, 2010 CFR

2010-07-01

... (CONTINUED) STANDARDS OF PERFORMANCE FOR NEW STATIONARY SOURCES Standards of Performance for Fossil-Fuel... from liquid fossil fuel or liquid fossil fuel and wood residue. (2) 520 ng/J heat input (1.2 lb/MMBtu) derived from solid fossil fuel or solid fossil fuel and wood residue, except as provided in paragraph (e...

40 CFR 60.44 - Standard for nitrogen oxides (NOX).

Code of Federal Regulations, 2010 CFR

2010-07-01

... (CONTINUED) STANDARDS OF PERFORMANCE FOR NEW STATIONARY SOURCES Standards of Performance for Fossil-Fuel...) derived from gaseous fossil fuel. (2) 129 ng/J heat input (0.30 lb/MMBtu) derived from liquid fossil fuel, liquid fossil fuel and wood residue, or gaseous fossil fuel and wood residue. (3) 300 ng/J heat input (0...

40 CFR 60.43 - Standard for sulfur dioxide (SO2).

Code of Federal Regulations, 2011 CFR

2011-07-01

... (CONTINUED) STANDARDS OF PERFORMANCE FOR NEW STATIONARY SOURCES Standards of Performance for Fossil-Fuel... from liquid fossil fuel or liquid fossil fuel and wood residue. (2) 520 ng/J heat input (1.2 lb/MMBtu) derived from solid fossil fuel or solid fossil fuel and wood residue, except as provided in paragraph (e...

40 CFR 60.44 - Standard for nitrogen oxides (NOX).

Code of Federal Regulations, 2011 CFR

2011-07-01

... (CONTINUED) STANDARDS OF PERFORMANCE FOR NEW STATIONARY SOURCES Standards of Performance for Fossil-Fuel...) derived from gaseous fossil fuel. (2) 129 ng/J heat input (0.30 lb/MMBtu) derived from liquid fossil fuel, liquid fossil fuel and wood residue, or gaseous fossil fuel and wood residue. (3) 300 ng/J heat input (0...

40 CFR 600.207-93 - Calculation of fuel economy values for a model type.

Code of Federal Regulations, 2011 CFR

2011-07-01

... economy data from tests conducted on these vehicle configuration(s) at high altitude to calculate the fuel... city, highway, and combined fuel economy values from the tests performed using gasoline or diesel test fuel. (ii) Calculate the city, highway, and combined fuel economy values from the tests performed using...

Spectral tailoring device

DOEpatents

Brager, H.R.; Schenter, R.E.; Carter, L.L.; Karnesky, R.A.

1987-08-05

A spectral tailoring device for altering the neutron energy spectra and flux of neutrons in a fast reactor thereby selectively to enhance or inhibit the transmutation rate of a target metrical to form a product isotope. Neutron moderators, neutron filters, neutron absorbers and neutron reflectors may be used as spectral tailoring devices. Depending on the intended use for the device, a member from each of these four classes of materials could be used singularly, or in combination, to provide a preferred neutron energy spectra and flux of the neutrons in the region of the target material. In one embodiment of the invention, an assembly is provided for enhancing the production of isotopes, such as cobalt 60 and gadolinium 153. In another embodiment of the invention, a spectral tailoring device is disposed adjacent a target material which comprises long lived or volatile fission products and the device is used to shift the neutron energy spectra and flux of neutrons in the region of the fission products to preferentially transmute them to produce a less volatile fission product inventory. 6 figs.

Sensitivity Studies and Experimental Evaluation for Optimizing Transcurium Isotope Production

DOE PAGES

Hogle, Susan L.; Alexander, Charles W.; Burns, Jonathan D.; ...

2017-03-01

This work applies to recent initiatives at the Radiochemical Engineering Development Center at Oak Ridge National Laboratory to optimize the production of transcurium isotopes in the High Flux Isotope Reactor in such a way as to prolong the use of high quality heavy curium feedstock. By studying the sensitivity of fission and transmutation reaction rates to the neutron flux spectrum a means of increasing the fraction of (n,γ) reactions per neutron absorption is explored. Filter materials which preferentially absorb neutrons at energies considered detrimental to optimal transcurium production are identified and transmutation rates are examined with high energy resolution. Experimentalmore » capsules are irradiated employing filter materials and the resulting fission and activation products studied to validate the filtering methodology. Improvement is seen in the production efficiency of heavier curium isotopes in 244Cm and 245Cm targets, and potentially in production of 252Cf from mixed californium targets. Finally, further analysis is recommended to evaluate longer duration irradiations more representative of typical transcurium production.« less

How to Use Benchmark and Cross-section Studies to Improve Data Libraries and Models

NASA Astrophysics Data System (ADS)

Wagner, V.; Suchopár, M.; Vrzalová, J.; Chudoba, P.; Svoboda, O.; Tichý, P.; Krása, A.; Majerle, M.; Kugler, A.; Adam, J.; Baldin, A.; Furman, W.; Kadykov, M.; Solnyshkin, A.; Tsoupko-Sitnikov, S.; Tyutyunikov, S.; Vladimirovna, N.; Závorka, L.

2016-06-01

Improvements of the Monte Carlo transport codes and cross-section libraries are very important steps towards usage of the accelerator-driven transmutation systems. We have conducted a lot of benchmark experiments with different set-ups consisting of lead, natural uranium and moderator irradiated by relativistic protons and deuterons within framework of the collaboration “Energy and Transmutation of Radioactive Waste”. Unfortunately, the knowledge of the total or partial cross-sections of important reactions is insufficient. Due to this reason we have started extensive studies of different reaction cross-sections. We measure cross-sections of important neutron reactions by means of the quasi-monoenergetic neutron sources based on the cyclotrons at Nuclear Physics Institute in Řež and at The Svedberg Laboratory in Uppsala. Measurements of partial cross-sections of relativistic deuteron reactions were the second direction of our studies. The new results obtained during last years will be shown. Possible use of these data for improvement of libraries, models and benchmark studies will be discussed.

The influence of dislocation and hydrogen on thermal helium desorption behavior in Fe9Cr alloys

NASA Astrophysics Data System (ADS)

Zhu, Te; Jin, Shuoxue; Gong, Yihao; Lu, Eryang; Song, Ligang; Xu, Qiu; Guo, Liping; Cao, Xingzhong; Wang, Baoyi

2017-11-01

Transmutation helium may causes serious embrittlement which is considered to be due to helium from clustering as a bubble in materials. Suppression of transmutation helium can be achieved by introducing trapping sites such as dislocations and impurities in materials. Here, effects of intentionally-induced dislocations and hydrogen on helium migrate and release behaviors were investigated using thermal desorption spectrometry (TDS) technique applied to well-annealed and cold-worked Fe9Cr alloys irradiated by energetic helium/hydrogen ions. Synchronous desorption of helium and hydrogen was observed, and the microstructure states during helium release at different temperatures were analyzed. High thermally stable HenD type complexes formed in cold-worked specimens, resulting in the retardation of helium migration and release. The existence of hydrogen will strongly affect the thermal helium desorption which could be reflected in the TDS spectrum. It was confirmed that hydrogen retained in the specimens can result in obvious delay of helium desorption.

[Lethal effect after transmutation of 33P incorporated into bacteriophage S 13 and mechanisms of DNA double helix rupture].

PubMed

Apelgot, S

1980-04-01

The experiments show the lethal effect of the beta decay of 33P incorporated in DNA of bacteriophage S 13. The lethal efficiency is high, 0.72 at 0 degrees C and 0.55 at--197 degrees C. The presence of a radical scavenger like AET has no influence. It was found previously that for such phages with single-stranded DNA, the lethal efficiency of 32P decay is unity, and that the lethal event is a DNA single-strand break, owing to the high energy of the nucleogenic 32S atom. As the recoil energy of the 33S atom is too low to account for such a break, it is suggested that the reorganization of the phosphate molecule into sulphate is able to bring about a DNA single-strand break with an efficiency as high as 0.7, at 0 degrees C. A model for the DNA double-strand-break produced by a transmutation processes is suggested.

Plasmon transmutation: inducing new modes in nanoclusters by adding dielectric nanoparticles.

PubMed

Wen, Fangfang; Ye, Jian; Liu, Na; Van Dorpe, Pol; Nordlander, Peter; Halas, Naomi J

2012-09-12

Planar clusters of coupled plasmonic nanoparticles support nanoscale electromagnetic "hot spots" and coherent effects, such as Fano resonances, with unique near and far field signatures, currently of prime interest for sensing applications. Here we show that plasmonic cluster properties can be substantially modified by the addition of individual, discrete dielectric nanoparticles at specific locations on the cluster, introducing new plasmon modes, or transmuting existing plasmon modes to new ones, in the resulting metallodielectric nanocomplex. Depositing a single carbon nanoparticle in the junction between a pair of adjacent nanodisks induces a metal-dielectric-metal quadrupolar plasmon mode. In a ten-membered cluster, placement of several carbon nanoparticles in junctions between multiple adjacent nanoparticles introduces a collective magnetic plasmon mode into the Fano dip, giving rise to an additional subradiant mode in the metallodielectric nanocluster response. These examples illustrate that adding dielectric nanoparticles to metallic nanoclusters expands the number and types of plasmon modes supported by these new mixed-media nanoscale assemblies.

An omnidirectional retroreflector based on the transmutation of dielectric singularities.

PubMed

Ma, Yun Gui; Ong, C K; Tyc, Tomás; Leonhardt, Ulf

2009-08-01

Transformation optics is a concept used in some metamaterials to guide light on a predetermined path. In this approach, the materials implement coordinate transformations on electromagnetic waves to create the illusion that the waves are propagating through a virtual space. Transforming space by appropriately designed materials makes devices possible that have been deemed impossible. In particular, transformation optics has led to the demonstration of invisibility cloaking for microwaves, surface plasmons and infrared light. Here, on the basis of transformation optics, we implement a microwave device that would normally require a dielectric singularity, an infinity in the refractive index. To fabricate such a device, we transmute a dielectric singularity in virtual space into a mere topological defect in a real metamaterial. In particular, we demonstrate an omnidirectional retroreflector, a device for faithfully reflecting images and for creating high visibility from all directions. Our method is robust, potentially broadband and could also be applied to visible light using similar techniques.

A theoretical model for the production of Ac-225 for cancer therapy by neutron capture transmutation of Ra-226.

PubMed

Melville, G; Melville, P

2013-02-01

Radium needles that were once implanted into tumours as a cancer treatment are now obsolete and constitute a radioactive waste problem, as their half-life is 1600 years. We are investigating the reduction of radium by transmutation by bombarding Ra-226 with high-energy neutrons from a neutron source to produce Ra-225 and hence Ac-225, which can be used as a generator to produce Bi-213 for use in 'Targeted Alpha Therapy' for cancer. This paper examines the possibility of producing Ac-225 by neutron capture using a theoretical model in which neutron energy is convoluted with the corresponding neutron cross sections of Ra-226. The total integrated yield can then be obtained. This study shows that an intense beam of high-energy neutrons could initiate neutron capture on Ra-226 to produce Ra-225 and hence practical amounts of Ac-225 and a useful reduction of Ra-226. Copyright © 2012 Elsevier Ltd. All rights reserved.

Laser-Bioplasma Interaction: The Blood Type Transmutation Induced by Multiple Ultrashort Wavelength Laser Beams

NASA Astrophysics Data System (ADS)

Stefan, V. Alexander

2015-11-01

The interaction of ultrashort wavelength multi laser beams with the flowing blood thin films leads to the transmutation of the blood types A, B, and AB into O type. This is a novel mechanism of importance for the transfusion medicine. Laser radiation is in resonance with the eigen-frequency modes of the antigen proteins and forces the proteins to parametrically oscillate until they get kicked out from the surface. The stripping away of antigens is done by the scanning-multiple-lasers of a high repetition rate in the blue-purple frequency domain. The guiding-lasers are in the red-green frequency domain. The laser force, (parametric interaction with the antigen eigen-oscillation), upon the antigen protein molecule must exceed its weight. The scanning laser beam is partially reflected as long as the antigen(s) is not eliminated. The process of the protein detachment can last a few minutes. Supported by Nikola Tesla Labs., Stefan University.

Study of radioactive impurities in neutron transmutation doped germanium

NASA Astrophysics Data System (ADS)

Mathimalar, S.; Dokania, N.; Singh, V.; Nanal, V.; Pillay, R. G.; Shrivastava, A.; Jagadeesan, K. C.; Thakare, S. V.

2015-02-01

A program to develop low temperature (mK) sensors with neutron transmutation doped Ge for rare event studies with a cryogenic bolometer has been initiated. For this purpose, semiconductor grade Ge wafers are irradiated with thermal neutron flux from Dhruva reactor at Bhabha Atomic Research Centre (BARC), Mumbai. Spectroscopic studies of irradiated samples have revealed that the environment of the capsule used for irradiating the sample leads to significant levels of 65Zn, 110mAg and 182Ta impurities, which can be reduced by chemical etching of approximately 50 μm thick surface layer. From measurements of the etched samples in the low background counting setup, activity due to trace impurities of 123Sb in bulk Ge is estimated to be 1 Bq / g after irradiation. These estimates indicate that in order to use the NTD Ge sensors for rare event studies, a cooldown period of 2 years would be necessary to reduce the radioactive background to ≤ 1 mBq / g.

Study on strontium isotope abundance-ratio measurements by using a 13-MeV proton beam

NASA Astrophysics Data System (ADS)

Jeong, Cheol-Ki; Jang, Han; Lee, Goung-Jin

2016-09-01

The Rb-Sr dating method is used in dating Paleozoic and Precambrian rocks. This method measures the 87Rb and the 87Sr concentrations by using thermal ionization mass spectrometry (TIMS) [J. Hefne et al., Inter. J. Phys. Sci. 3(1), 28 (2008)]. In addition, it calculates the initial 87Sr/86Sr ratio to increase the reliability of Rb-Sr dating. In this study, the 87Sr/86Sr ratio was measured by using a 13-MeV proton accelerator. Proton kinetic energies are in the range of tens of megaelectronvolts, and protons have large absorption cross-sections for ( p, n) reactions with most substances. After absorbing a proton with such a high kinetic energy, an element is converted into a nuclide with its atomic number increased by one via nuclear transmutation. These nuclides usually have short half-lives and return to the original state through radioactive decay. When a strontium sample is irradiated with protons, nuclear transmutation occurs; thus, the strontium isotope present in the sample changes to a yttrium isotope, which is an activated radioisotope. Based on this, the 87Sr/86Sr ratio was calculated by analyzing the gamma-rays emitted by each yttrium isotope. The KIRAMS-13 cyclotron at the Cyclotron Center of Chosun University, where 13-MeV protons can be extracted, was utilized in our experiment. The 87Sr/86Sr isotope ratio was computed for samples irradiated with these protons, and the result was similar to the isotope ratio for the Standard Reference Material, i.e., 98.2 ± 3.4%. As part of the analysis, proton activation analyses were performed using 13-MeV protons, and the experimental results of this research suggest a possible approach for measuring the strontium-isotope abundance ratio of samples.

Spent nuclear fuel integrity during dry storage - performance tests and demonstrations

DOE Office of Scientific and Technical Information (OSTI.GOV)

McKinnon, M.A.; Doherty, A.L.

1997-06-01

This report summarizes the results of fuel integrity surveillance determined from gas sampling during and after performance tests and demonstrations conducted from 1983 through 1996 by or in cooperation with the US DOE Office of Commercial Radioactive Waste Management (OCRWM). The cask performance tests were conducted at Idaho National Engineering Laboratory (INEL) between 1984 and 1991 and included visual observation and ultrasonic examination of the condition of the cladding, fuel rods, and fuel assembly hardware before dry storage and consolidation of fuel, and a qualitative determination of the effects of dry storage and fuel consolidation on fission gas release frommore » the spent fuel rods. The performance tests consisted of 6 to 14 runs involving one or two loading, usually three backfill environments (helium, nitrogen, and vacuum backfills), and one or two storage system orientations. The nitrogen and helium backfills were sampled and analyzed to detect leaking spent fuel rods. At the end of each performance test, periodic gas sampling was conducted on each cask. A spent fuel behavior project (i.e., enhanced surveillance, monitoring, and gas sampling activities) was initiated by DOE in 1994 for intact fuel in a CASTOR V/21 cask and for consolidated fuel in a VSC-17 cask. The results of the gas sampling activities are included in this report. Information on spent fuel integrity is of interest in evaluating the impact of long-term dry storage on the behavior of spent fuel rods. Spent fuel used during cask performance tests at INEL offers significant opportunities for confirmation of the benign nature of long-term dry storage. Supporting cask demonstration included licensing and operation of an independent spent fuel storage installation (ISFSI) at the Virginia Power (VP) Surry reactor site. A CASTOR V/21, an MC-10, and a Nuclear Assurance NAC-I28 have been loaded and placed at the VP ISFSI as part of the demonstration program. 13 refs., 14 figs., 9 tabs.« less

Performance of miniaturized direct methanol fuel cell (DMFC) devices using micropump for fuel delivery

NASA Astrophysics Data System (ADS)

Zhang, Tao; Wang, Qing-Ming

A fuel cell is a device that can convert chemical energy into electricity directly. Among various types of fuel cells, both polymer electrolyte membrane fuel cells (PEMFCs) and direct methanol fuel cells (DMFCs) can work at low temperature (<80 °C). Therefore, they can be used to supply power for commercial portable electronics such as laptop computers, digital cameras, PDAs and cell phones. The focus of this paper is to investigate the performance of a miniaturized DMFC device using a micropump to deliver fuel. The core of this micropump is a piezoelectric ring-type bending actuator and the associated nozzle/diffuser for directing fuel flow. Based on the experimental measurements, it is found that the performance of the fuel cell can be significantly improved if enough fuel flow is induced by the micropump at anode. Three factors may contribute to the performance enhancement including replenishment of methanol, decrease of diffusion resistance and removal of carbon dioxide. In comparison with conventional mini pumps, the size of the piezoelectric micropump is much smaller and the energy consumption is much lower. Thus, it is very viable and effective to use a piezoelectric valveless micropump for fuel delivery in miniaturized DMFC power systems.

Effects of Gas Turbine Component Performance on Engine and Rotary Wing Vehicle Size and Performance

NASA Technical Reports Server (NTRS)

Snyder, Christopher A.; Thurman, Douglas R.

2010-01-01

In support of the Fundamental Aeronautics Program, Subsonic Rotary Wing Project, further gas turbine engine studies have been performed to quantify the effects of advanced gas turbine technologies on engine weight and fuel efficiency and the subsequent effects on a civilian rotary wing vehicle size and mission fuel. The Large Civil Tiltrotor (LCTR) vehicle and mission and a previous gas turbine engine study will be discussed as a starting point for this effort. Methodology used to assess effects of different compressor and turbine component performance on engine size, weight and fuel efficiency will be presented. A process to relate engine performance to overall LCTR vehicle size and fuel use will also be given. Technology assumptions and levels of performance used in this analysis for the compressor and turbine components performances will be discussed. Optimum cycles (in terms of power specific fuel consumption) will be determined with subsequent engine weight analysis. The combination of engine weight and specific fuel consumption will be used to estimate their effect on the overall LCTR vehicle size and mission fuel usage. All results will be summarized to help suggest which component performance areas have the most effect on the overall mission.

Impact of future fuel properties on aircraft engines and fuel systems

NASA Technical Reports Server (NTRS)

Rudey, R. A.; Grobman, J. S.

1978-01-01

The effect of modifications in hydrocarbon jet fuels specifications on engine performance, component durability and maintenance, and aircraft fuel system performance is discussed. Specific topics covered include: specific fuel consumption; ignition at relight limits; exhaust emissions; combustor liner temperatures; carbon deposition; gum formation in fuel nozzles, erosion and corrosion of turbine blades and vanes; deposits in fuel system heat exchangers; and pumpability and flowability of the fuel. Data that evaluate the ability of current technology aircraft to accept fuel specification changes are presented, and selected technological advances that can reduce the severity of the problems are described and discussed.

40 CFR 600.208-12 - Calculation of FTP-based and HFET-based fuel economy, CO2 emissions, and carbon-related exhaust...

Code of Federal Regulations, 2014 CFR

2014-07-01

..., and carbon-related exhaust emissions from the tests performed using gasoline or diesel test fuel. (ii... from the tests performed using alcohol or natural gas test fuel. (b) For each model type, as determined... from the tests performed using gasoline or diesel test fuel. (ii) Calculate the city, highway, and...

40 CFR 600.208-12 - Calculation of FTP-based and HFET-based fuel economy, CO2 emissions, and carbon-related exhaust...

Code of Federal Regulations, 2012 CFR

2012-07-01

..., and carbon-related exhaust emissions from the tests performed using gasoline or diesel test fuel. (ii... from the tests performed using alcohol or natural gas test fuel. (b) For each model type, as determined... from the tests performed using gasoline or diesel test fuel. (ii) Calculate the city, highway, and...

40 CFR 600.208-12 - Calculation of FTP-based and HFET-based fuel economy, CO2 emissions, and carbon-related exhaust...

Code of Federal Regulations, 2013 CFR

2013-07-01

..., and carbon-related exhaust emissions from the tests performed using gasoline or diesel test fuel. (ii... from the tests performed using alcohol or natural gas test fuel. (b) For each model type, as determined... from the tests performed using gasoline or diesel test fuel. (ii) Calculate the city, highway, and...

Steady-State Thermal-Hydraulics Analyses for the Conversion of BR2 to Low Enriched Uranium Fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Licht, J.; Bergeron, A.; Dionne, B.

The code PLTEMP/ANL version 4.2 was used to perform the steady-state thermal-hydraulic analyses of the BR2 research reactor for conversion from Highly-Enriched to Low Enriched Uranium fuel (HEU and LEU, respectively). Calculations were performed to evaluate different fuel assemblies with respect to the onset of nucleate boiling (ONB), flow instability (FI), critical heat flux (CHF) and fuel temperature at beginning of cycle conditions. The fuel assemblies were characteristic of fresh fuel (0% burnup), highest heat flux (16% burnup), highest power (32% burnup) and highest burnup (46% burnup). Results show that the high heat flux fuel element is limiting for ONB,more » FI, and CHF, for both HEU and LEU fuel, but that the high power fuel element produces similar margin in a few cases. The maximum fuel temperature similarly occurs in both the high heat flux and high power fuel assemblies for both HEU and LEU fuel. A sensitivity study was also performed to evaluate the variation in fuel temperature due to uncertainties in the thermal conductivity degradation associated with burnup.« less

40 CFR 60.43 - Standard for sulfur dioxide (SO2).

Code of Federal Regulations, 2014 CFR

2014-07-01

... (CONTINUED) STANDARDS OF PERFORMANCE FOR NEW STATIONARY SOURCES Standards of Performance for Fossil-Fuel.../J heat input (0.80 lb/MMBtu) derived from liquid fossil fuel or liquid fossil fuel and wood residue. (2) 520 ng/J heat input (1.2 lb/MMBtu) derived from solid fossil fuel or solid fossil fuel and wood...

40 CFR 60.43 - Standard for sulfur dioxide (SO2).

Code of Federal Regulations, 2012 CFR

2012-07-01

... (CONTINUED) STANDARDS OF PERFORMANCE FOR NEW STATIONARY SOURCES Standards of Performance for Fossil-Fuel.../J heat input (0.80 lb/MMBtu) derived from liquid fossil fuel or liquid fossil fuel and wood residue. (2) 520 ng/J heat input (1.2 lb/MMBtu) derived from solid fossil fuel or solid fossil fuel and wood...

40 CFR 60.43 - Standard for sulfur dioxide (SO2).

Code of Federal Regulations, 2013 CFR

2013-07-01

... (CONTINUED) STANDARDS OF PERFORMANCE FOR NEW STATIONARY SOURCES Standards of Performance for Fossil-Fuel.../J heat input (0.80 lb/MMBtu) derived from liquid fossil fuel or liquid fossil fuel and wood residue. (2) 520 ng/J heat input (1.2 lb/MMBtu) derived from solid fossil fuel or solid fossil fuel and wood...

BISON Fuel Performance Analysis of IFA-796 Rod 3 & 4 and Investigation of the Impact of Fuel Creep

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wirth, Brian; Terrani, Kurt A.; Sweet, Ryan T.

In order to improve the accident tolerance of light water reactor (LWR) fuel, alternative cladding materials have been proposed to replace the currently used zirconium (Zr)-based alloys. Of these materials, there is a particular focus on iron-chromiumaluminum (FeCrAl) alloys because they exhibit slower oxidation kinetics in high-temperature steam than Zr-alloys. This should decrease the energy release due to oxidation and slow cladding consumption in the presence of high temperature steam. These alloys should also exhibit increased “coping time” in the event of an accident scenario by improving the mechanical performance at high temperatures, allowing greater flexibility to achieve core cooling.more » As a continuation of the development of these alloys, in-reactor irradiation testing of FeCrAl cladded fuel rods has started. In order to provide insight on the possible behavior of these fuel rods as they undergo irradiation in the Halden Boiling Water Reactor, engineering analysis has been performed using FeCrAl material models implemented into the BISON fuel performance code. This milestone report provides an update on the ongoing development of modeling capability to predict FeCrAl cladding fuel performance and to provide an early look at the possible behavior of planned in-reactor FeCrAl cladding experiments. In particular, this report consists of two separate analyses. The first analysis consists of fuel performance simulations of IFA-796 rod 4 and two segments of rod 3. These simulations utilize previously implemented material models for the C35M FeCrAl alloy and UO2 to provide a bounding behavior analysis corresponding to variation of the initial fuel cladding gap thickness within the fuel rod. The second analysis is an assessment of the fuel and cladding stress states after modification of the fuel creep model that is currently implemented in the BISON fuel performance code. Effects from modifying the fuel creep model were identified for the BISON simulations of the IFA-796 rod 4 experiment, but show that varying the creep model (within the range investigated here) only provide a minimal increase in the fuel radius and maximum cladding hoop stress. Continued investigation of fuel behavioral models will include benchmarking the modified fuel creep model against available experimental data, as well as an investigation of the role that fuel cracking will play in the compliance of the fuel. Correctly calculating stress evolution in the fuel is key to assessing fuel behavior up to gap closure and the subsequent deformation of the cladding due to PCMI. The inclusion of frictional contact should also be investigated to determine the axial elongation of the fuel rods for comparison with data from this experiment.« less

Effect of aviation fuel type and fuel injection conditions on the spray characteristics of pressure swirl and hybrid air blast fuel injectors

NASA Astrophysics Data System (ADS)

Feddema, Rick

Feddema, Rick T. M.S.M.E., Purdue University, December 2013. Effect of Aviation Fuel Type and Fuel Injection Conditions on the Spray Characteristics of Pressure Swirl and Hybrid Air Blast Fuel Injectors. Major Professor: Dr. Paul E. Sojka, School of Mechanical Engineering Spray performance of pressure swirl and hybrid air blast fuel injectors are central to combustion stability, combustor heat management, and pollutant formation in aviation gas turbine engines. Next generation aviation gas turbine engines will optimize spray atomization characteristics of the fuel injector in order to achieve engine efficiency and emissions requirements. Fuel injector spray atomization performance is affected by the type of fuel injector, fuel liquid properties, fuel injection pressure, fuel injection temperature, and ambient pressure. Performance of pressure swirl atomizer and hybrid air blast nozzle type fuel injectors are compared in this study. Aviation jet fuels, JP-8, Jet A, JP-5, and JP-10 and their effect on fuel injector performance is investigated. Fuel injector set conditions involving fuel injector pressure, fuel temperature and ambient pressure are varied in order to compare each fuel type. One objective of this thesis is to contribute spray patternation measurements to the body of existing drop size data in the literature. Fuel droplet size tends to increase with decreasing fuel injection pressure, decreasing fuel injection temperature and increasing ambient injection pressure. The differences between fuel types at particular set conditions occur due to differences in liquid properties between fuels. Liquid viscosity and surface tension are identified to be fuel-specific properties that affect the drop size of the fuel. An open aspect of current research that this paper addresses is how much the type of aviation jet fuel affects spray atomization characteristics. Conventional aviation fuel specifications are becoming more important with new interest in alternative fuels. Optical patternation data and line of sight laser diffraction data show that there is significant difference between jet fuels. Particularly at low fuel injection pressures (0.345 MPa) and cold temperatures (-40 C), the patternation data shows that the total surface area in the spray at 38.1 mm from the pressure swirl injector for the JP-10 fuel type is one-sixth the amount of the JP-8. Finally, this study compares the atomizer performance of a pressure swirl nozzle to a hybrid air blast nozzle. The total surface area for both the hybrid air blast nozzle and the pressure swirl nozzle show a similar decline in atomization performance at low fuel injection pressures and cold temperatures. However, the optical patternator radial profile data and the line of sight laser diffraction data show that the droplet size and spray distribution data are less affected by injection conditions and fuel type in the hybrid air blast nozzle, than they are in the pressure swirl nozzle. One explanation is that the aerodynamic forces associated with the swirler on the hybrid air blast nozzle control the distribution droplets in the spray. This is in contrast to the pressure swirl nozzle droplet distribution that is controlled by internal geometry and droplet ballistics.

Method of improving fuel cell performance by removing at least one metal oxide contaminant from a fuel cell electrode

DOEpatents

Kim, Yu Seung [Los Alamos, NM; Choi, Jong-Ho [Los Alamos, NM; Zelenay, Piotr [Los Alamos, NM

2009-08-18

A method of removing contaminants from a fuel cell catalyst electrode. The method includes providing a getter electrode and a fuel cell catalyst electrode having at least one contaminant to a bath and applying a voltage sufficient to drive the contaminant from the fuel cell catalyst electrode to the getter electrode. Methods of removing contaminants from a membrane electrode assembly of a fuel cell and of improving performance of a fuel cell are also provided.

Sharing Information among Various Organizations in Relief Efforts

DTIC Science & Technology

2005-09-01

in which two or more social actors work together toward a singular common end that requires the transmutation of materials, ideas, and/or social...women and children to malnutrition and famine, sexual violence, and displacement. The women’s risk of unemployment and loss of income-generating work

Highly durable, coking and sulfur tolerant, fuel-flexible protonic ceramic fuel cells.

PubMed

Duan, Chuancheng; Kee, Robert J; Zhu, Huayang; Karakaya, Canan; Chen, Yachao; Ricote, Sandrine; Jarry, Angelique; Crumlin, Ethan J; Hook, David; Braun, Robert; Sullivan, Neal P; O'Hayre, Ryan

2018-05-01

Protonic ceramic fuel cells, like their higher-temperature solid-oxide fuel cell counterparts, can directly use both hydrogen and hydrocarbon fuels to produce electricity at potentially more than 50 per cent efficiency 1,2 . Most previous direct-hydrocarbon fuel cell research has focused on solid-oxide fuel cells based on oxygen-ion-conducting electrolytes, but carbon deposition (coking) and sulfur poisoning typically occur when such fuel cells are directly operated on hydrocarbon- and/or sulfur-containing fuels, resulting in severe performance degradation over time 3-6 . Despite studies suggesting good performance and anti-coking resistance in hydrocarbon-fuelled protonic ceramic fuel cells 2,7,8 , there have been no systematic studies of long-term durability. Here we present results from long-term testing of protonic ceramic fuel cells using a total of 11 different fuels (hydrogen, methane, domestic natural gas (with and without hydrogen sulfide), propane, n-butane, i-butane, iso-octane, methanol, ethanol and ammonia) at temperatures between 500 and 600 degrees Celsius. Several cells have been tested for over 6,000 hours, and we demonstrate excellent performance and exceptional durability (less than 1.5 per cent degradation per 1,000 hours in most cases) across all fuels without any modifications in the cell composition or architecture. Large fluctuations in temperature are tolerated, and coking is not observed even after thousands of hours of continuous operation. Finally, sulfur, a notorious poison for both low-temperature and high-temperature fuel cells, does not seem to affect the performance of protonic ceramic fuel cells when supplied at levels consistent with commercial fuels. The fuel flexibility and long-term durability demonstrated by the protonic ceramic fuel cell devices highlight the promise of this technology and its potential for commercial application.

Performance analysis of a SOFC under direct internal reforming conditions

NASA Astrophysics Data System (ADS)

Janardhanan, Vinod M.; Heuveline, Vincent; Deutschmann, Olaf

This paper presents the performance analysis of a planar solid-oxide fuel cell (SOFC) under direct internal reforming conditions. A detailed solid-oxide fuel cell model is used to study the influences of various operating parameters on cell performance. Significant differences in efficiency and power density are observed for isothermal and adiabatic operational regimes. The influence of air number, specific catalyst area, anode thickness, steam to carbon (s/c) ratio of the inlet fuel, and extend of pre-reforming on cell performance is analyzed. In all cases except for the case of pre-reformed fuel, adiabatic operation results in lower performance compared to isothermal operation. It is further discussed that, though direct internal reforming may lead to cost reduction and increased efficiency by effective utilization of waste heat, the efficiency of the fuel cell itself is higher for pre-reformed fuel compared to non-reformed fuel. Furthermore, criteria for the choice of optimal operating conditions for cell stacks operating under direct internal reforming conditions are discussed.

Design and Performance of LPG Fuel Mixer for Dual Fuel Diesel Engine

NASA Astrophysics Data System (ADS)

Desrial; Saputro, W.; Garcia, P. P.

2018-05-01

Small horizontal diesel engines are commonly used for agricultural machinery, however, availability of diesel fuel become one of big problems especially in remote area. Conversely, in line with government policy for conversion of kerosene into LPG for cooking, then LPG become more popular and available even in remote area. Therefore, LPG is potential fuel to replace the shortage of diesel fuel for operating diesel engine in remote area. The purpose of this study was to design mixing device for using dual fuel i.e. LPG and diesel fuel and evaluate its performance accordingly. Simulation by using CFD was done in order to analyze mixture characteristics of LPG in air intake manifold. The performance test was done by varying the amount of LPG injected in intake air at 20%, 25%, 30%, 35%, until 40%, respectively. Result of CFD contour simulation showed the best combination when mixing 30% LPG into the intake air. Performance test of this research revealed that mixing LPG in air intake can reduce the diesel fuel consumption about 0.7 l/hour (without load) and 1.14 l/hour (with load). Diesel engine revolution increases almost 300 rpm faster than when using diesel fuel only. Based on economic analysis, using the fuel combination (diesel fuel – LPG) is not recommended in the area near SPBU where the price of diesel fuel is standard. However, using the fuel combination LPG-diesel fuel is highly recommended in the remote areas in Indonesia where price of diesel fuel is comparatively expensive which will provide cheaper total fuel cost for diesel engine operation.

Fully Ceramic Microencapsulated Fuel Development for LWR Applications

DOE Office of Scientific and Technical Information (OSTI.GOV)

Snead, Lance Lewis; Besmann, Theodore M; Terrani, Kurt A

2012-01-01

The concept, fabrication, and key feasibility issues of a new fuel form based on the microencapsulated (TRISO-type) fuel which has been specifically engineered for LWR application and compacted within a SiC matrix will be presented. This fuel, the so-called fully ceramic microencapsulated fuel is currently undergoing development as an accident tolerant fuel for potential UO2 replacement in commercial LWRs. While the ability of this fuel to facilitate normal LWR cycle performance is an ongoing effort within the program, this will not be a focus of this paper. Rather, key feasibility and performance aspects of the fuel will be presented includingmore » the ability to fabricate a LWR-specific TRISO, the need for and route to a high thermal conductivity and fully dense matrix that contains neutron poisons, and the performance of that matrix under irradiation and the interaction of the fuel with commercial zircaloy clad.« less

The COPERNIC3 project: how AREVA is successfully developing an advanced global fuel rod performance code

DOE Office of Scientific and Technical Information (OSTI.GOV)

Garnier, Ch.; Mailhe, P.; Sontheimer, F.

2007-07-01

Fuel performance is a key factor for minimizing operating costs in nuclear plants. One of the important aspects of fuel performance is fuel rod design, based upon reliable tools able to verify the safety of current fuel solutions, prevent potential issues in new core managements and guide the invention of tomorrow's fuels. AREVA is developing its future global fuel rod code COPERNIC3, which is able to calculate the thermal-mechanical behavior of advanced fuel rods in nuclear plants. Some of the best practices to achieve this goal are described, by reviewing the three pillars of a fuel rod code: the database,more » the modelling and the computer and numerical aspects. At first, the COPERNIC3 database content is described, accompanied by the tools developed to effectively exploit the data. Then is given an overview of the main modelling aspects, by emphasizing the thermal, fission gas release and mechanical sub-models. In the last part, numerical solutions are detailed in order to increase the computational performance of the code, with a presentation of software configuration management solutions. (authors)« less

Evaluation of neutron total and capture cross sections on 99Tc in the unresolved resonance region

NASA Astrophysics Data System (ADS)

Iwamoto, Nobuyuki; Katabuchi, Tatsuya

2017-09-01

Long-lived fission product Technetium-99 is one of the most important radioisotopes for nuclear transmutation. The reliable nuclear data are indispensable for a wide energy range up to a few MeV, in order to develop environmental load reducing technology. The statistical analyses of resolved resonances were performed by using the truncated Porter-Thomas distribution, coupled-channels optical model, nuclear level density model and Bayes' theorem on conditional probability. The total and capture cross sections were calculated by a nuclear reaction model code CCONE. The resulting cross sections have statistical consistency between the resolved and unresolved resonance regions. The evaluated capture data reproduce those recently measured at ANNRI of J-PARC/MLF above resolved resonance region up to 800 keV.

Calculations vs. measurements of remnant dose rates for SNS spent structures

NASA Astrophysics Data System (ADS)

Popova, I. I.; Gallmeier, F. X.; Trotter, S.; Dayton, M.

2018-06-01

Residual dose rate measurements were conducted on target vessel #13 and proton beam window #5 after extraction from their service locations. These measurements were used to verify calculation methods of radionuclide inventory assessment that are typically performed for nuclear waste characterization and transportation of these structures. Neutronics analyses for predicting residual dose rates were carried out using the transport code MCNPX and the transmutation code CINDER90. For transport analyses complex and rigorous geometry model of the structures and their surrounding are applied. The neutronics analyses were carried out using Bertini and CEM high energy physics models for simulating particles interaction. Obtained preliminary calculational results were analysed and compared to the measured dose rates and overall are showing good agreement with in 40% in average.

Calculations vs. measurements of remnant dose rates for SNS spent structures

DOE Office of Scientific and Technical Information (OSTI.GOV)

Popova, Irina I.; Gallmeier, Franz X.; Trotter, Steven M.

Residual dose rate measurements were conducted on target vessel #13 and proton beam window #5 after extraction from their service locations. These measurements were used to verify calculation methods of radionuclide inventory assessment that are typically performed for nuclear waste characterization and transportation of these structures. Neutronics analyses for predicting residual dose rates were carried out using the transport code MCNPX and the transmutation code CINDER90. For transport analyses complex and rigorous geometry model of the structures and their surrounding are applied. The neutronics analyses were carried out using Bertini and CEM high energy physics models for simulating particles interaction.more » Obtained preliminary calculational results were analysed and compared to the measured dose rates and overall are showing good agreement with in 40% in average.« less

Optimisation of air cooled, open-cathode fuel cells: Current of lowest resistance and electro-thermal performance mapping

NASA Astrophysics Data System (ADS)

Meyer, Quentin; Ronaszegi, Krisztian; Pei-June, Gan; Curnick, Oliver; Ashton, Sean; Reisch, Tobias; Adcock, Paul; Shearing, Paul R.; Brett, Daniel J. L.

2015-09-01

Selecting the ideal operating point for a fuel cell depends on the application and consequent trade-off between efficiency, power density and various operating considerations. A systematic methodology for determining the optimal operating point for fuel cells is lacking; there is also the need for a single-value metric to describe and compare fuel cell performance. This work shows how the 'current of lowest resistance' can be accurately measured using electrochemical impedance spectroscopy and used as a useful metric of fuel cell performance. This, along with other measures, is then used to generate an 'electro-thermal performance map' of fuel cell operation. A commercial air-cooled open-cathode fuel cell is used to demonstrate how the approach can be used; in this case leading to the identification of the optimum operating temperature of ∼45 °C.

40 CFR 600.209-08 - Calculation of vehicle-specific 5-cycle fuel economy values for a model type.

Code of Federal Regulations, 2013 CFR

2013-07-01

... intended for sale at high altitude, the Administrator may use fuel economy data from tests conducted on... from the tests performed using gasoline or diesel test fuel. (ii) If 5-cycle testing was performed on the alcohol or natural gas test fuel, calculate the city and highway fuel economy values from the...

40 CFR 63.11213 - What fuel analyses and procedures must I use for the performance tests?

Code of Federal Regulations, 2013 CFR

2013-07-01

... What fuel analyses and procedures must I use for the performance tests? (a) You must conduct fuel... approximately equal intervals during a test run period. (c) Determine the concentration of mercury in the fuel... 40 Protection of Environment 15 2013-07-01 2013-07-01 false What fuel analyses and procedures must...

40 CFR 600.209-08 - Calculation of vehicle-specific 5-cycle fuel economy values for a model type.

Code of Federal Regulations, 2012 CFR

2012-07-01

... intended for sale at high altitude, the Administrator may use fuel economy data from tests conducted on... from the tests performed using gasoline or diesel test fuel. (ii) If 5-cycle testing was performed on the alcohol or natural gas test fuel, calculate the city and highway fuel economy values from the...

40 CFR 600.206-12 - Calculation and use of FTP-based and HFET-based fuel economy and carbon-related exhaust emission...

Code of Federal Regulations, 2011 CFR

2011-07-01

... exhaust emission values from the tests performed using gasoline or diesel test fuel. (ii) Calculate the city, highway, and combined fuel economy and carbon-related exhaust emission values from the tests performed using alcohol or natural gas test fuel. (b) If only one equivalent petroleum-based fuel economy...

40 CFR 600.207-08 - Calculation and use of vehicle-specific 5-cycle-based fuel economy values for vehicle...

Code of Federal Regulations, 2012 CFR

2012-07-01

... economy values from the tests performed using gasoline or diesel test fuel. (ii)(A) Calculate the 5-cycle city and highway fuel economy values from the tests performed using alcohol or natural gas test fuel...-specific 5-cycle-based fuel economy values for vehicle configurations. 600.207-08 Section 600.207-08...

40 CFR 63.11213 - What fuel analyses and procedures must I use for the performance tests?

Code of Federal Regulations, 2011 CFR

2011-07-01

... What fuel analyses and procedures must I use for the performance tests? (a) You must conduct fuel... approximately equal intervals during a test run period. (c) Determine the concentration of mercury in the fuel... 40 Protection of Environment 14 2011-07-01 2011-07-01 false What fuel analyses and procedures must...

40 CFR 63.11213 - What fuel analyses and procedures must I use for the performance tests?

Code of Federal Regulations, 2014 CFR

2014-07-01

... What fuel analyses and procedures must I use for the performance tests? (a) You must conduct fuel... approximately equal intervals during a test run period. (c) Determine the concentration of mercury in the fuel... 40 Protection of Environment 15 2014-07-01 2014-07-01 false What fuel analyses and procedures must...

40 CFR 600.208-08 - Calculation of FTP-based and HFET-based fuel economy values for a model type.

Code of Federal Regulations, 2013 CFR

2013-07-01

... tests conducted on these vehicle configuration(s) at high altitude to calculate the fuel economy for the... combined fuel economy values from the tests performed using gasoline or diesel test fuel. (ii) Calculate the city, highway, and combined fuel economy values from the tests performed using alcohol or natural...

40 CFR 600.208-08 - Calculation of FTP-based and HFET-based fuel economy values for a model type.

Code of Federal Regulations, 2012 CFR

2012-07-01

... tests conducted on these vehicle configuration(s) at high altitude to calculate the fuel economy for the... combined fuel economy values from the tests performed using gasoline or diesel test fuel. (ii) Calculate the city, highway, and combined fuel economy values from the tests performed using alcohol or natural...

40 CFR 600.207-08 - Calculation and use of vehicle-specific 5-cycle-based fuel economy values for vehicle...

Code of Federal Regulations, 2013 CFR

2013-07-01

... economy values from the tests performed using gasoline or diesel test fuel. (ii)(A) Calculate the 5-cycle city and highway fuel economy values from the tests performed using alcohol or natural gas test fuel...-specific 5-cycle-based fuel economy values for vehicle configurations. 600.207-08 Section 600.207-08...

40 CFR 63.11213 - What fuel analyses and procedures must I use for the performance tests?

Code of Federal Regulations, 2012 CFR

2012-07-01

... What fuel analyses and procedures must I use for the performance tests? (a) You must conduct fuel... approximately equal intervals during a test run period. (c) Determine the concentration of mercury in the fuel... 40 Protection of Environment 15 2012-07-01 2012-07-01 false What fuel analyses and procedures must...

2007 international meeting on Reduced Enrichment for Research and Test Reactors (RERTR). Abstracts and available papers presented at the meeting

DOE Office of Scientific and Technical Information (OSTI.GOV)

NONE

2008-07-15

The Meeting papers discuss research and test reactor fuel performance, manufacturing and testing. Some of the main topics are: conversion from HEU to LEU in different reactors and corresponding problems and activities; flux performance and core lifetime analysis with HEU and LEU fuels; physics and safety characteristics; measurement of gamma field parameters in core with LEU fuel; nondestructive analysis of RERTR fuel; thermal hydraulic analysis; fuel interactions; transient analyses and thermal hydraulics for HEU and LEU cores; microstructure research reactor fuels; post irradiation analysis and performance; computer codes and other related problems.

Assessment and comparison of 100-MW coal gasification phosphoric acid fuel cell power plants

NASA Technical Reports Server (NTRS)

Lu, Cheng-Yi

1988-01-01

One of the advantages of fuel cell (FC) power plants is fuel versatility. With changes only in the fuel processor, the power plant will be able to accept a variety of fuels. This study was performed to design process diagrams, evaluate performance, and to estimate cost of 100 MW coal gasifier (CG)/phosphoric acid fuel cell (PAFC) power plant systems utilizing coal, which is the largest single potential source of alternate hydrocarbon liquids and gases in the United States, as the fuel. Results of this study will identify the most promising integrated CG/PAFC design and its near-optimal operating conditions. The comparison is based on the performance and cost of electricity which is calculated under consistent financial assumptions.

Redwing: A MOOSE application for coupling MPACT and BISON

DOE Office of Scientific and Technical Information (OSTI.GOV)

Frederick N. Gleicher; Michael Rose; Tom Downar

Fuel performance and whole core neutron transport programs are often used to analyze fuel behavior as it is depleted in a reactor. For fuel performance programs, internal models provide the local intra-pin power density, fast neutron flux, burnup, and fission rate density, which are needed for a fuel performance analysis. The fuel performance internal models have a number of limitations. These include effects on the intra-pin power distribution by nearby assembly elements, such as water channels and control rods, and the further limitation of applicability to a specified fuel type such as low enriched UO2. In addition, whole core neutronmore » transport codes need an accurate intra-pin temperature distribution in order to calculate neutron cross sections. Fuel performance simulations are able to model the intra-pin fuel displacement as the fuel expands and densifies. These displacements must be accurately modeled in order to capture the eventual mechanical contact of the fuel and the clad; the correct radial gap width is needed for an accurate calculation of the temperature distribution of the fuel rod. Redwing is a MOOSE-based application that enables coupling between MPACT and BISON for transport and fuel performance coupling. MPACT is a 3D neutron transport and reactor core simulator based on the method of characteristics (MOC). The development of MPACT began at the University of Michigan (UM) and now is under the joint development of ORNL and UM as part of the DOE CASL Simulation Hub. MPACT is able to model the effects of local assembly elements and is able calculate intra-pin quantities such as the local power density on a volumetric mesh for any fuel type. BISON is a fuel performance application of Multi-physics Object Oriented Simulation Environment (MOOSE), which is under development at Idaho National Laboratory. BISON is able to solve the nonlinearly coupled mechanical deformation and heat transfer finite element equations that model a fuel element as it is depleted in a nuclear reactor. Redwing couples BISON and MPACT in a single application. Redwing maps and transfers the individual intra-pin quantities such as fission rate density, power density, and fast neutron flux from the MPACT volumetric mesh to the individual BISON finite element meshes. For a two-way coupling Redwing maps and transfers the individual pin temperature field and axially dependent coolant densities from the BISON mesh to the MPACT volumetric mesh. Details of the mapping are given. Redwing advances the simulation with the MPACT solution for each depletion time step and then advances the multiple BISON simulations for fuel performance calculations. Sub-cycle advancement can be applied to the individual BISON simulations and allows multiple time steps to be applied to the fuel performance simulations. Currently, only loose coupling where data from a previous time step is applied to the current time step is performed.« less

High performance direct methanol fuel cell with thin electrolyte membrane

NASA Astrophysics Data System (ADS)

Wan, Nianfang

2017-06-01

A high performance direct methanol fuel cell is achieved with thin electrolyte membrane. 320 mW cm-2 of peak power density and over 260 mW cm-2 at 0.4 V are obtained when working at 90 °C with normal pressure air supply. It is revealed that the increased anode half-cell performance with temperature contributes primarily to the enhanced performance at elevated temperature. From the comparison of iR-compensated cathode potential of methanol/air with that of H2/air fuel cell, the impact of methanol crossover on cathode performance decreases with current density and becomes negligible at high current density. Current density is found to influence fuel efficiency and methanol crossover significantly from the measurement of fuel efficiency at different current density. At high current density, high fuel efficiency can be achieved even at high temperature, indicating decreased methanol crossover.

76 FR 3517 - Standards of Performance for Fossil-Fuel-Fired, Electric Utility, Industrial-Commercial...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-01-20

... Standards of Performance for Fossil-Fuel-Fired, Electric Utility, Industrial-Commercial-Institutional, and... following: Category NAICS \\1\\ Examples of regulated entities Industry 221112 Fossil fuel-fired electric utility steam generating units. Federal Government 22112 Fossil fuel-fired electric utility steam...

Comparison of Performance of AN-F-58 Fuel and Gasoline in J34-WE-22 Turbojet Engine

NASA Technical Reports Server (NTRS)

Dowman, Harry W; Younger, George G

1949-01-01

As part of an investigation of the performance of AN-F-58 fuel in various types of turbojet engine, the performance of this fuel in a 3000-pound-thrust turbojet engine has been investigated in an altitude test chamber together with the comparative performance of 62-octane gasoline. The investigation of normal engine performance, which covered a range of engine speeds at altitudes from 5000 to 50,000 feet and flight Mach numbers up to 1.00, showed that both the net thrust and average turbine-outlet temperatures were approximately the same for both fuels. The specific fuel consumption and the combustion efficiency at the maximum engine speeds investigated were approximately the same for both fuels at altitudes up to 35,000 feet, but at an altitude of 50,000 feet the specific fuel consumption was about 9 percent higher and the combustion efficiency was correspondingly lower with the AN-F-58 fuel than with gasoline. The low-engine-speed blow-out limits were about the same for both fuels. Ignition of AN-F-58 fuel with the standard spark plug was possible only with the spark plug in a clean condition; ignition was impossible at all flight conditions investigated when the plug was fouled by an accumulation of liquid fuel from a preceding false start. Use of an extended-electrode spark plug provided satisfactory ignition over a slightly smaller range of altitudes and flight Mach numbers than for gasoline with the standard spark plug.

Recent advances on Zeolite modification for direct alcohol fuel cells (DAFCs)

NASA Astrophysics Data System (ADS)

Makertihartha, I. G. B. N.; Zunita, M.; Rizki, Z.; Dharmawijaya, P. T.

2017-03-01

The increase of energy demand and global warming issues has driven studies of alternative energy sources. The polymer electrolyte membrane fuel cell (PEMFC) can be an alternative energy source by (partially) replacing the use of fossil fuel which is in line with the green technology concept. However, the usage of hydrogen as a fuel has several disadvantages mainly transportation and storage related to its safety aspects. Recently, alcohol has gained attention as an energy source for fuel cell application, namely direct alcohol fuel cell (DAFC). Among alcohols, high-mass energy density methanol and ethanol are widely used as direct methanol fuel cell (DMFC) and direct ethanol fuel cell (DEFC), respectively. Currently, the performance of DMFC is still rudimentary. Furthermore, the use of ethanol gives some additional privileges such as non-toxic property, renewable, ease of production in great quantity by the fermentation of sugar-containing raw materials. Direct alcohol fuel cell (DAFC) still has weakness in the low proton conductivity and high alcohol crossover. Therefore, to increase the performance of DAFC, modification using zeolite has been performed to improve proton conductivity and decrease alcohol crossover. Zeolite also has high thermal resistance properties, thereby increasing DAFC performance. This paper will discuss briefly about modification of catalyst and membrane for DAFC using zeolite. Zeolite modification effect on fuel cell performance especially proton conductivity and alcohol crossover will be presented in detail.

Modelling of diesel engine fuelled with biodiesel using engine simulation software

NASA Astrophysics Data System (ADS)

Said, Mohd Farid Muhamad; Said, Mazlan; Aziz, Azhar Abdul

2012-06-01

This paper is about modelling of a diesel engine that operates using biodiesel fuels. The model is used to simulate or predict the performance and combustion of the engine by simplified the geometry of engine component in the software. The model is produced using one-dimensional (1D) engine simulation software called GT-Power. The fuel properties library in the software is expanded to include palm oil based biodiesel fuels. Experimental works are performed to investigate the effect of biodiesel fuels on the heat release profiles and the engine performance curves. The model is validated with experimental data and good agreement is observed. The simulation results show that combustion characteristics and engine performances differ when biodiesel fuels are used instead of no. 2 diesel fuel.

Fuel Performance Calculations for FeCrAl Cladding in BWRs

DOE Office of Scientific and Technical Information (OSTI.GOV)

George, Nathan; Sweet, Ryan; Maldonado, G. Ivan

2015-01-01

This study expands upon previous neutronics analyses of the reactivity impact of alternate cladding concepts in boiling water reactor (BWR) cores and directs focus toward contrasting fuel performance characteristics of FeCrAl cladding against those of traditional Zircaloy. Using neutronics results from a modern version of the 3D nodal simulator NESTLE, linear power histories were generated and supplied to the BISON-CASL code for fuel performance evaluations. BISON-CASL (formerly Peregrine) expands on material libraries implemented in the BISON fuel performance code and the MOOSE framework by providing proprietary material data. By creating material libraries for Zircaloy and FeCrAl cladding, the thermomechanical behaviormore » of the fuel rod (e.g., strains, centerline fuel temperature, and time to gap closure) were investigated and contrasted.« less

Multidimensional Fuel Performance Code: BISON

DOE Office of Scientific and Technical Information (OSTI.GOV)

BISON is a finite element based nuclear fuel performance code applicable to a variety of fuel forms including light water reactor fuel rods, TRISO fuel particles, and metallic rod and plate fuel (Refs. [a, b, c]). It solves the fully-coupled equations of thermomechanics and species diffusion and includes important fuel physics such as fission gas release and material property degradation with burnup. BISON is based on the MOOSE framework (Ref. [d]) and can therefore efficiently solve problems on 1-, 2- or 3-D meshes using standard workstations or large high performance computers. BISON is also coupled to a MOOSE-based mesoscale phasemore » field material property simulation capability (Refs. [e, f]). As described here, BISON includes the code library named FOX, which was developed concurrent with BISON. FOX contains material and behavioral models that are specific to oxide fuels.« less

40 CFR 600.207-12 - Calculation and use of vehicle-specific 5-cycle-based fuel economy and CO2 emission values for...

Code of Federal Regulations, 2012 CFR

2012-07-01

... economy and CO2 emission values from the tests performed using gasoline or diesel test fuel. (ii) Calculate the 5-cycle city and highway fuel economy and CO2 emission values from the tests performed using alcohol or natural gas test fuel, if 5-cycle testing has been performed. Otherwise, the procedure in § 600...

40 CFR 600.207-12 - Calculation and use of vehicle-specific 5-cycle-based fuel economy and CO2 emission values for...

Code of Federal Regulations, 2014 CFR

2014-07-01

... economy and CO2 emission values from the tests performed using gasoline or diesel test fuel. (ii) Calculate the 5-cycle city and highway fuel economy and CO2 emission values from the tests performed using alcohol or natural gas test fuel, if 5-cycle testing has been performed. Otherwise, the procedure in § 600...

40 CFR 600.207-12 - Calculation and use of vehicle-specific 5-cycle-based fuel economy and CO2 emission values for...

Code of Federal Regulations, 2013 CFR

2013-07-01

... economy and CO2 emission values from the tests performed using gasoline or diesel test fuel. (ii) Calculate the 5-cycle city and highway fuel economy and CO2 emission values from the tests performed using alcohol or natural gas test fuel, if 5-cycle testing has been performed. Otherwise, the procedure in § 600...

Sustainable Thorium Nuclear Fuel Cycles: A Comparison of Intermediate and Fast Neutron Spectrum Systems

DOE PAGES

Brown, Nicholas R.; Powers, Jeffrey J.; Feng, B.; ...

2015-05-21

This paper presents analyses of possible reactor representations of a nuclear fuel cycle with continuous recycling of thorium and produced uranium (mostly U-233) with thorium-only feed. The analysis was performed in the context of a U.S. Department of Energy effort to develop a compendium of informative nuclear fuel cycle performance data. The objective of this paper is to determine whether intermediate spectrum systems, having a majority of fission events occurring with incident neutron energies between 1 eV and 10 5 eV, perform as well as fast spectrum systems in this fuel cycle. The intermediate spectrum options analyzed include tight latticemore » heavy or light water-cooled reactors, continuously refueled molten salt reactors, and a sodium-cooled reactor with hydride fuel. All options were modeled in reactor physics codes to calculate their lattice physics, spectrum characteristics, and fuel compositions over time. Based on these results, detailed metrics were calculated to compare the fuel cycle performance. These metrics include waste management and resource utilization, and are binned to accommodate uncertainties. The performance of the intermediate systems for this selfsustaining thorium fuel cycle was similar to a representative fast spectrum system. However, the number of fission neutrons emitted per neutron absorbed limits performance in intermediate spectrum systems.« less

Joint Force Quarterly. Number 6, Autumn/Winter 1994-95

DTIC Science & Technology

1994-11-01

enhances the per- formance of the other. Indeed, the strategic challenge often is to find ways to transmute success in one environment into good enough...President Clinton “with contemptuous ease” on issues such as revising military policy toward homo- sexuals and using force in ex-Yugoslavia. Kohn accuses

Guidelines for Developing Habitat-Based Threatened and Endangered Species Population Goals on Army Installations

DTIC Science & Technology

2005-10-01

Many plants have both sexual and vegetative reproduction, and this may be best handled with separate classes that reflect meaningful differ- ences in... Transmutation and functional representation of heterogeneous landscapes.” Landscape Ecology 5: 239-253. LaGro, J.Jr. 1991. “Assessing patch

Transmuting Common Substances: The Cold Fusion Controversy and the Rhetoric of Science.

ERIC Educational Resources Information Center

Thacker, Brad; Stratman, James F.

1995-01-01

Explores the relationship among forensic, deliberative, and epideictic modes of rhetoric in the cold fusion controversy. Shows the interactions between three modes of rhetoric. Examines the ways in which the modes have shaped the emerging scientific consensus. Supports Robert Sanders' contention that rhetorical practices interact with scientific…

Development of advanced technological systems for accelerator transmutation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Batskikh, G.I.; Bondarev, B.I.; Durkin, A.P.

1995-10-01

A development concept of the accelerator nuclear energy reactors is considered for energy generation and nuclear power plant waste conversion into short-lived nuclides along with the requirements imposed on the technological systems necessary for implementation of such projects. The state of art in the field is discussed.

Solid state lift for micrometering in a fuel injector

DOEpatents

Milam, David M.; Carroll, Thomas S.; Lee, Chien-Chang; Miller, Charles R.

2002-01-01

A fuel injector performs main fuel injection by raising fuel pressure in a nozzle chamber to lift a check valve member to a fully open position, and performs preinjection or microinjection by operating a solid state motor to lift the check valve member a much smaller distance.

40 CFR 60.41 - Definitions.

Code of Federal Regulations, 2014 CFR

2014-07-01

.... Fossil fuel means natural gas, petroleum, coal, and any form of solid, liquid, or gaseous fuel derived from such materials for the purpose of creating useful heat. Fossil fuel and wood residue-fired steam... PERFORMANCE FOR NEW STATIONARY SOURCES Standards of Performance for Fossil-Fuel-Fired Steam Generators § 60.41...

40 CFR 60.41 - Definitions.

Code of Federal Regulations, 2013 CFR

2013-07-01

.... Fossil fuel means natural gas, petroleum, coal, and any form of solid, liquid, or gaseous fuel derived from such materials for the purpose of creating useful heat. Fossil fuel and wood residue-fired steam... PERFORMANCE FOR NEW STATIONARY SOURCES Standards of Performance for Fossil-Fuel-Fired Steam Generators § 60.41...

40 CFR 60.41 - Definitions.

Code of Federal Regulations, 2012 CFR

2012-07-01

.... Fossil fuel means natural gas, petroleum, coal, and any form of solid, liquid, or gaseous fuel derived from such materials for the purpose of creating useful heat. Fossil fuel and wood residue-fired steam... PERFORMANCE FOR NEW STATIONARY SOURCES Standards of Performance for Fossil-Fuel-Fired Steam Generators § 60.41...

Screening of advanced cladding materials and UN-U3Si5 fuel

NASA Astrophysics Data System (ADS)

Brown, Nicholas R.; Todosow, Michael; Cuadra, Arantxa

2015-07-01

In the aftermath of Fukushima, a focus of the DOE-NE Advanced Fuels Campaign has been the development of advanced nuclear fuel and cladding options with the potential for improved performance in an accident. Uranium dioxide (UO2) fuels with various advanced cladding materials were analyzed to provide a reference for cladding performance impacts. For advanced cladding options with UO2 fuel, most of the cladding materials have some reactivity and discharge burn-up penalty (in GWd/t). Silicon carbide is one exception in that the reactor physics performance is predicted to be very similar to zirconium alloy cladding. Most candidate claddings performed similar to UO2-Zr fuel-cladding in terms of safety coefficients. The clear exception is that Mo-based materials were identified as potentially challenging from a reactor physics perspective due to high resonance absorption. This paper also includes evaluation of UN-U3Si5 fuels with Kanthal AF or APMT cladding. The objective of the U3Si5 phase in the UN-U3Si5 fuel concept is to shield the nitride phase from water. It was shown that UN-U3Si5 fuels with Kanthal AF or APMT cladding have similar reactor physics and fuel management performance over a wide parameter space of phase fractions when compared to UO2-Zr fuel-cladding. There will be a marginal penalty in discharge burn-up (in GWd/t) and the sensitivity to 14N content in UN ceramic composites is high. Analysis of the rim effect due to self-shielding in the fuel shows that the UN-based ceramic fuels are not expected to have significantly different relative burn-up distributions at discharge relative to the UO2 reference fuel. However, the overall harder spectrum in the UN ceramic composite fuels increases transuranic build-up, which will increase long-term activity in a once-thru fuel cycle but is expected to be a significant advantage in a fuel cycle with continuous recycling of transuranic material. It is recognized that the fuel and cladding properties assumed in these assessments are preliminary, and that additional data are necessary for these materials, most significantly under irradiation.

40 CFR 600.206-12 - Calculation and use of FTP-based and HFET-based fuel economy, CO2 emissions, and carbon-related...

Code of Federal Regulations, 2013 CFR

2013-07-01

... values from the tests performed using gasoline or diesel test fuel. (ii) Calculate the city, highway, and combined fuel economy, CO2 emissions, and carbon-related exhaust emission values from the tests performed using alcohol or natural gas test fuel. (b) If only one equivalent petroleum-based fuel economy value...

40 CFR 600.206-12 - Calculation and use of FTP-based and HFET-based fuel economy, CO2 emissions, and carbon-related...

Code of Federal Regulations, 2014 CFR

2014-07-01

... values from the tests performed using gasoline or diesel test fuel. (ii) Calculate the city, highway, and combined fuel economy, CO2 emissions, and carbon-related exhaust emission values from the tests performed using alcohol or natural gas test fuel. (b) If only one equivalent petroleum-based fuel economy value...

40 CFR 600.206-12 - Calculation and use of FTP-based and HFET-based fuel economy, CO2 emissions, and carbon-related...

Code of Federal Regulations, 2012 CFR

2012-07-01

... values from the tests performed using gasoline or diesel test fuel. (ii) Calculate the city, highway, and combined fuel economy, CO2 emissions, and carbon-related exhaust emission values from the tests performed using alcohol or natural gas test fuel. (b) If only one equivalent petroleum-based fuel economy value...

NREL Fuels and Engines R&D Revs Up Vehicle Efficiency, Performance

DOE Office of Scientific and Technical Information (OSTI.GOV)

None

NREL bridges fuels and engines R&D to maximize vehicle efficiency and performance. The lab’s fuels and engines research covers the full spectrum of innovation—from fuel chemistry, conversion, and combustion to the evaluation of how fuels interact with engine and vehicle design. This innovative approach has the potential to positively impact our economy, national energy security, and air quality.

Used Nuclear Fuel Loading and Structural Performance Under Normal Conditions of Transport- Demonstration of Approach and Results on Used Fuel Performance Characterization

DOE Office of Scientific and Technical Information (OSTI.GOV)

Adkins, Harold; Geelhood, Ken; Koeppel, Brian

2013-09-30

This document addresses Oak Ridge National Laboratory milestone M2FT-13OR0822015 Demonstration of Approach and Results on Used Nuclear Fuel Performance Characterization. This report provides results of the initial demonstration of the modeling capability developed to perform preliminary deterministic evaluations of moderate-to-high burnup used nuclear fuel (UNF) mechanical performance under normal conditions of storage (NCS) and normal conditions of transport (NCT) conditions. This report also provides results from the sensitivity studies that have been performed. Finally, discussion on the long-term goals and objectives of this initiative are provided.

Low-temperature fuel cell systems for commercial airplane auxiliary power.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Curgus, Dita Brigitte; Pratt, Joseph William; Akhil, Abbas Ali

2010-11-01

This presentation briefly describes the ongoing study of fuel cell systems on-board a commercial airplane. Sandia's current project is focused on Proton Exchange Membrane (PEM) fuel cells applied to specific on-board electrical power needs. They are trying to understand how having a fuel cell on an airplane would affect overall performance. The fuel required to accomplish a mission is used to quantify the performance. Our analysis shows the differences between the base airplane and the airplane with the fuel cell. There are many ways of designing a system, depending on what you do with the waste heat. A system thatmore » requires ram air cooling has a large mass penalty due to increased drag. The bottom-line impact can be expressed as additional fuel required to complete the mission. Early results suggest PEM fuel cells can be used on airplanes with manageable performance impact if heat is rejected properly. For PEMs on aircraft, we are continuing to perform: (1) thermodynamic analysis (investigate configurations); (2) integrated electrical design (with dynamic modeling of the micro grid); (3) hardware assessment (performance, weight, and volume); and (4) galley and peaker application.« less

40 CFR 63.11220 - When must I conduct subsequent performance tests or fuel analyses?

Code of Federal Regulations, 2014 CFR

2014-07-01

... performance tests or fuel analyses? 63.11220 Section 63.11220 Protection of Environment ENVIRONMENTAL... When must I conduct subsequent performance tests or fuel analyses? (a) If your boiler has a heat input... performance (stack) tests according to § 63.11212 on a triennial basis, except as specified in paragraphs (b...

40 CFR 63.11220 - When must I conduct subsequent performance tests or fuel analyses?

Code of Federal Regulations, 2013 CFR

2013-07-01

... performance tests or fuel analyses? 63.11220 Section 63.11220 Protection of Environment ENVIRONMENTAL... When must I conduct subsequent performance tests or fuel analyses? (a) If your boiler has a heat input... performance (stack) tests according to § 63.11212 on a triennial basis, except as specified in paragraphs (b...

Review of cell performance in anion exchange membrane fuel cells

NASA Astrophysics Data System (ADS)

Dekel, Dario R.

2018-01-01

Anion exchange membrane fuel cells (AEMFCs) have recently received increasing attention since in principle they allow for the use of non-precious metal catalysts, which dramatically reduces the cost per kilowatt of power in fuel cell devices. Until not long ago, the main barrier in the development of AEMFCs was the availability of highly conductive anion exchange membranes (AEMs); however, improvements on this front in the past decade show that newly developed AEMs have already reached high levels of conductivity, leading to satisfactory cell performance. In recent years, a growing number of research studies have reported AEMFC performance results. In the last three years, new records in performance were achieved. Most of the literature reporting cell performance is based on hydrogen-AEMFCs, although an increasing number of studies have also reported the use of fuels others than hydrogen - such as alcohols, non-alcohol C-based fuels, as well as N-based fuels. This article reviews the cell performance and performance stability achieved in AEMFCs through the years since the first reports in the early 2000s.

Impact of hydrogen SAE J2601 fueling methods on fueling time of light-duty fuel cell electric vehicles

DOE PAGES

Reddi, Krishna; Elgowainy, Amgad; Rustagi, Neha; ...

2017-05-16

Hydrogen fuel cell electric vehicles (HFCEVs) are zero-emission vehicles (ZEVs) that can provide drivers a similar experience to conventional internal combustion engine vehicles (ICEVs), in terms of fueling time and performance (i.e. power and driving range). The Society of Automotive Engineers (SAE) developed fueling protocol J2601 for light-duty HFCEVs to ensure safe vehicle fills while maximizing fueling performance. This study employs a physical model that simulates and compares the fueling performance of two fueling methods, known as the “lookup table” method and the “MC formula” method, within the SAE J2601 protocol. Both the fueling methods provide fast fueling of HFCEVsmore » within minutes, but the MC formula method takes advantage of active measurement of precooling temperature to dynamically control the fueling process, and thereby provides faster vehicle fills. Here, the MC formula method greatly reduces fueling time compared to the lookup table method at higher ambient temperatures, as well as when the precooling temperature falls on the colder side of the expected temperature window for all station types. Although the SAE J2601 lookup table method is the currently implemented standard for refueling hydrogen fuel cell vehicles, the MC formula method provides significant fueling time advantages in certain conditions; these warrant its implementation in future hydrogen refueling stations for better customer satisfaction with fueling experience of HFCEVs.« less

Impact of hydrogen SAE J2601 fueling methods on fueling time of light-duty fuel cell electric vehicles

DOE Office of Scientific and Technical Information (OSTI.GOV)

Reddi, Krishna; Elgowainy, Amgad; Rustagi, Neha

Hydrogen fuel cell electric vehicles (HFCEVs) are zero-emission vehicles (ZEVs) that can provide drivers a similar experience to conventional internal combustion engine vehicles (ICEVs), in terms of fueling time and performance (i.e. power and driving range). The Society of Automotive Engineers (SAE) developed fueling protocol J2601 for light-duty HFCEVs to ensure safe vehicle fills while maximizing fueling performance. This study employs a physical model that simulates and compares the fueling performance of two fueling methods, known as the “lookup table” method and the “MC formula” method, within the SAE J2601 protocol. Both the fueling methods provide fast fueling of HFCEVsmore » within minutes, but the MC formula method takes advantage of active measurement of precooling temperature to dynamically control the fueling process, and thereby provides faster vehicle fills. Here, the MC formula method greatly reduces fueling time compared to the lookup table method at higher ambient temperatures, as well as when the precooling temperature falls on the colder side of the expected temperature window for all station types. Although the SAE J2601 lookup table method is the currently implemented standard for refueling hydrogen fuel cell vehicles, the MC formula method provides significant fueling time advantages in certain conditions; these warrant its implementation in future hydrogen refueling stations for better customer satisfaction with fueling experience of HFCEVs.« less

FY16 Status Report for the Uranium-Molybdenum Fuel Concept

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bennett, Wendy D.; Doherty, Ann L.; Henager, Charles H.

2016-09-22

The Fuel Cycle Research and Development program of the Office of Nuclear Energy has implemented a program to develop a Uranium-Molybdenum metal fuel for light water reactors. Uranium-Molybdenum fuel has the potential to provide superior performance based on its thermo-physical properties. With sufficient development, it may be able to provide the Light Water Reactor industry with a melt-resistant, accident-tolerant fuel with improved safety response. The Pacific Northwest National Laboratory has been tasked with extrusion development and performing ex-reactor corrosion testing to characterize the performance of Uranium-Molybdenum fuel in both these areas. This report documents the results of the fiscal yearmore » 2016 effort to develop the Uranium-Molybdenum metal fuel concept for light water reactors.« less

Performance of a diesel engine operating on raw coal-diesel fuel and solvent refined coal-diesel fuel slurries. Final report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Marshall, H.P.

1980-03-01

Performance tests using an 11 kW single cylinder diesel engine were made to determine the effects of three different micronized coal-fuel oil slurries being considered as alternative fuels. Slurries containing 20, 32, and 40%-wt micronized raw coal in No. 2 fuel oil were used. Results are presented indicating the changes in the concentrations of SO/sub X/ and NO/sub X/ in the exhaust, exhaust opacity, power and efficiency, and in wear rates relative to operation on fuel oil No. 2. The engine was operated for 10 h at full load and 1400 rpm on al fuels except the 40%-wt slurry. Thismore » test was discontinued because of extremely poor performance.« less

Performance of a diesel engine operating on raw coal-diesel fuel and solvent refined coal-diesel fuel slurries. Final report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Marshall, H.P.

1980-03-01

Performance tests using an 11 kw single cylinder diesel engine were made to determine the effects of three different micronized coal-fuel oil slurries being considered as alternative fuels. Slurries containing 20, 32, and 40 percent by weight micronized raw coal in No. 2 fuel oil were used. Results are presented indicating the changes in the concentrations of SO/sub X/ and NO/sub X/ in the exhaust, exhaust opacity, power and efficiency, and in wear rates relative to operation on fuel oil No. 2. The engine was operated for 10 hrs at full load and 1400 rpm on all fuels except themore » 40% by weight slurry. This test was discontinued because of extremely poor performance.« less

Fuel properties effect on the performance of a small high temperature rise combustor

NASA Technical Reports Server (NTRS)

Acosta, Waldo A.; Beckel, Stephen A.

1989-01-01

The performance of an advanced small high temperature rise combustor was experimentally determined at NASA-Lewis. The combustor was designed to meet the requirements of advanced high temperature, high pressure ratio turboshaft engines. The combustor featured an advanced fuel injector and an advanced segmented liner design. The full size combustor was evaluated at power conditions ranging from idle to maximum power. The effect of broad fuel properties was studied by evaluating the combustor with three different fuels. The fuels used were JP-5, a blend of Diesel Fuel Marine/Home Heating Oil, and a blend of Suntec C/Home Heating Oil. The fuel properties effect on the performance of the combustion in terms of pattern factor, liner temperatures, and exhaust emissions are documented.

Imaging performance of annular apertures. IV - Apodization and point spread functions. V - Total and partial energy integral functions

NASA Technical Reports Server (NTRS)

Tschunko, H. F. A.

1983-01-01

Reference is made to a study by Tschunko (1979) in which it was discussed how apodization modifies the modulation transfer function for various central obstruction ratios. It is shown here how apodization, together with the central obstruction ratio, modifies the point spread function, which is the basic element for the comparison of imaging performance and for the derivation of energy integrals and other functions. At high apodization levels and lower central obstruction (less than 0.1), new extended radial zones are formed in the outer part of the central ring groups. These transmutation of the image functions are of more than theoretical interest, especially if the irradiance levels in the outer ring zones are to be compared to the background irradiance levels. Attention is then given to the energy distribution in point images generated by annular apertures apodized by various transmission functions. The total energy functions are derived; partial energy integrals are determined; and background irradiance functions are discussed.

In-pile test results of U-silicide or U-nitride coated U-7Mo particle dispersion fuel in Al

NASA Astrophysics Data System (ADS)

Kim, Yeon Soo; Park, J. M.; Lee, K. H.; Yoo, B. O.; Ryu, H. J.; Ye, B.

2014-11-01

U-silicide or U-nitride coated U-Mo particle dispersion fuel in Al (U-Mo/Al) was in-pile tested to examine the effectiveness of the coating as a diffusion barrier between the U-7Mo fuel kernels and Al matrix. This paper reports the PIE data and analyses focusing on the effectiveness of the coating in terms of interaction layer (IL) growth and general fuel performance. The U-silicide coating showed considerable success, but it also provided evidence for additional improvement for coating process. The U-nitride coated specimen showed largely inefficient results in reducing IL growth. From the test, important observations were also made that can be utilized to improve U-Mo/Al fuel performance. The heating process for coating turned out to be beneficial to suppress fuel swelling. The use of larger fuel particles confirmed favorable effects on fuel performance.

Proton Exchange Membrane Fuel Cells for Electrical Power Generation On-Board Commercial Airplanes

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pratt, Joesph W.; Klebanoff, Leonard E.; Munoz-Ramos, Karina

2011-05-01

Deployed on a commercial airplane, proton exchange membrane fuel cells may offer emissions reductions, thermal efficiency gains, and enable locating the power near the point of use. This work seeks to understand whether on-board fuel cell systems are technically feasible, and, if so, if they offer a performance advantage for the airplane as a whole. Through hardware analysis and thermodynamic and electrical simulation, we found that while adding a fuel cell system using today’s technology for the PEM fuel cell and hydrogen storage is technically feasible, it will not likely give the airplane a performance benefit. However, when we re-didmore » the analysis using DOE-target technology for the PEM fuel cell and hydrogen storage, we found that the fuel cell system would provide a performance benefit to the airplane (i.e., it can save the airplane some fuel), depending on the way it is configured.« less

Proton exchange membrane fuel cells for electrical power generation on-board commercial airplanes.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Curgus, Dita Brigitte; Munoz-Ramos, Karina; Pratt, Joseph William

2011-05-01

Deployed on a commercial airplane, proton exchange membrane fuel cells may offer emissions reductions, thermal efficiency gains, and enable locating the power near the point of use. This work seeks to understand whether on-board fuel cell systems are technically feasible, and, if so, if they offer a performance advantage for the airplane as a whole. Through hardware analysis and thermodynamic and electrical simulation, we found that while adding a fuel cell system using today's technology for the PEM fuel cell and hydrogen storage is technically feasible, it will not likely give the airplane a performance benefit. However, when we re-didmore » the analysis using DOE-target technology for the PEM fuel cell and hydrogen storage, we found that the fuel cell system would provide a performance benefit to the airplane (i.e., it can save the airplane some fuel), depending on the way it is configured.« less

Use of multi-functional flexible micro-sensors for in situ measurement of temperature, voltage and fuel flow in a proton exchange membrane fuel cell.

PubMed

Lee, Chi-Yuan; Chan, Pin-Cheng; Lee, Chung-Ju

2010-01-01

Temperature, voltage and fuel flow distribution all contribute considerably to fuel cell performance. Conventional methods cannot accurately determine parameter changes inside a fuel cell. This investigation developed flexible and multi-functional micro sensors on a 40 μm-thick stainless steel foil substrate by using micro-electro-mechanical systems (MEMS) and embedded them in a proton exchange membrane fuel cell (PEMFC) to measure the temperature, voltage and flow. Users can monitor and control in situ the temperature, voltage and fuel flow distribution in the cell. Thereby, both fuel cell performance and lifetime can be increased.

Thermal-mechanical performance modeling of thorium-plutonium oxide fuel and comparison with on-line irradiation data

NASA Astrophysics Data System (ADS)

Insulander Björk, Klara; Kekkonen, Laura

2015-12-01

Thorium-plutonium Mixed OXide (Th-MOX) fuel is considered for use in light water reactors fuel due to some inherent benefits over conventional fuel types in terms of neutronic properties. The good material properties of ThO2 also suggest benefits in terms of thermal-mechanical fuel performance, but the use of Th-MOX fuel for commercial power production demands that its thermal-mechanical behavior can be accurately predicted using a well validated fuel performance code. Given the scant operational experience with Th-MOX fuel, no such code is available today. This article describes the first phase of the development of such a code, based on the well-established code FRAPCON 3.4, and in particular the correlations reviewed and chosen for the fuel material properties. The results of fuel temperature calculations with the code in its current state of development are shown and compared with data from a Th-MOX test irradiation campaign which is underway in the Halden research reactor. The results are good for fresh fuel, whereas experimental complications make it difficult to judge the adequacy of the code for simulations of irradiated fuel.

77 FR 73968 - Reconsideration of Certain New Source and Startup/Shutdown Issues: National Emission Standards...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-12-12

... Units and Standards of Performance for Fossil-Fuel-Fired Electric Utility, Industrial-Commercial...- and Oil-fired Electric Utility Steam Generating Units and Standards of Performance for Fossil-Fuel... Units and Standards of Performance for Fossil-Fuel-Fired Electric Utility, Industrial-Commercial...

77 FR 9303 - National Emission Standards for Hazardous Air Pollutants From Coal- and Oil-Fired Electric...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-02-16

... Utility Steam Generating Units and Standards of Performance for Fossil-Fuel-Fired Electric Utility... Performance for Fossil-Fuel-Fired Electric Utility, Industrial-Commercial-Institutional, and Small Industrial... electric utility steam generating units (EGUs) and standards of performance for fossil-fuel-fired electric...

PERFORMANCE OF TWO LIQUID METAL TURBOPROP ENGINES UTILIZING A CIRCULATING FUEL REACTOR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tiedemann, H.J.; Mathews, L.

1955-01-20

The performance of two all-nuclear turboprop engines utilizing the circulating fuel reactor with a fluoride fuel temperature of I500 deg F was investigated. Data are presented for off-match-point and modified match-point performances. Results are given in graph form. (M.C.G.)

Effect of fuel properties on performance of a single aircraft turbojet combustor

NASA Technical Reports Server (NTRS)

Butze, H. F.; Ehlers, R. C.

1975-01-01

The performance of a single-can JT8D combustor was investigated with a number of fuels exhibiting wide variations in chemical composition and volatility. Performance parameters investigated were combustion efficiency, emissions of CO, unburned hydrocarbons and NOx, as well as liner temperatures and smoke. At the simulated idle condition no significant differences in performance were observed. At cruise, liner temperatures and smoke increased sharply with decreasing hydrogen content of the fuel. No significant differences were observed in the performance of an oil-shale derived JP-5 and a petroleum-based Jet A fuel except for emissions of NOx which were higher with the oil-shale JP-5. The difference is attributed to the higher concentration of fuel-bound nitrogen in the oil-shale JP-5.

40 CFR 60.46 - Test methods and procedures.

Code of Federal Regulations, 2012 CFR

2012-07-01

... (CONTINUED) STANDARDS OF PERFORMANCE FOR NEW STATIONARY SOURCES Standards of Performance for Fossil-Fuel... fossil fuels or fossil fuel and wood residue are fired, the owner or operator (in order to compute the...

40 CFR 60.46 - Test methods and procedures.

Code of Federal Regulations, 2014 CFR

2014-07-01

... (CONTINUED) STANDARDS OF PERFORMANCE FOR NEW STATIONARY SOURCES Standards of Performance for Fossil-Fuel... fossil fuels or fossil fuel and wood residue are fired, the owner or operator (in order to compute the...

40 CFR 60.46 - Test methods and procedures.

Code of Federal Regulations, 2013 CFR

2013-07-01

... (CONTINUED) STANDARDS OF PERFORMANCE FOR NEW STATIONARY SOURCES Standards of Performance for Fossil-Fuel... fossil fuels or fossil fuel and wood residue are fired, the owner or operator (in order to compute the...