Cadmium depletion impacts on hardening neutron spectrum for advanced fuel testing in ATR

International Nuclear Information System (INIS)

Chang, Gray S.

2011-01-01

For transmuting long-lived isotopes contained in spent nuclear fuel into shorter-lived fission products effectively is in a fast neutron spectrum reactor. In the absence of a fast spectrum test reactor in the United States of America (USA), 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. A test region is achieved with a Cadmium (Cd) filter which can harden the neutron spectrum to a spectrum similar (although still somewhat softer) to that of the liquid metal fast breeder reactor (LMFBR). A fuel test loop with a Cd-filter has been installed within the East Flux Trap (EFT) of the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL). A detailed comparison analyses between the cadmium (Cd) filter hardened neutron spectrum in the ATR and the LMFBR fast neutron spectrum have been performed using MCWO. MCWO is a set of scripting tools that are used to couple the Monte Carlo transport code MCNP with the isotope depletion and buildup code ORIGEN-2.2. The MCWO-calculated results indicate that the Cd-filter can effectively flatten the Rim-Effect and reduce the linear heat rate (LHGR) to meet the advanced fuel testing project requirements at the beginning of irradiation (BOI). However, the filtering characteristics of Cd as a strong absorber quickly depletes over time, and the Cd-filter must be replaced for every two typical operating cycles within the EFT of the ATR. The designed Cd-filter can effectively depress the LHGR in experimental fuels and harden the neutron spectrum enough to adequately flatten the Rim-Effect in the test region. (author)

Initiation of depleted uranium oxide and spent fuel testing for the spent fuel sabotage aerosol ratio program

Energy Technology Data Exchange (ETDEWEB)

Molecke, M.A.; Gregson, M.W.; Sorenson, K.B. [Sandia National Labs. (United States); Billone, M.C.; Tsai, H. [Argonne National Lab. (United States); Koch, W.; Nolte, O. [Fraunhofer Inst. fuer Toxikologie und Experimentelle Medizin (Germany); Pretzsch, G.; Lange, F. [Gesellschaft fuer Anlagen- und Reaktorsicherheit (Germany); Autrusson, B.; Loiseau, O. [Inst. de Radioprotection et de Surete Nucleaire (France); Thompson, N.S.; Hibbs, R.S. [U.S. Dept. of Energy (United States); Young, F.I.; Mo, T. [U.S. Nuclear Regulatory Commission (United States)

2004-07-01

We provide a detailed overview of an ongoing, multinational test program that is developing aerosol data for some spent fuel sabotage scenarios on spent fuel transport and storage casks. Experiments are being performed to quantify the aerosolized materials plus volatilized fission products generated from actual spent fuel and surrogate material test rods, due to impact by a high energy density device, HEDD. The program participants in the U.S. plus Germany, France, and the U.K., part of the international Working Group for Sabotage Concerns of Transport and Storage Casks, WGSTSC have strongly supported and coordinated this research program. Sandia National Laboratories, SNL, has the lead role for conducting this research program; test program support is provided by both the U.S. Department of Energy and Nuclear Regulatory Commission. WGSTSC partners need this research to better understand potential radiological impacts from sabotage of nuclear material shipments and storage casks, and to support subsequent risk assessments, modeling, and preventative measures. We provide a summary of the overall, multi-phase test design and a description of all explosive containment and aerosol collection test components used. We focus on the recently initiated tests on ''surrogate'' spent fuel, unirradiated depleted uranium oxide, and forthcoming actual spent fuel tests. The depleted uranium oxide test rodlets were prepared by the Institut de Radioprotection et de Surete Nucleaire, in France. These surrogate test rodlets closely match the diameter of the test rodlets of actual spent fuel from the H.B. Robinson reactor (high burnup PWR fuel) and the Surry reactor (lower, medium burnup PWR fuel), generated from U.S. reactors. The characterization of the spent fuels and fabrication into short, pressurized rodlets has been performed by Argonne National Laboratory, for testing at SNL. The ratio of the aerosol and respirable particles released from HEDD-impacted spent

Nuclear Fuel Depletion Analysis Using Matlab Software

Science.gov (United States)

Faghihi, F.; Nematollahi, M. R.

Coupled first order IVPs are frequently used in many parts of engineering and sciences. In this article, we presented a code including three computer programs which are joint with the Matlab software to solve and plot the solutions of the first order coupled stiff or non-stiff IVPs. Some engineering and scientific problems related to IVPs are given and fuel depletion (production of the 239Pu isotope) in a Pressurized Water Nuclear Reactor (PWR) are computed by the present code.

Theoretical analysis of nuclear reactors (Phase I), I-V, Part IV, Nuclear fuel depletion

International Nuclear Information System (INIS)

Pop-Jordanov, J.

1962-07-01

Nuclear fuel depletion is analyzed in order to estimate the qualitative and quantitative fuel property changes during irradiation and the influence of changes on the reactivity during long-term reactor operation. The changes of fuel properties are described by changes of neutron absorption and fission cross sections. Part one of this report covers the economic significance of fuel burnup and the review of fuel isotopic changes during depletion. Pat two contains the analysis of the U 235 chain, analytical expressions for the concentrations of U 235 , U 236 and Np 237 as a function of burnup. Part three contains the analysis of neutron spectrum influence on the Westcott method for calculating the cross sections. Part four contains the calculation method applied on Calder Hall type reactor. The results were obtained by applying ZUSE-22 R digital computer

Separation and recovery method for depleted uranium from spent fuel

International Nuclear Information System (INIS)

Imoto, Yoshie; Fujita, Reiko.

1993-01-01

Spent oxide fuels are reduced in a molten salt of CaCl 2 -CaF 2 to convert them into metals, then melted in an Fe-U bath disposed in an electrolytic refining vessel and brought into contact with molten Mg, to extract transuranium elements and rare earth elements contained in the Fe-U bath as metals in the molten Mg. Then molten Mg is removed and the residue is brought into contact with KCl-LiCl molten salt and electrolyzed using the Fe-U as an anode. Then, uranium is recovered by deposition on an iron cathode disposed in chloride electrolytes of the electrolytic refining vessel. Uranium and transuranium elements can be thus separated and, for example, depleted uranium for use in blanket fuels can be recovered easily. This can greatly reduce the temporary storage amount of depleted uranium, to eliminate requirement for a large-scaled facility used exclusively for storing uranium and long time management for uranium. (T.M.)

EPRI depletion benchmark calculations using PARAGON

International Nuclear Information System (INIS)

Kucukboyaci, Vefa N.

2015-01-01

Highlights: • PARAGON depletion calculations are benchmarked against the EPRI reactivity decrement experiments. • Benchmarks cover a wide range of enrichments, burnups, cooling times, and burnable absorbers, and different depletion and storage conditions. • Results from PARAGON-SCALE scheme are more conservative relative to the benchmark data. • ENDF/B-VII based data reduces the excess conservatism and brings the predictions closer to benchmark reactivity decrement values. - Abstract: In order to conservatively apply burnup credit in spent fuel pool criticality analyses, code validation for both fresh and used fuel is required. Fresh fuel validation is typically done by modeling experiments from the “International Handbook.” A depletion validation can determine a bias and bias uncertainty for the worth of the isotopes not found in the fresh fuel critical experiments. Westinghouse’s burnup credit methodology uses PARAGON™ (Westinghouse 2-D lattice physics code) and its 70-group cross-section library, which have been benchmarked, qualified, and licensed both as a standalone transport code and as a nuclear data source for core design simulations. A bias and bias uncertainty for the worth of depletion isotopes, however, are not available for PARAGON. Instead, the 5% decrement approach for depletion uncertainty is used, as set forth in the Kopp memo. Recently, EPRI developed a set of benchmarks based on a large set of power distribution measurements to ascertain reactivity biases. The depletion reactivity has been used to create 11 benchmark cases for 10, 20, 30, 40, 50, and 60 GWd/MTU and 3 cooling times 100 h, 5 years, and 15 years. These benchmark cases are analyzed with PARAGON and the SCALE package and sensitivity studies are performed using different cross-section libraries based on ENDF/B-VI.3 and ENDF/B-VII data to assess that the 5% decrement approach is conservative for determining depletion uncertainty

Nonlinear analyses of spent-fuel racks for consolidated fuel loading

International Nuclear Information System (INIS)

Kabir, A.F.; Godha, P.C.; Malik, L.E.; Bolourchi, S.

1987-01-01

Storage racks for spent-fuel assemblies in nuclear power plants are designed to withstand various combinations of loads generated by gravity, seismic, thermal, and accidental fuel drops. Due to the need for storing increased amounts of spent fuel in the existing fuel pools, many nuclear power utilities are evaluating existing fuel racks to safely carry the additional loads. The current study presents the seismic analyses of existing fuel racks of Northeast Utility Company's Millstone Unit Number 1 (BWR Mark I) nuclear plant to accommodate a 2:1 fuel consolidation. This objective requires rigorous nonlinear analyses to establish the full available capacities of the racks and thereby avoid expensive modifications or minimize any needed upgrades

Implantation of a new calculation method of fuel depletion in the CITHAM code

International Nuclear Information System (INIS)

Alvarenga, M.A.B.

1985-01-01

It is evaluated the accuracy of the linear aproximation method used in the CITHAN code to obtain the solution of depletion equations. Results are compared with the Benchmark problem. The convenience of depletion chain before criticality calculations is analysed. The depletion calculation was modified using linear combination technic of linear chains. (M.C.K.) [pt

Application of depletion perturbation theory to fuel cycle burnup analysis

International Nuclear Information System (INIS)

White, J.R.

1979-01-01

Over the past several years static perturbation theory methods have been increasingly used for reactor analysis in lieu of more detailed and costly direct computations. Recently, perturbation methods incorporating time dependence have also received attention, and several authors have demonstrated their applicability to fuel burnup analysis. The objective of the work described here is to demonstrate that a time-dependent perturbation method can be easily and accurately applied to realistic depletion problems

Comparison of MCNPX-C90 and TRIPOLI-4-D for fuel depletion calculations of a Gas-cooled Fast Reactor

International Nuclear Information System (INIS)

Reyes-Ramirez, Ricardo; Martin-del-Campo, Cecilia; Francois, Juan-Luis; Brun, Emeric; Dumonteil, Eric; Malvagi, Fausto

2010-01-01

The Gas-cooled Fast Reactor is one of the reactor concepts selected by the Generation IV International Forum for the next generation of innovative nuclear energy systems. Several fuel design concepts are being investigated. Burnup depletion of mixed fuel of uranium and plutonium, cooled with gas in a fast neutron energy spectrum must be simulated. Various codes are being developed and/or adapted to improve the quality of the results, and also to reduce the computing time required for the simulations. The main objective of this work is to compare the fuel depletion results obtained with MCNPX-CINDER90 code and the new TRIPOLI-4-Depletion code (developed by the Commissariat a l'Energie Atomique) of a fuel design concept for the Gas-cooled Fast Reactor. Calculations were made for an equivalent homogeneous model of fuel rods in a hexagonal mesh assembly. Total reflection conditions were applied on the six lateral faces and the two axial faces of the assembly. The materials used in the fuel assembly are: carbide of uranium and plutonium as fuel, silicon carbide as cladding, and helium gas as coolant. JEFF libraries of effective cross sections were used in both codes. Two methods of burnup step calculations were performed with TRIPOLI-4-D, the Euler and the CSADA, and their results were compared with the MCNPX-CINDER90 CSADA method. A period of 300 days of irradiation time was considered, which was divided into 12 steps. Results of the infinite multiplication factor as function of the irradiation time, and the evolution of the isotope concentrations for a selected group of nuclides were compared. The main conclusion is that very similar results were obtained for the three types of depletion calculations which were compared: (1) MCNPX-C90 CSADA; (2) TRIPOLI-4-D CSADA, and (3) TRIPOLI-4-D EULER. The best calculation time was obtained with the TRIPOLI-4-D EULER method, which needed approximately half the time than the other two. In summary, it is sufficiently good to use

Constraints of fossil fuels depletion on global warming projections

International Nuclear Information System (INIS)

Chiari, Luca; Zecca, Antonio

2011-01-01

A scientific debate is in progress about the intersection of climate change with the new field of fossil fuels depletion geology. Here, new projections of atmospheric CO 2 concentration and global-mean temperature change are presented, should fossil fuels be exploited at a rate limited by geological availability only. The present work starts from the projections of fossil energy use, as obtained from ten independent sources. From such projections an upper bound, a lower bound and an ensemble mean profile for fossil CO 2 emissions until 2200 are derived. Using the coupled gas-cycle/climate model MAGICC, the corresponding climatic projections out to 2200 are obtained. We find that CO 2 concentration might increase up to about 480 ppm (445-540 ppm), while the global-mean temperature increase w.r.t. 2000 might reach 1.2 deg. C (0.9-1.6 deg. C). However, future improvements of fossil fuels recovery and discoveries of new resources might lead to higher emissions; hence our climatic projections are likely to be underestimated. In the absence of actions of emissions reduction, a level of dangerous anthropogenic interference with the climate system might be already experienced toward the middle of the 21st century, despite the constraints imposed by the exhaustion of fossil fuels. - Highlights: → CO 2 and global temperature are projected under fossil fuels exhaustion scenarios. → Temperature is projected to reach a minimum of 2 deg. C above pre-industrial. → Temperature projections are possibly lower than the IPCC ones. → Fossil fuels exhaustion will not avoid dangerous global warming.

Constraints of fossil fuels depletion on global warming projections

Energy Technology Data Exchange (ETDEWEB)

Chiari, Luca, E-mail: chiari@science.unitn.it [Department of Physics, University of Trento, Via Sommarive 14, 38123 Povo (Italy); Zecca, Antonio, E-mail: zecca@science.unitn.it [Department of Physics, University of Trento, Via Sommarive 14, 38123 Povo (Italy)

2011-09-15

A scientific debate is in progress about the intersection of climate change with the new field of fossil fuels depletion geology. Here, new projections of atmospheric CO{sub 2} concentration and global-mean temperature change are presented, should fossil fuels be exploited at a rate limited by geological availability only. The present work starts from the projections of fossil energy use, as obtained from ten independent sources. From such projections an upper bound, a lower bound and an ensemble mean profile for fossil CO{sub 2} emissions until 2200 are derived. Using the coupled gas-cycle/climate model MAGICC, the corresponding climatic projections out to 2200 are obtained. We find that CO{sub 2} concentration might increase up to about 480 ppm (445-540 ppm), while the global-mean temperature increase w.r.t. 2000 might reach 1.2 deg. C (0.9-1.6 deg. C). However, future improvements of fossil fuels recovery and discoveries of new resources might lead to higher emissions; hence our climatic projections are likely to be underestimated. In the absence of actions of emissions reduction, a level of dangerous anthropogenic interference with the climate system might be already experienced toward the middle of the 21st century, despite the constraints imposed by the exhaustion of fossil fuels. - Highlights: > CO{sub 2} and global temperature are projected under fossil fuels exhaustion scenarios. > Temperature is projected to reach a minimum of 2 deg. C above pre-industrial. > Temperature projections are possibly lower than the IPCC ones. > Fossil fuels exhaustion will not avoid dangerous global warming.

Real depletion in nodal diffusion codes

International Nuclear Information System (INIS)

Petkov, P.T.

2002-01-01

The fuel depletion is described by more than one hundred fuel isotopes in the advanced lattice codes like HELIOS, but only a few fuel isotopes are accounted for even in the advanced steady-state diffusion codes. The general assumption that the number densities of the majority of the fuel isotopes depend only on the fuel burnup is seriously in error if high burnup is considered. The real depletion conditions in the reactor core differ from the asymptotic ones at the stage of lattice depletion calculations. This study reveals which fuel isotopes should be explicitly accounted for in the diffusion codes in order to predict adequately the real depletion effects in the core. A somewhat strange conclusion is that if the real number densities of the main fissionable isotopes are not explicitly accounted for in the diffusion code, then Sm-149 should not be accounted for either, because the net error in k-inf is smaller (Authors)

Investigation of Burnup Credit Issues in BWR Fuel

International Nuclear Information System (INIS)

Broadhead, B.L.; DeHart, M.D.

1999-01-01

Calculations for long-term-disposal criticality safety of spent nuclear fuel requires the application of burnup credit because of the large mass of fissile material that will be present in the repository. Burnup credit calculations are based on depletion calculations that provide a conservative estimate of spent fuel contents, followed by criticality calculations to assess the value of keff for a spent fuel cask or a fuel configuration under a variety of probabilistically derived events. In order to ensure that the depletion calculation is conservative, it is necessary to both qualify and quantify assumptions that can be made in depletion models used to characterize spent fuel. Most effort in the United States this decade has focused on burnup issues related to pressurized-water reactors. However, requirements for the permanent disposal of fuel from boiling-water reactors has necessitated development of methods for prediction of spent fuel contents for such fuels. Concomitant with such analyses, validation is also necessary. This paper provides a summary of initial efforts at the Oak Ridge National Laboratory to better understand and validate spent fuel analyses for boiling-water-reactor fuel

Contribution to fuel depletion study in PWR type reactors, reactor core with three and four regions of enrichment

International Nuclear Information System (INIS)

Teixeira, M.C.C.

1977-03-01

The main methods for calculation of fuel depletion are studied and some approaches to do it are mentioned; the LEOPARD Code is described and full details are given for each subroutine, flow charts are included; the method given by the code for calculation of fuel depletion is described; some imperfections from the IPR's version are listed, and corrected, for instance: the method for burn-up calculation of heavy isotopes; the results of calculations for a reference reactor based on data of the Preliminary Safety Analysis Report (PSAR) for Angra I Nuclear Power Plant are presented and discussed. (author)

Verification of the depletion capabilities of the MCNPX code on a LWR MOX fuel assembly

International Nuclear Information System (INIS)

Cerba, S.; Hrncir, M.; Necas, V.

2012-01-01

The study deals with the verification of the depletion capabilities of the MCNPX code, which is a linked Monte-Carlo depletion code. For such a purpose the IV-B phase of the OECD NEA Burnup credit benchmark has been chosen. The mentioned benchmark is a code to code comparison of the multiplication coefficient k eff and the isotopic composition of a LWR MOX fuel assembly at three given burnup levels and after five years of cooling. The benchmark consists of 6 cases, 2 different Pu vectors and 3 geometry models, however in this study only the fuel assembly calculations with two Pu vectors were performed. The aim of this study was to compare the obtained result with data from the participants of the OECD NEA Burnup Credit project and confirm the burnup capability of the MCNPX code. (Authors)

Initiation of depleted uranium oxide and spent fuel testing for the spent fuel sabotage aerosol ratio programme

International Nuclear Information System (INIS)

Molecke, M.A.; Gregson, M.W.; Sorenson, K.B.

2004-01-01

We provide a detailed overview of an on-going, multinational test programme that is developing aerosol data for some spent fuel sabotage scenarios on spent fuel transport and storage casks. Experiments are being performed to quantify the aerosolised materials plus volatilised fission products generated from actual spent fuel and surrogate material test rods, due to impact by a high-energy/density device. The programme participants in the United States plus Germany, France and the United Kingdom, part of the international Working Group for Sabotage Concerns of Transport and Storage Casks (WGSTSC) have strongly supported and coordinated this research programme. Sandia National Laboratories has the lead role for conducting this research programme; test programme support is provided by both the US Department of Energy and the US Nuclear Regulatory Commission. We provide a summary of the overall, multiphase test design and a description of all explosive containment and aerosol collection test components used. We focus on the recently initiated tests on 'surrogate' spent fuel, unirradiated depleted uranium oxide and forthcoming actual spent fuel tests. We briefly summarise similar results from completed surrogate tests that used non-radioactive, sintered cerium oxide ceramic pellets in test rods. (author)

Improvements of MCOR: A Monte Carlo depletion code system for fuel assembly reference calculations

Energy Technology Data Exchange (ETDEWEB)

Tippayakul, C.; Ivanov, K. [Pennsylvania State Univ., Univ. Park (United States); Misu, S. [AREVA NP GmbH, An AREVA and SIEMENS Company, Erlangen (Germany)

2006-07-01

This paper presents the improvements of MCOR, a Monte Carlo depletion code system for fuel assembly reference calculations. The improvements of MCOR were initiated by the cooperation between the Penn State Univ. and AREVA NP to enhance the original Penn State Univ. MCOR version in order to be used as a new Monte Carlo depletion analysis tool. Essentially, a new depletion module using KORIGEN is utilized to replace the existing ORIGEN-S depletion module in MCOR. Furthermore, the online burnup cross section generation by the Monte Carlo calculation is implemented in the improved version instead of using the burnup cross section library pre-generated by a transport code. Other code features have also been added to make the new MCOR version easier to use. This paper, in addition, presents the result comparisons of the original and the improved MCOR versions against CASMO-4 and OCTOPUS. It was observed in the comparisons that there were quite significant improvements of the results in terms of k{sub inf}, fission rate distributions and isotopic contents. (authors)

Depleted uranium oxides as spent-nuclear-fuel waste-package invert and backfill materials

International Nuclear Information System (INIS)

Forsberg, C.W.; Haire, M.J.

1997-01-01

A new technology has been proposed in which depleted uranium, in the form of oxides or silicates, is placed around the outside of the spent nuclear fuel waste packages in the geological repository. This concept may (1) reduce the potential for repository nuclear criticality events and (2) reduce long-term release of radionuclides from the repository. As a new concept, there are significant uncertainties

Fossil fuel depletion and socio-economic scenarios: An integrated approach

International Nuclear Information System (INIS)

Capellán-Pérez, Iñigo; Mediavilla, Margarita; Castro, Carlos de; Carpintero, Óscar; Miguel, Luis Javier

2014-01-01

The progressive reduction of high-quality-easy-to-extract energy is a widely recognized and already ongoing process. Although depletion studies for individual fuels are relatively abundant, few of them offer a global perspective of all energy sources and their potential future developments, and even fewer include the demand of the socio-economic system. This paper presents an Economy-Energy-Environment model based on System Dynamics which integrates all those aspects: the physical restrictions (with peak estimations for oil, gas, coal and uranium), the techno-sustainable potential of renewable energy estimated by a novel top-down methodology, the socio-economic energy demands, the development of alternative technologies and the net CO 2 emissions. We confront our model with the basic assumptions of previous Global Environmental Assessment (GEA) studies. The results show that demand-driven evolution, as performed in the past, might be unfeasible: strong energy-supply scarcity is found in the next two decades, especially in the transportation sector before 2020. Electricity generation is unable to fulfill its demand in 2025–2040, and a large expansion of electric renewable energies move us close to their limits. In order to find achievable scenarios, we are obliged to set hypotheses which are hardly used in GEA scenarios, such as zero or negative economic growth. - Highlights: • The paper presents and describes a new Energy–Economy–Environment global model. • GEA scenario dynamics have the potential to lead us to energy resource scarcity in the next 2 decades. • Global forecasts of international agencies show inconsistency in energy constraints. • Renewable energies are only partially able to replace fossil fuels depletion. • Climate change still reaches dangerous dimensions

Monte Carlo simulation in UWB1 depletion code

International Nuclear Information System (INIS)

Lovecky, M.; Prehradny, J.; Jirickova, J.; Skoda, R.

2015-01-01

U W B 1 depletion code is being developed as a fast computational tool for the study of burnable absorbers in the University of West Bohemia in Pilsen, Czech Republic. In order to achieve higher precision, the newly developed code was extended by adding a Monte Carlo solver. Research of fuel depletion aims at development and introduction of advanced types of burnable absorbers in nuclear fuel. Burnable absorbers (BA) allow the compensation of the initial reactivity excess of nuclear fuel and result in an increase of fuel cycles lengths with higher enriched fuels. The paper describes the depletion calculations of VVER nuclear fuel doped with rare earth oxides as burnable absorber based on performed depletion calculations, rare earth oxides are divided into two equally numerous groups, suitable burnable absorbers and poisoning absorbers. According to residual poisoning and BA reactivity worth, rare earth oxides marked as suitable burnable absorbers are Nd, Sm, Eu, Gd, Dy, Ho and Er, while poisoning absorbers include Sc, La, Lu, Y, Ce, Pr and Tb. The presentation slides have been added to the article

Development of methods for theoretical analysis of nuclear reactors (Phase II), I-V, Part IV, Fuel depletion

International Nuclear Information System (INIS)

Pop-Jordanov, J.

1962-10-01

This report includes the analysis of plutonium isotopes from U 238 depletion chain. Two theoretical approaches for solving the depletion of fuel are shown. One results in the system of differential equations that can be solved only by using electronic calculators and the second, Machinari-Goto method enables obtaining analytical equations for approximative values of particular nuclei. In addition, differential equations are given for different approximation levels in calculating Pu 239 , as well as relations between the released energy and irradiation [sr

Statistical implications in Monte Carlo depletions - 051

International Nuclear Information System (INIS)

Zhiwen, Xu; Rhodes, J.; Smith, K.

2010-01-01

As a result of steady advances of computer power, continuous-energy Monte Carlo depletion analysis is attracting considerable attention for reactor burnup calculations. The typical Monte Carlo analysis is set up as a combination of a Monte Carlo neutron transport solver and a fuel burnup solver. Note that the burnup solver is a deterministic module. The statistical errors in Monte Carlo solutions are introduced into nuclide number densities and propagated along fuel burnup. This paper is towards the understanding of the statistical implications in Monte Carlo depletions, including both statistical bias and statistical variations in depleted fuel number densities. The deterministic Studsvik lattice physics code, CASMO-5, is modified to model the Monte Carlo depletion. The statistical bias in depleted number densities is found to be negligible compared to its statistical variations, which, in turn, demonstrates the correctness of the Monte Carlo depletion method. Meanwhile, the statistical variation in number densities generally increases with burnup. Several possible ways of reducing the statistical errors are discussed: 1) to increase the number of individual Monte Carlo histories; 2) to increase the number of time steps; 3) to run additional independent Monte Carlo depletion cases. Finally, a new Monte Carlo depletion methodology, called the batch depletion method, is proposed, which consists of performing a set of independent Monte Carlo depletions and is thus capable of estimating the overall statistical errors including both the local statistical error and the propagated statistical error. (authors)

Work plan for improving the DARWIN2.3 depleted material balance calculation of nuclides of interest for the fuel cycle

Directory of Open Access Journals (Sweden)

Rizzo Axel

2017-01-01

Full Text Available DARWIN2.3 is the reference package used for fuel cycle applications in France. It solves the Boltzmann and Bateman equations in a coupling way, with the European JEFF-3.1.1 nuclear data library, to compute the fuel cycle values of interest. It includes both deterministic transport codes APOLLO2 (for light water reactors and ERANOS2 (for fast reactors, and the DARWIN/PEPIN2 depletion code, each of them being developed by CEA/DEN with the support of its industrial partners. The DARWIN2.3 package has been experimentally validated for pressurized and boiling water reactors, as well as for sodium fast reactors; this experimental validation relies on the analysis of post-irradiation experiments (PIE. The DARWIN2.3 experimental validation work points out some isotopes for which the depleted concentration calculation can be improved. Some other nuclides have no available experimental validation, and their concentration calculation uncertainty is provided by the propagation of a priori nuclear data uncertainties. This paper describes the work plan of studies initiated this year to improve the accuracy of the DARWIN2.3 depleted material balance calculation concerning some nuclides of interest for the fuel cycle.

Work plan for improving the DARWIN2.3 depleted material balance calculation of nuclides of interest for the fuel cycle

Science.gov (United States)

Rizzo, Axel; Vaglio-Gaudard, Claire; Martin, Julie-Fiona; Noguère, Gilles; Eschbach, Romain

2017-09-01

DARWIN2.3 is the reference package used for fuel cycle applications in France. It solves the Boltzmann and Bateman equations in a coupling way, with the European JEFF-3.1.1 nuclear data library, to compute the fuel cycle values of interest. It includes both deterministic transport codes APOLLO2 (for light water reactors) and ERANOS2 (for fast reactors), and the DARWIN/PEPIN2 depletion code, each of them being developed by CEA/DEN with the support of its industrial partners. The DARWIN2.3 package has been experimentally validated for pressurized and boiling water reactors, as well as for sodium fast reactors; this experimental validation relies on the analysis of post-irradiation experiments (PIE). The DARWIN2.3 experimental validation work points out some isotopes for which the depleted concentration calculation can be improved. Some other nuclides have no available experimental validation, and their concentration calculation uncertainty is provided by the propagation of a priori nuclear data uncertainties. This paper describes the work plan of studies initiated this year to improve the accuracy of the DARWIN2.3 depleted material balance calculation concerning some nuclides of interest for the fuel cycle.

Power distribution and fuel depletion calculation for a PWR, using LEOPARD and CITATION codes

International Nuclear Information System (INIS)

Batista, J.L.

1982-01-01

By modifying LEOPARD a new program, LEOCIT, has been developed in which additional subroutines prepare cross-section libraries in 1, 2 or 4 energy groups and subsequently record these on disc or tape in a format appropriate for direct input to the CITATION code. Use of LEOCIT in conjunction with CITATION is demonstrated by simulating the first depletion cycle of Angra Unit 1. In these calculations two energy groups are used in 1/4, X - Y geometry to give the soluble boron curve, the fuel depletion and the point to point power distribution in Angra 1. Finally relevant results obtained here are compared with those published by Westinghouse, CNEN and Furnas and recommendations are made to improve the system of neutronic calculation developed in this work. (Author) [pt

Depletion of gadolinium burnable poison in a PWR assembly with high burnup fuel

Energy Technology Data Exchange (ETDEWEB)

Refeat, Riham Mahmoud [Nuclear and Radiological Regulatory Authority (NRRA), Cairo (Egypt). Safety Engineering Dept.

2015-12-15

A tendency to increase the discharge burnup of nuclear fuel for Advanced Pressurized Water Reactors (PWR) has been a characteristic of its operation for many years. It will be able to burn at very high burnup of about 70 GWd/t with UO{sub 2} fuels. The U-235 enrichment must be higher than 5 %, which leads to the necessity of using an extremely efficient burnable poison like Gadolinium oxide. Using gadolinium isotope is significant due to its particular depletion behavior (''Onion-Skin'' effect). In this paper, the MCNPX2.7 code is used to calculate the important neutronic parameters of the next generation fuels of PWR. K-infinity, local peaking factor and fission rate distributions are calculated for a PWR assembly which burn at very high burnup reaching 70 GWd/t. The calculations are performed using the recently released evaluated Gadolinium cross section data. The results obtained are close to those of a LWR next generation fuel benchmark problem. This demonstrates that the calculation scheme used is able to accurately model a PWR assembly that operates at high burnup values.

Optimal reload and depletion method for pressurized water reactors

International Nuclear Information System (INIS)

Ahn, D.H.

1984-01-01

A new method has been developed to automatically reload and deplete a PWR so that both the enriched inventory requirements during the reactor cycle and the cost of reloading the core are minimized. This is achieved through four stepwise optimization calculations: 1) determination of the minimum fuel requirement for an equivalent three-region core model, 2) optimal selection and allocation of fuel requirement for an equivalent three-region core model, 2) optimal selection and allocation of fuel assemblies for each of the three regions to minimize the cost of the fresh reload fuel, 3) optimal placement of fuel assemblies to conserve regionwise optimal conditions and 4) optimal control through poison management to deplete individual fuel assemblies to maximize EOC k/sub eff/. Optimizing the fuel cost of reloading and depleting a PWR reactor cycle requires solutions to two separate optimization calculations. One of these minimizes the enriched fuel inventory in the core by optimizing the EOC k/sub eff/. The other minimizes the cost of the fresh reload cost. Both of these optimization calculations have now been combined to provide a new method for performing an automatic optimal reload of PWR's. The new method differs from previous methods in that the optimization process performs all tasks required to reload and deplete a PWR

MOx Depletion Calculation Benchmark

International Nuclear Information System (INIS)

San Felice, Laurence; Eschbach, Romain; Dewi Syarifah, Ratna; Maryam, Seif-Eddine; Hesketh, Kevin

2016-01-01

Under the auspices of the NEA Nuclear Science Committee (NSC), the Working Party on Scientific Issues of Reactor Systems (WPRS) has been established to study the reactor physics, fuel performance, radiation transport and shielding, and the uncertainties associated with modelling of these phenomena in present and future nuclear power systems. The WPRS has different expert groups to cover a wide range of scientific issues in these fields. The Expert Group on Reactor Physics and Advanced Nuclear Systems (EGRPANS) was created in 2011 to perform specific tasks associated with reactor physics aspects of present and future nuclear power systems. EGRPANS provides expert advice to the WPRS and the nuclear community on the development needs (data and methods, validation experiments, scenario studies) for different reactor systems and also provides specific technical information regarding: core reactivity characteristics, including fuel depletion effects; core power/flux distributions; Core dynamics and reactivity control. In 2013 EGRPANS published a report that investigated fuel depletion effects in a Pressurised Water Reactor (PWR). This was entitled 'International Comparison of a Depletion Calculation Benchmark on Fuel Cycle Issues' NEA/NSC/DOC(2013) that documented a benchmark exercise for UO 2 fuel rods. This report documents a complementary benchmark exercise that focused on PuO 2 /UO 2 Mixed Oxide (MOX) fuel rods. The results are especially relevant to the back-end of the fuel cycle, including irradiated fuel transport, reprocessing, interim storage and waste repository. Saint-Laurent B1 (SLB1) was the first French reactor to use MOx assemblies. SLB1 is a 900 MWe PWR, with 30% MOx fuel loading. The standard MOx assemblies, used in Saint-Laurent B1 reactor, include three zones with different plutonium enrichments, high Pu content (5.64%) in the center zone, medium Pu content (4.42%) in the intermediate zone and low Pu content (2.91%) in the peripheral zone

DUSCOBS - a depleted-uranium silicate backfill for transport, storage, and disposal of spent nuclear fuel

International Nuclear Information System (INIS)

Forsberg, C.W.; Pope, R.B.; Ashline, R.C.; DeHart, M.D.; Childs, K.W.; Tang, J.S.

1995-01-01

A Depleted Uranium Silicate COntainer Backfill System (DUSCOBS) is proposed that would use small, isotopically-depleted uranium silicate glass beads as a backfill material inside storage, transport, and repository waste packages containing spent nuclear fuel (SNF). The uranium silicate glass beads would fill all void space inside the package including the coolant channels inside SNF assemblies. Based on preliminary analysis, the following benefits have been identified. DUSCOBS improves repository waste package performance by three mechanisms. First, it reduces the radionuclide releases from SNF when water enters the waste package by creating a local uranium silicate saturated groundwater environment that suppresses (1) the dissolution and/or transformation of uranium dioxide fuel pellets and, hence, (2) the release of radionuclides incorporated into the SNF pellets. Second, the potential for long-term nuclear criticality is reduced by isotopic exchange of enriched uranium in SNF with the depleted uranium (DU) in the glass. Third, the backfill reduces radiation interactions between SNF and the local environment (package and local geology) and thus reduces generation of hydrogen, acids, and other chemicals that degrade the waste package system. In addition, the DUSCOBS improves the integrity of the package by acting as a packing material and ensures criticality control for the package during SNF storage and transport. Finally, DUSCOBS provides a potential method to dispose of significant quantities of excess DU from uranium enrichment plants at potential economic savings. DUSCOBS is a new concept. Consequently, the concept has not been optimized or demonstrated in laboratory experiments

Use of depleted uranium silicate glass to minimize release of radionuclides from spent nuclear fuel waste packages

International Nuclear Information System (INIS)

Forsberg, C.W.

1996-01-01

A Depleted Uranium Silicate Container Backfill System (DUSCOBS) is proposed that would use small, isotopically-depleted uranium silicate glass beads as a backfill material inside repository waste packages containing spent nuclear fuel (SNF). The uranium silicate glass beads would fill the void space inside the package including the coolant channels inside SNF assemblies. Based on preliminary analysis, the following benefits have been identified. DUSCOBS improves repository waste package performance by three mechanisms. First, it reduces the radionuclide releases from SNF when water enters the waste package by creating a local uranium silicate saturated groundwater environment that suppresses (a) the dissolution and/or transformation of uranium dioxide fuel pellets and, hence, (b) the release of radionuclides incorporated into the SNF pellets. Second, the potential for long-term nuclear criticality is reduced by isotopic exchange of enriched uranium in SNF with the depleted uranium (DU) in the glass. Third, the backfill reduces radiation interactions between SNF and the local environment (package and local geology) and thus reduces generation of hydrogen, acids, and other chemicals that degrade the waste package system. Finally, DUSCOBS provides a potential method to dispose of significant quantities of excess DU from uranium enrichment plants at potential economic savings. DUSCOBS is a new concept. Consequently, the concept has not been optimized or demonstrated in laboratory experiments

Accelerated fuel depreciation as an economic incentive for low-leakage fuel management

International Nuclear Information System (INIS)

Downar, T.J.

1986-01-01

An analysis is presented which evaluates the tax depreciation advantage which results from the increased rate of fuel depletion achieved in the current low-leakage fuel-management LWR core reload designs. An analytical fuel-cycle cost model is used to examine the important cost parameters which are then validated using the fuel-cycle cost code CINCAS and data from the Maine Yankee PWR. Results show that low-leakage fuel management, through the tax depreciation advantage from accelerated fuel depletion, provides an improvement of several percent in fuel-cycle costs compared to traditional out-in fuel management and a constant fuel depletion rate. (author)

Used Fuel Management System Interface Analyses - 13578

Energy Technology Data Exchange (ETDEWEB)

Howard, Robert; Busch, Ingrid [Oak Ridge National Laboratory, P.O. Box 2008, Bldg. 5700, MS-6170, Oak Ridge, TN 37831 (United States); Nutt, Mark; Morris, Edgar; Puig, Francesc [Argonne National Laboratory (United States); Carter, Joe; Delley, Alexcia; Rodwell, Phillip [Savannah River National Laboratory (United States); Hardin, Ernest; Kalinina, Elena [Sandia National Laboratories (United States); Clark, Robert [U.S. Department of Energy (United States); Cotton, Thomas [Complex Systems Group (United States)

2013-07-01

Preliminary system-level analyses of the interfaces between at-reactor used fuel management, consolidated storage facilities, and disposal facilities, along with the development of supporting logistics simulation tools, have been initiated to provide the U.S. Department of Energy (DOE) and other stakeholders with information regarding the various alternatives for managing used nuclear fuel (UNF) generated by the current fleet of light water reactors operating in the United States. An important UNF management system interface consideration is the need for ultimate disposal of UNF assemblies contained in waste packages that are sized to be compatible with different geologic media. Thermal analyses indicate that waste package sizes for the geologic media under consideration by the Used Fuel Disposition Campaign may be significantly smaller than the canisters being used for on-site dry storage by the nuclear utilities. Therefore, at some point along the UNF disposition pathway, there could be a need to repackage fuel assemblies already loaded and being loaded into the dry storage canisters currently in use. The implications of where and when the packaging or repackaging of commercial UNF will occur are key questions being addressed in this evaluation. The analysis demonstrated that thermal considerations will have a major impact on the operation of the system and that acceptance priority, rates, and facility start dates have significant system implications. (authors)

PLUTON, Isotope Generation and Depletion in Highly Irradiated LWR Fuel Rods

International Nuclear Information System (INIS)

Lemehov, Sergei; Motoe, Suzuki

2003-01-01

1 - Description of program or function: The PLUTON-PC is a three-group neutronic code analyzing, as functions of time and burnup, the change of radial profiles, together with average values, of power density, burnup, concentration of trans-uranium elements, plutonium buildup, depletion of fissile elements, and fission product generation in water reactor fuel rod with standard UO 2 , UO 2 -Gd 2 O 3 , inhomogeneous MOX, and UO 2 -ThO 2 . The PLUTON-PC code, which has been designed to be run on Windows PC, has adopted a theoretical shape function of neutron attenuation in pellet, which enables users to perform a very fast and accurate calculation easily. The code includes the irradiation conditions of the Halden Reactor which gives verification data for the code. Verification has been performed up to 83 GWd/tU, and a satisfactory agreement has been obtained. 2 - Methods: Based upon cumulative yields, the PLUTON-PC code calculates as a function of radial position and local burnup concentrations of fission products, macroscopic scattering cross-sections and self-shielding effect which is important for standard fuel (for Pu-242 mainly) and more importantly for homogeneous and inhomogeneous MOX fuel because of higher concentrations of fissile and fertile isotopes of plutonium. The code results in burnup dependent fission rate density profiles throughout the in-reactor irradiation of LWR fuel rods. The isotopes included in calculations have been extended to cover all trans-uranium groups (plutonium plus higher actinides) of fissile and fertile isotopes. Self-shielding problem and scattering effects have been revised and solved for all isotopes in the calculations for adequacy at high burnup, different irradiation conditions and cladding materials

Radiochemical analyses of several spent fuel Approved Testing Materials

International Nuclear Information System (INIS)

Guenther, R.J.; Blahnik, D.E.; Wildung, N.J.

1994-09-01

Radiochemical characterization data are described for UO 2 and UO 2 plus 3 wt% Gd 2 O 3 commercial spent nuclear fuel taken from a series of Approved Testing Materials (ATMs). These full-length nuclear fuel rods include MLA091 of ATM-103, MKP070 of ATM-104, NBD095 and NBD131 of ATM-106, and ADN0206 of ATM-108. ATMs 103, 104, and 106 were all irradiated in the Calvert Cliffs Nuclear Power Plant (Reactor No.1), a pressurized-water reactor that used fuel fabricated by Combustion Engineering. ATM-108 was part of the same fuel bundle designed as ATM-105 and came from boiling-water reactor fuel fabricated by General Electric and irradiated in the Cooper Nuclear Power Plant. Rod average burnups and expected fission gas releases ranged from 2,400 to 3,700 GJ/kgM. (25 to 40 Mwd/kgM) and from less than 1% to greater than 10%, respectively, depending on the specific ATM. The radiochemical analyses included uranium and plutonium isotopes in the fuel, selected fission products in the fuel, fuel burnup, cesium and iodine on the inner surfaces of the cladding, 14 C in the fuel and cladding, and analyses of the gases released to the rod plenum. Supporting examinations such as fuel rod design and material descriptions, power histories, and gamma scans used for sectioning diagrams are also included. These ATMs were examined as part of the Materials Characterization Center Program conducted at Pacific Northwest Laboratory provide a source of well-characterized spent fuel for testing in support of the US Department of Energy Office of Civilian Radioactive Waste Management Program

Parallel processing of neutron transport in fuel assembly calculation

International Nuclear Information System (INIS)

Song, Jae Seung

1992-02-01

Group constants, which are used for reactor analyses by nodal method, are generated by fuel assembly calculations based on the neutron transport theory, since one or a quarter of the fuel assembly corresponds to a unit mesh in the current nodal calculation. The group constant calculation for a fuel assembly is performed through spectrum calculations, a two-dimensional fuel assembly calculation, and depletion calculations. The purpose of this study is to develop a parallel algorithm to be used in a parallel processor for the fuel assembly calculation and the depletion calculations of the group constant generation. A serial program, which solves the neutron integral transport equation using the transmission probability method and the linear depletion equation, was prepared and verified by a benchmark calculation. Small changes from the serial program was enough to parallelize the depletion calculation which has inherent parallel characteristics. In the fuel assembly calculation, however, efficient parallelization is not simple and easy because of the many coupling parameters in the calculation and data communications among CPU's. In this study, the group distribution method is introduced for the parallel processing of the fuel assembly calculation to minimize the data communications. The parallel processing was performed on Quadputer with 4 CPU's operating in NURAD Lab. at KAIST. Efficiencies of 54.3 % and 78.0 % were obtained in the fuel assembly calculation and depletion calculation, respectively, which lead to the overall speedup of about 2.5. As a result, it is concluded that the computing time consumed for the group constant generation can be easily reduced by parallel processing on the parallel computer with small size CPU's

Management of depleted uranium

International Nuclear Information System (INIS)

2001-01-01

Large stocks of depleted uranium have arisen as a result of enrichment operations, especially in the United States and the Russian Federation. Countries with depleted uranium stocks are interested in assessing strategies for the use and management of depleted uranium. The choice of strategy depends on several factors, including government and business policy, alternative uses available, the economic value of the material, regulatory aspects and disposal options, and international market developments in the nuclear fuel cycle. This report presents the results of a depleted uranium study conducted by an expert group organised jointly by the OECD Nuclear Energy Agency and the International Atomic Energy Agency. It contains information on current inventories of depleted uranium, potential future arisings, long term management alternatives, peaceful use options and country programmes. In addition, it explores ideas for international collaboration and identifies key issues for governments and policy makers to consider. (authors)

POST-IRRADIATION ANALYSES OF U-MO DISPERSION FUEL RODS OF KOMO TESTS AT HANARO

Directory of Open Access Journals (Sweden)

H.J. RYU

2013-12-01

Full Text Available Since 2001, a series of five irradiation test campaigns for atomized U-Mo dispersion fuel rods, KOMO-1, -2, -3, -4, and -5, has been conducted at HANARO (Korea in order to develop high performance low enriched uranium dispersion fuel for research reactors. The KOMO irradiation tests provided valuable information on the irradiation behavior of U-Mo fuel that results from the distinct fuel design and irradiation conditions of the rod fuel for HANARO. Full size U-Mo dispersion fuel rods of 4–5 g-U/cm3 were irradiated at a maximum linear power of approximately 105 kW/m up to 85% of the initial U-235 depletion burnup without breakaway swelling or fuel cladding failure. Electron probe microanalyses of the irradiated samples showed localized distribution of the silicon that was added in the matrix during fuel fabrication and confirmed its beneficial effect on interaction layer growth during irradiation. The modifications of U-Mo fuel particles by the addition of a ternary alloying element (Ti or Zr, additional protective coatings (silicide or nitride, and the use of larger fuel particles resulted in significantly reduced interaction layers between fuel particles and Al.

Review of HEDL fuel pin transient analyses analytical programs

International Nuclear Information System (INIS)

Scott, J.H.; Baars, R.E.

1975-05-01

Methods for analysis of transient fuel pin performance are described, as represented by the steady-state SIEX code and the PECT series of codes used for steady-state and transient mechanical analyses. The empirical fuel failure correlation currently in use for analysis of transient overpower accidents is described. (U.S.)

DANDE: a linked code system for core neutronics/depletion analysis

International Nuclear Information System (INIS)

LaBauve, R.J.; England, T.R.; George, D.C.; MacFarlane, R.E.; Wilson, W.B.

1985-06-01

This report describes DANDE - a modular neutronics, depletion code system for reactor analysis. It consists of nuclear data processing, core physics, and fuel depletion modules, and allows one to use diffusion and transport methods interchangeably in core neutronics calculations. This latter capability is especially important in the design of small modular cores. Additional unique features include the capability of updating the nuclear data file during a calculation; a detailed treatment of depletion, burnable poisons as well as fuel; and the ability to make geometric changes such as control rod repositioning and fuel relocation in the course of a calculation. The detailed treatment of reactor fuel burnup, fission-product creation and decay, as well as inventories of higher-order actinides is a necessity when predicting the behavior of reactor fuel under increased burn conditions. The operation of the code system is made clear in this report by following a sample problem

DANDE: a linked code system for core neutronics/depletion analysis

International Nuclear Information System (INIS)

LaBauve, R.J.; England, T.R.; George, D.C.; MacFarlane, R.E.; Wilson, W.B.

1986-01-01

This report describes DANDE - a modular neutronics, depletion code system for reactor analysis. It consists of nuclear data processing, core physics, and fuel depletion modules, and allows one to use diffusion and transport methods interchangeably in core neutronics calculations. This latter capability is especially important in the design of small modular cores. Additional unique features include the capability of updating the nuclear data file during a calculation; a detailed treatment of depletion, burnable poisons as well as fuel; and the ability to make geometric changes such as control rod repositioning and fuel relocation in the cource of a calculation. The detailed treatment of reactor fuel burnup, fission-product creation and decay, as well as inventories of higher-order actinides is a necessity when predicting the behavior of reactor fuel under increased burn conditions. The operation of the code system is illustrated in this report by two sample problems. 25 refs

DANDE-a linked code system for core neutronics/depletion analysis

International Nuclear Information System (INIS)

LaBauve, R.J.; England, T.R.; George, D.C.; MacFarlane, R.E.; Wilson, W.B.

1986-01-01

This report describes DANDE-a modular neutronics, depletion code system for reactor analysis. It consists of nuclear data processing, core physics, and fuel depletion modules, and allows one to use diffusion and transport methods interchangeably in core neutronics calculations. This latter capability is especially important in the design of small modular cores. Additional unique features include the capability of updating the nuclear data file during a calculation; a detailed treatment of depletion, burnable poisons as well as fuel; and the ability to make geometric changes such as control rod repositioning and fuel relocation in the course of a calculation. The detailed treatment of reactor fuel burnup, fission-product creation and decay, as well as inventories of higher-order actinides is a necessity when predicting the behavior of the reactor fuel under increased burn conditions. The operation of the code system is illustrated in this report by two actual problems

Comparison of scale/triton and helios burnup calculations for high burnup LWR fuel

Energy Technology Data Exchange (ETDEWEB)

Tittelbach, S.; Mispagel, T.; Phlippen, P.W. [WTI Wissenschaftlich-Technische Ingenieurberatung GmbH, Juelich (Germany)

2009-07-01

The presented analyses provide information about the suitability of the lattice burnup code HELIOS and the recently developed code SCALE/TRITON for the prediction of isotopic compositions of high burnup LWR fuel. The accurate prediction of the isotopic inventory of high burnt spent fuel is a prerequisite for safety analyses in and outside of the reactor core, safe loading of spent fuel into storage casks, design of next generation spent fuel casks and for any consideration of burnup credit. Depletion analyses are performed with both burnup codes for PWR and BWR fuel samples which were irradiated far beyond 50 GWd/t within the LWR-PROTEUS Phase II project. (orig.)

Citham a computer code for calculating fuel depletion-description, tests, modifications and evaluation

International Nuclear Information System (INIS)

Alvarenga, M.A.B.

1984-12-01

The CITHAN computer code was developed at IPEN (Instituto de Pesquisas Energeticas e Nucleares) to link the HAMMER computer code with a fuel depletion routine and to provide neutron cross sections to be read with the appropriate format of the CITATION code. The problem arised due to the efforts to addapt the new version denomined HAMMER-TECHION with the routine refered. The HAMMER-TECHION computer code was elaborated by Haifa Institute, Israel within a project with EPRI. This version is at CNEN to be used in multigroup constant generation for neutron diffusion calculation in the scope of the new methodology to be adopted by CNEN. The theoretical formulation of CITHAM computer code, tests and modificatins are described. (Author) [pt

Relationship between basic nuclear data and LWR fuel cycle parameters

International Nuclear Information System (INIS)

Becker, M.; Harris, D.R.; Quan, B.; Ryskamp, J.M.

1979-01-01

An interactive system has been developed at RPI to analyze the sensitivity of water reactor fuel cycle parameters and costs to uncertainties in nuclear data. A sequence of batch depletion, core analysis, and fuel cost codes (referred to as Path B) determines the changes in fuel cycle parameters and costs for changes in few-group microscopic cross sections, in fission yields, and in decay data. For cases that are found to be significant from Part B analysis, the sensitivities of few-group data to basic nuclear data are determined by detailed calculations (referred to as Path A). Analyses of pressurized and boiling water reactors with recycle and throwaway options show substantial sensitivities of fuel cycle parameters and costs, particularly to thermal and resonance nuclear data for fissile nuclides. The results bring out the importance for power reactor sensitivity analysis of dealing with the full fuel cycle including depletion of initially-loaded fuel and the building-in of actinides and fission products

Parametric neutronic analyses related to burnup credit cask design

International Nuclear Information System (INIS)

Parks, C.V.

1989-01-01

The consideration of spent fuel histories (burnup credit) in the design of spent fuel shipping casks will result in cost savings and public risk benefits in the overall fuel transportation system. The purpose of this paper is to describe the depletion and criticality analyses performed in conjunction with and supplemental to the referenced analysis. Specifically, the objectives are to indicate trends in spent fuel isotopic composition with burnup and decay time; provide spent fuel pin lattice values as a function of burnup, decay time, and initial enrichment; demonstrate the variation of k eff for infinite arrays of spent fuel assemblies separated by generic cask basket designs (borated and unborated) of varying thicknesses; and verify the potential cask reactivity margin available with burnup credit via analysis with generic cask models

Generalized perturbation theory for LWR depletion analysis and core design applications

International Nuclear Information System (INIS)

White, J.R.; Frank, B.R.

1986-01-01

A comprehensive time-dependent perturbation theory formulation that includes macroscopic depletion, thermal-hydraulic and poison feedback effects, and a criticality reset mechanism is developed. The methodology is compatible with most current LWR design codes. This new development allows GTP/DTP methods to be used quantitatively in a variety of realistic LWR physics applications that were not possible prior to this work. A GTP-based optimization technique for incore fuel management analyses is addressed as a promising application of the new formulation

Comparison of KANEXT and SERPENT for fuel depletion calculations of a sodium fast reactor

International Nuclear Information System (INIS)

Lopez-Solis, R.C.; Francois, J.L.; Becker, M.; Sanchez-Espinoza, V.H.

2014-01-01

As most of Generation-IV systems are in development, efficient and reliable computational tools are needed to obtain accurate results in reasonably computer time. In this study, KANEXT code system is presented and validated against the well-known Monte Carlo SERPENT code, for fuel depletion calculations of a sodium fast reactor (SFR). The KArlsruhe Neutronic EXtended Tool (KANEXT) is a modular code system for deterministic reactor calculations, consisting of one kernel and several modules. Results obtained with KANEXT for the SFR core are in good agreement with the ones of SERPENT, e.g. the neutron multiplication factor and the isotopes evolution with burn-up. (author)

Depletion optimization of lumped burnable poisons in pressurized water reactors

International Nuclear Information System (INIS)

Kodah, Z.H.

1982-01-01

Techniques were developed to construct a set of basic poison depletion curves which deplete in a monotonical manner. These curves were combined to match a required optimized depletion profile by utilizing either linear or non-linear programming methods. Three computer codes, LEOPARD, XSDRN, and EXTERMINATOR-2 were used in the analyses. A depletion routine was developed and incorporated into the XSDRN code to allow the depletion of fuel, fission products, and burnable poisons. The Three Mile Island Unit-1 reactor core was used in this work as a typical PWR core. Two fundamental burnable poison rod designs were studied. They are a solid cylindrical poison rod and an annular cylindrical poison rod with water filling the central region.These two designs have either a uniform mixture of burnable poisons or lumped spheroids of burnable poisons in the poison region. Boron and gadolinium are the two burnable poisons which were investigated in this project. Thermal self-shielding factor calculations for solid and annular poison rods were conducted. Also expressions for overall thermal self-shielding factors for one or more than one size group of poison spheroids inside solid and annular poison rods were derived and studied. Poison spheroids deplete at a slower rate than the poison mixture because each spheroid exhibits some self-shielding effects of its own. The larger the spheroid, the higher the self-shielding effects due to the increase in poison concentration

Numerical analyses of an ex-core fuel incident: Results of the OECD-IAEA Paks Fuel Project

Energy Technology Data Exchange (ETDEWEB)

Hozer, Z., E-mail: hozer@aeki.kfki.h [Hungarian Academy of Sciences KFKI Atomic Energy Research Institute, H-1525 Budapest, P.O. Box 49 (Hungary); Aszodi, A. [BME NTI Budapest (Hungary); Barnak, M. [IVS, Trnava (Slovakia); Boros, I. [BME NTI Budapest (Hungary); Fogel, M. [VUJE, Trnava (Slovakia); Guillard, V. [IRSN, Cadarache (France); Gyori, Cs. [ITU, EU, Karlsruhe (Germany); Hegyi, G. [Hungarian Academy of Sciences KFKI Atomic Energy Research Institute, H-1525 Budapest, P.O. Box 49 (Hungary); Horvath, G.L. [VEIKI, Budapest (Hungary); Nagy, I. [Hungarian Academy of Sciences KFKI Atomic Energy Research Institute, H-1525 Budapest, P.O. Box 49 (Hungary); Junninen, P. [VTT, Espoo (Finland); Kobzar, V. [KI, Moscow (Russian Federation); Legradi, G. [BME NTI Budapest (Hungary); Molnar, A. [Hungarian Academy of Sciences KFKI Atomic Energy Research Institute, H-1525 Budapest, P.O. Box 49 (Hungary); Pietarinen, K. [VTT, Espoo (Finland); Perneczky, L. [Hungarian Academy of Sciences KFKI Atomic Energy Research Institute, H-1525 Budapest, P.O. Box 49 (Hungary); Makihara, Y. [ATMEA, Paris (France); Matejovic, P. [IVS, Trnava (Slovakia); Perez-Fero, E.; Slonszki, E. [Hungarian Academy of Sciences KFKI Atomic Energy Research Institute, H-1525 Budapest, P.O. Box 49 (Hungary)

2010-03-15

The OECD-IAEA Paks Fuel Project was developed to support the understanding of fuel behaviour in accident conditions on the basis of analyses of the Paks-2 incident. Numerical simulation of the most relevant aspects of the event and comparison of the calculation results with the available data from the incident was carried out between 2006 and 2007. A database was compiled to provide input for the code calculations. The activities covered the following three areas: (a) Thermal hydraulic calculations described the cooling conditions possibly established during the incident. (b) Simulation of fuel behaviour described the oxidation and degradation mechanisms of the fuel assemblies. (c) The release of fission products from the failed fuel rods was estimated and compared to available measured data. The applied used codes captured the most important events of the Paks-2 incident and the calculated results improved the understanding of the causes and mechanisms of fuel failure. The numerical analyses showed that the by-pass flow leading to insufficient cooling amounted to 75-90% of the inlet flow rate, the maximum temperature in the tank was between 1200 and 1400 deg. C, the degree of zirconium oxidation reached 4-12% and the mass of produced hydrogen was between 3 and 13 kg.

Automatic optimized reload and depletion method for a pressurized water reactor

International Nuclear Information System (INIS)

Ahn, D.H.; Levene, S.H.

1985-01-01

A new method has been developed to automatically reload and deplete a pressurized water reactor (PWR) so that both the enriched inventory requirements during the reactor cycle and the cost of reloading the core are minimized. This is achieved through four stepwise optimization calculations: (a) determination of the minimum fuel requirement for an equivalent three-region core model, (b) optimal selection and allocation of fuel assemblies for each of the three regions to minimize the reload cost, (c) optimal placement of fuel assemblies to conserve regionwise optimal conditions, and (d) optimal control through poison management to deplete individual fuel assemblies to maximize end-of-cycle k /SUB eff/ . The new method differs from previous methods in that the optimization process automatically performs all tasks required to reload and deplete a PWR. In addition, the previous work that developed optimization methods principally for the initial reactor cycle was modified to handle subsequent cycles with fuel assemblies having burnup at beginning of cycle. Application of the method to the fourth reactor cycle at Three Mile Island Unit 1 has shown that both the enrichment and the number of fresh reload fuel assemblies can be decreased and fully amortized fuel assemblies can be reused to minimize the fuel cost of the reactor

Theoretical analysis of nuclear reactors (Phase I), I-V, Part IV, Nuclear fuel depletion; Razrada metoda teorijske analize nuklearnih reaktora (I faza) I-V, IV Deo, Promena izotopnog sastava goriva

Energy Technology Data Exchange (ETDEWEB)

Pop-Jordanov, J [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

1962-07-15

Nuclear fuel depletion is analyzed in order to estimate the qualitative and quantitative fuel property changes during irradiation and the influence of changes on the reactivity during long-term reactor operation. The changes of fuel properties are described by changes of neutron absorption and fission cross sections. Part one of this report covers the economic significance of fuel burnup and the review of fuel isotopic changes during depletion. Pat two contains the analysis of the U{sup 235} chain, analytical expressions for the concentrations of U{sup 235}, U{sup 236} and Np{sup 237} as a function of burnup. Part three contains the analysis of neutron spectrum influence on the Westcott method for calculating the cross sections. Part four contains the calculation method applied on Calder Hall type reactor. The results were obtained by applying ZUSE-22 R digital computer.

ORIGEN-based Nuclear Fuel Inventory Module for Fuel Cycle Assessment: Final Project Report

Energy Technology Data Exchange (ETDEWEB)

Skutnik, Steven E. [Univ. of Tennessee, Knoxville, TN (United States). Dept. of Nuclear Engineering

2017-06-19

The goal of this project, “ORIGEN-based Nuclear Fuel Depletion Module for Fuel Cycle Assessment" is to create a physics-based reactor depletion and decay module for the Cyclus nuclear fuel cycle simulator in order to assess nuclear fuel inventories over a broad space of reactor operating conditions. The overall goal of this approach is to facilitate evaluations of nuclear fuel inventories for a broad space of scenarios, including extended used nuclear fuel storage and cascading impacts on fuel cycle options such as actinide recovery in used nuclear fuel, particularly for multiple recycle scenarios. The advantages of a physics-based approach (compared to a recipe-based approach which has been typically employed for fuel cycle simulators) is in its inherent flexibility; such an approach can more readily accommodate the broad space of potential isotopic vectors that may be encountered under advanced fuel cycle options. In order to develop this flexible reactor analysis capability, we are leveraging the Origen nuclear fuel depletion and decay module from SCALE to produce a standalone “depletion engine” which will serve as the kernel of a Cyclus-based reactor analysis module. The ORIGEN depletion module is a rigorously benchmarked and extensively validated tool for nuclear fuel analysis and thus its incorporation into the Cyclus framework can bring these capabilities to bear on the problem of evaluating long-term impacts of fuel cycle option choices on relevant metrics of interest, including materials inventories and availability (for multiple recycle scenarios), long-term waste management and repository impacts, etc. Developing this Origen-based analysis capability for Cyclus requires the refinement of the Origen analysis sequence to the point where it can reasonably be compiled as a standalone sequence outside of SCALE; i.e., wherein all of the computational aspects of Origen (including reactor cross-section library processing and interpolation, input and output

Benefits of the delta K of depletion benchmarks for burnup credit validation

International Nuclear Information System (INIS)

Lancaster, D.; Machiels, A.

2012-01-01

Pressurized Water Reactor (PWR) burnup credit validation is demonstrated using the benchmarks for quantifying fuel reactivity decrements, published as 'Benchmarks for Quantifying Fuel Reactivity Depletion Uncertainty,' EPRI Report 1022909 (August 2011). This demonstration uses the depletion module TRITON available in the SCALE 6.1 code system followed by criticality calculations using KENO-Va. The difference between the predicted depletion reactivity and the benchmark's depletion reactivity is a bias for the criticality calculations. The uncertainty in the benchmarks is the depletion reactivity uncertainty. This depletion bias and uncertainty is used with the bias and uncertainty from fresh UO 2 critical experiments to determine the criticality safety limits on the neutron multiplication factor, k eff . The analysis shows that SCALE 6.1 with the ENDF/B-VII 238-group cross section library supports the use of a depletion bias of only 0.0015 in delta k if cooling is ignored and 0.0025 if cooling is credited. The uncertainty in the depletion bias is 0.0064. Reliance on the ENDF/B V cross section library produces much larger disagreement with the benchmarks. The analysis covers numerous combinations of depletion and criticality options. In all cases, the historical uncertainty of 5% of the delta k of depletion ('Kopp memo') was shown to be conservative for fuel with more than 30 GWD/MTU burnup. Since this historically assumed burnup uncertainty is not a function of burnup, the Kopp memo's recommended bias and uncertainty may be exceeded at low burnups, but its absolute magnitude is small. (authors)

The development of depletion program coupled with Monte Carlo computer code

International Nuclear Information System (INIS)

Nguyen Kien Cuong; Huynh Ton Nghiem; Vuong Huu Tan

2015-01-01

The paper presents the development of depletion code for light water reactor coupled with MCNP5 code called the MCDL code (Monte Carlo Depletion for Light Water Reactor). The first order differential depletion system equations of 21 actinide isotopes and 50 fission product isotopes are solved by the Radau IIA Implicit Runge Kutta (IRK) method after receiving neutron flux, reaction rates in one group energy and multiplication factors for fuel pin, fuel assembly or whole reactor core from the calculation results of the MCNP5 code. The calculation for beryllium poisoning and cooling time is also integrated in the code. To verify and validate the MCDL code, high enriched uranium (HEU) and low enriched uranium (LEU) fuel assemblies VVR-M2 types and 89 fresh HEU fuel assemblies, 92 LEU fresh fuel assemblies cores of the Dalat Nuclear Research Reactor (DNRR) have been investigated and compared with the results calculated by the SRAC code and the MCNP R EBUS linkage system code. The results show good agreement between calculated data of the MCDL code and reference codes. (author)

Validation and application of a physics database for fast reactor fuel cycle analysis

International Nuclear Information System (INIS)

McKnight, R.D.; Stillman, J.A.; Toppel, B.J.; Khalil, H.S.

1994-01-01

An effort has been made to automate the execution of fast reactor fuel cycle analysis, using EBR-II as a demonstration vehicle, and to validate the analysis results for application to the IFR closed fuel cycle demonstration at EBR-II and its fuel cycle facility. This effort has included: (1) the application of the standard ANL depletion codes to perform core-follow analyses for an extensive series of EBR-II runs, (2) incorporation of the EBR-II data into a physics database, (3) development and verification of software to update, maintain and verify the database files, (4) development and validation of fuel cycle models and methodology, (5) development and verification of software which utilizes this physics database to automate the application of the ANL depletion codes, methods and models to perform the core-follow analysis, and (6) validation studies of the ANL depletion codes and of their application in support of anticipated near-term operations in EBR-II and the Fuel Cycle Facility. Results of the validation tests indicate the physics database and associated analysis codes and procedures are adequate to predict required quantities in support of early phases of FCF operations

Fuel cells:

DEFF Research Database (Denmark)

Sørensen, Bent

2013-01-01

A brief overview of the progress in fuel cell applications and basic technology development is presented, as a backdrop for discussing readiness for penetration into the marketplace as a solution to problems of depletion, safety, climate or environmental impact from currently used fossil and nucl......A brief overview of the progress in fuel cell applications and basic technology development is presented, as a backdrop for discussing readiness for penetration into the marketplace as a solution to problems of depletion, safety, climate or environmental impact from currently used fossil...... and nuclear fuel-based energy technologies....

ORIGEN-S: scale system module to calculate fuel depletion, actinide transmutation, fission product buildup and decay, and associated radiation source terms

International Nuclear Information System (INIS)

Anon.

1982-01-01

ORIGEN-S computes time-dependent concentrations and source terms of a large number of isotopes, which are simultaneously generated or depleted through neutronic transmutation, fission, radioactive decay, input feet rates and physical or chemical removal rates. The calculations may pertain to fuel irradiation within nuclear reactors, or the storage, management, transportation or subsequent chemical processing of removed fuel elements. The matrix exponential expansion model of the ORIGIN code is unaltered in ORIGEN-S. Essentially all features of ORIGEN were retained, expanded or supplemented within new computations. The primary objective of ORIGEN-S, as requested by the Nuclear Regulatory Commission, is that the calculations may utilize the multi-energy group cross sections from any currently processed standardized ENDF/B data base. This purpose has been implemented through the prior execution of codes within either the SCALE System or the AMPX System, developed at the Oak Ridge National Laboratory. These codes compute flux-weighted cross sections, simulating conditions within any given reactor fuel assembly, and convert the data into a library that can be input to ORIGEN-S. Time-dependent libraries may be produced, reflecting fuel composition variations during irradiation. Presented in the document are: detailed and condensed input instructions, model theory, features available, range of applicability, brief subroutine descriptions, sample input, and I/O requirements. Presently the code is operable on IBM 360/370 computers and may be converted for CDC computers. ORIGEN-S is a functional module in the SCALE System and will be one of the modules invoked in the SAS2 Control Module, presently being developed, or may be applied as a stand alone program. It can be used in nuclear reactor and processing plant design studies, radiation safety analyses, and environmental assessments

Depletion benchmarks calculation of random media using explicit modeling approach of RMC

International Nuclear Information System (INIS)

Liu, Shichang; She, Ding; Liang, Jin-gang; Wang, Kan

2016-01-01

Highlights: • Explicit modeling of RMC is applied to depletion benchmark for HTGR fuel element. • Explicit modeling can provide detailed burnup distribution and burnup heterogeneity. • The results would serve as a supplement for the HTGR fuel depletion benchmark. • The method of adjacent burnup regions combination is proposed for full-core problems. • The combination method can reduce memory footprint, keeping the computing accuracy. - Abstract: Monte Carlo method plays an important role in accurate simulation of random media, owing to its advantages of the flexible geometry modeling and the use of continuous-energy nuclear cross sections. Three stochastic geometry modeling methods including Random Lattice Method, Chord Length Sampling and explicit modeling approach with mesh acceleration technique, have been implemented in RMC to simulate the particle transport in the dispersed fuels, in which the explicit modeling method is regarded as the best choice. In this paper, the explicit modeling method is applied to the depletion benchmark for HTGR fuel element, and the method of combination of adjacent burnup regions has been proposed and investigated. The results show that the explicit modeling can provide detailed burnup distribution of individual TRISO particles, and this work would serve as a supplement for the HTGR fuel depletion benchmark calculations. The combination of adjacent burnup regions can effectively reduce the memory footprint while keeping the computational accuracy.

Analyses of the transportation of spent research reactor fuel in the United States

International Nuclear Information System (INIS)

Cashwell, J.W.; Neuhauser, K.S.

1989-01-01

We analyzed the impacts of transportation of research reactor spent fuel from US and foreign reactors for the US Department of Energy's (DOE) Office of Defense Programs. Two separate shipment programs were analyzed. The shipment of research reactor spent fuel from Taiwan to the US (Fuel Movement Program), and the return of research reactor spent fuels of US origin from foreign and domestic reactors (Research Reactor Fuel Return Program). To perform these analyses, a comprehensive methodology for analyzing the probabilities and consequences of transportation in coastal waters and port facilities, handling at the port, and shipment by truck to reprocessing facilities was developed. The Taiwanese fuel consists of low-burnup aluminum-clad metallic uranium research reactor spent fuel; the other fuels are primarily aluminum-clad oxide fuels. The Fuel Movement Program is ongoing, while the Fuel Return Program addresses future shipments over a ten-year period. The operational aspects of the Taiwanese shipments have been uniform, but several possible shipping configurations are possible for the Fuel Return Program shipments. The risks of transporting spent nuclear fuel and other radioactive materials by all modes have been analyzed extensively. Comprehensive assessments, which bound the impacts of spent fuel transport, demonstrate that when shipments are made in compliance with applicable regulations, the risks for all such transport are low. For comparison with previously licensed transport activities and to provide continuity with earlier analyses, the results for shipment of 150-day-old commercial pressurized water reactor (PWR) spent fuel are presented as part of this study

Fuel element transfer cask modelling using MCNP technique

International Nuclear Information System (INIS)

Rosli Darmawan

2009-01-01

Full text: After operating for more than 25 years, some of the Reaktor TRIGA PUSPATI (RTP) fuel elements would have been depleted. A few addition and fuel reconfiguration exercises have to be conducted in order to maintain RTP capacity. Presently, RTP spent fuels are stored at the storage area inside RTP tank. The need to transfer the fuel element outside of RTP tank may be prevalence in the near future. The preparation shall be started from now. A fuel element transfer cask has been designed according to the recommendation by the fuel manufacturer and experience of other countries. A modelling using MCNP code has been conducted to analyse the design. The result shows that the design of transfer cask fuel element is safe for handling outside the RTP tank according to recent regulatory requirement. (author)

Fuel Element Transfer Cask Modelling Using MCNP Technique

International Nuclear Information System (INIS)

Darmawan, Rosli; Topah, Budiman Naim

2010-01-01

After operating for more than 25 years, some of the Reaktor TRIGA Puspati (RTP) fuel elements would have been depleted. A few addition and fuel reconfiguration exercises have to be conducted in order to maintain RTP capacity. Presently, RTP spent fuels are stored at the storage area inside RTP tank. The need to transfer the fuel element outside of RTP tank may be prevalence in the near future. The preparation shall be started from now. A fuel element transfer cask has been designed according to the recommendation by the fuel manufacturer and experience of other countries. A modelling using MCNP code has been conducted to analyse the design. The result shows that the design of transfer cask fuel element is safe for handling outside the RTP tank according to recent regulatory requirement.

Coupling of channel thermalhydraulics and fuel behaviour in ACR-1000 safety analyses

International Nuclear Information System (INIS)

Huang, F.L.; Lei, Q.M.; Zhu, W.; Bilanovic, Z.

2008-01-01

Channel thermalhydraulics and fuel thermal-mechanical behaviour are interlinked. This paper describes a channel thermalhydraulics and fuel behaviour coupling methodology that has been used in ACR-1000 safety analyses. The coupling is done for all 12 fuel bundles in a fuel channel using the channel thermalhydraulics code CATHENA MOD-3.5d/Rev 2 and the transient fuel behaviour code ELOCA 2.2. The coupling approach can be used for every fuel element or every group of fuel elements in the channel. Test cases are presented where a total of 108 fuel element models are set up to allow a full coupling between channel thermalhydraulics and detailed fuel analysis for a channel containing a string of 12 fuel bundles. An additional advantage of this coupling approach is that there is no need for a separate detailed fuel analysis because the coupling analysis, once done, provides detailed calculations for the fuel channel (fuel bundles, pressure tube, and calandria tube) as well as all the fuel elements (or element groups) in the channel. (author)

Performance assessment analyses unique to Department of Energy spent nuclear fuel

International Nuclear Information System (INIS)

Loo, H.H.; Duguid, J.J.

2000-01-01

This paper describes the iterative process of grouping and performance assessment that has led to the current grouping of the U.S. Department of Energy (DOE) spent nuclear fuel (SNF). The unique sensitivity analyses that form the basis for incorporating DOE fuel into the total system performance assessment (TSPA) base case model are described. In addition, the chemistry that results from dissolution of DOE fuel and high level waste (HLW) glass in a failed co-disposal package, and the effects of disposal of selected DOE SNF in high integrity cans are presented

Analyses of the transportation of spent research reactor fuel in the United States

International Nuclear Information System (INIS)

Cashwell, J.W.; Neuhauser, K.S.

1989-01-01

The Transportation Technology Center at Sandia National Laboratories has analyzed the impacts of transportation of research reactor spent fuel from US and foreign reactors for the US Department of Energy (DOE) Office of Defense Programs. This effort represents the first comprehensive analytical evaluation of the risks of transporting high-, medium-, and low-enriched uranium spent research reactor fuel by both sea and land. Two separate shipment programs have been analyzed: the shipment of research reactor spent fuel from Taiwan to the US (Fuel Movement Program), and the return of research reactor spent fuels of US origin from foreign and domestic reactors (Research Reactor Fuel Return Program). In order to perform these analyses, a comprehensive methodology for analyzing the probabilities and consequences of transportation in coastal waters and port facilities, handling at the port, and shipment by truck to reprocessing facilities was developed. The Taiwanese fuel consists of low-burnup aluminum-clad metallic uranium research reactor spent fuel; the other fuels are primarily aluminum-clad oxide fuels. The Fuel Movement Program is ongoing, while the Fuel Return Program addresses future shipments over a ten-year period. The operational aspects of the Taiwanese shipments have been uniform, but several possible shipping configurations are possible for the Fuel Return Program shipments. Comprehensive assessments, which bound the impacts of spent fuel transport, demonstrate that when shipments are made in compliance with applicable regulations, the risks for all such transport are low. For comparison with previously licensed transport activities and to provide continuity with earlier analyses, the results for shipment of 150-day-old commercial pressurized water reactor (PWR) spent fuel are presented as part of this study

Used fuel management system architecture and interface analyses

Energy Technology Data Exchange (ETDEWEB)

Nutt, Mark [Argonne National Laboratory, Argonne, IL (United States); Howard, Robert; Busch, Ingrid [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Carter, Joe; Delley, Alexcia [Savannah River National Laboratory, Aiken, SC (United States); Hardin, Ernest; Kalinina, Elena [Sandia National Laboratories, Albuquerque NM (United States); Cotton, Thomas [Complex Systems LLC, Washington, DC (United States)

2013-07-01

between at-reactor used fuel management, consolidated storage facilities, and disposal facilities, along with the development of supporting logistics simulation tools, have been initiated to provide the U.S. Department of Energy (DOE) and other stakeholders with information regarding the various alternatives for managing used nuclear fuel (UNF) generated by the current fleet of light water reactors operating in the United States. An important UNF management system interface consideration is the need for ultimate disposal of UNF assemblies contained in waste packages that are sized to be compatible with different geologic media. Thermal analyses indicate that waste package sizes for the geologic media under consideration by the Used Fuel Disposition Campaign may be significantly smaller than the canisters being used for on-site dry storage by the nuclear utilities. Therefore, at some point along the UNF disposition pathway, there could be a need to repackage fuel assemblies already loaded and being loaded into the dry storage canisters currently in use. The implications of where and when the packaging or repackaging of commercial UNF will occur are key questions being addressed in this evaluation. The analysis demonstrated that thermal considerations will have a major impact on the operation of the system and that acceptance priority, rates, and facility start dates have significant system implications. (authors)

Used fuel management system architecture and interface analyses

International Nuclear Information System (INIS)

Nutt, Mark; Howard, Robert; Busch, Ingrid; Carter, Joe; Delley, Alexcia; Hardin, Ernest; Kalinina, Elena; Cotton, Thomas

2013-01-01

between at-reactor used fuel management, consolidated storage facilities, and disposal facilities, along with the development of supporting logistics simulation tools, have been initiated to provide the U.S. Department of Energy (DOE) and other stakeholders with information regarding the various alternatives for managing used nuclear fuel (UNF) generated by the current fleet of light water reactors operating in the United States. An important UNF management system interface consideration is the need for ultimate disposal of UNF assemblies contained in waste packages that are sized to be compatible with different geologic media. Thermal analyses indicate that waste package sizes for the geologic media under consideration by the Used Fuel Disposition Campaign may be significantly smaller than the canisters being used for on-site dry storage by the nuclear utilities. Therefore, at some point along the UNF disposition pathway, there could be a need to repackage fuel assemblies already loaded and being loaded into the dry storage canisters currently in use. The implications of where and when the packaging or repackaging of commercial UNF will occur are key questions being addressed in this evaluation. The analysis demonstrated that thermal considerations will have a major impact on the operation of the system and that acceptance priority, rates, and facility start dates have significant system implications. (authors)

An Integrated Fuel Depletion Calculator for Fuel Cycle Options Analysis

Energy Technology Data Exchange (ETDEWEB)

Schneider, Erich [Univ. of Texas, Austin, TX (United States); Scopatz, Anthony [Univ. of Wisconsin, Madison, WI (United States)

2016-04-25

Bright-lite is a reactor modeling software developed at the University of Texas Austin to expand upon the work done with the Bright [1] reactor modeling software. Originally, bright-lite was designed to function as a standalone reactor modeling software. However, this aim was refocused t couple bright-lite with the Cyclus fuel cycle simulator [2] to make it a module for the fuel cycle simulator.

Evaluation of spent fuel isotopics, radiation spectra and decay heat using the scale computational system

International Nuclear Information System (INIS)

Parks, C.V.; Hermann, O.W.; Ryman, J.C.

1986-01-01

In order to be a self-sufficient system for transport/storage cask shielding and heat transfer analysis, the SCALE system developers included modules to evaluate spent fuel radiation spectra and decay heat. The primary module developed for these analyses is ORIGEN-S which is an updated verision of the original ORIGEN code. The COUPLE module was also developed to enable ORIGEN-S to easily utilize multigroup cross sections and neutron flux data during a depletion analysis. Finally, the SAS2 control module was developed for automating the depletion and decay via ORIGEN-S while using burnup-dependent neutronic data based on a user-specified fuel assembly and reactor history. The ORIGEN-S data libraries available for depletion and decay have also been significantly updated from that developed with the original ORIGEN code

Probabilistic fuel rod analyses using the TRANSURANUS code

Energy Technology Data Exchange (ETDEWEB)

Lassmann, K; O` Carroll, C; Laar, J Van De [CEC Joint Research Centre, Karlsruhe (Germany)

1997-08-01

After more than 25 years of fuel rod modelling research, the basic concepts are well established and the limitations of the specific approaches are known. However, the widely used mechanistic approach leads in many cases to discrepancies between theoretical predictions and experimental evidence indicating that models are not exact and that some of the physical processes encountered are of stochastic nature. To better understand uncertainties and their consequences, the mechanistic approach must therefore be augmented by statistical analyses. In the present paper the basic probabilistic methods are briefly discussed. Two such probabilistic approaches are included in the fuel rod performance code TRANSURANUS: the Monte Carlo method and the Numerical Noise Analysis. These two techniques are compared and their capabilities are demonstrated. (author). 12 refs, 4 figs, 2 tabs.

Spent fuel shipping costs for transportation logistics analyses

International Nuclear Information System (INIS)

Cole, B.M.; Cross, R.E.; Cashwell, J.W.

1983-05-01

Logistics analyses supplied to the nuclear waste management programs of the U.S. Department of Energy through the Transportation Technology Center (TTC) at Sandia National Laboratories are used to predict nuclear waste material logistics, transportation packaging demands, shipping and receiving rates and transportation-related costs for alternative strategies. This study is an in-depth analysis of the problems and contingencies associated with the costs of shipping irradiated reactor fuel. These costs are extremely variable however, and have changed frequently (sometimes monthly) during the past few years due to changes in capital, fuel, and labor costs. All costs and charges reported in this study are based on January 1982 data using existing transport cask systems and should be used as relative indices only. Actual shipping costs would be negotiable for each origin-destination combination

SFCOMPO: A new database of isotopic compositions of spent nuclear fuel

International Nuclear Information System (INIS)

Michel-Sendis, Franco; Gauld, Ian

2014-01-01

The numerous applications of nuclear fuel depletion simulations impact all areas related to nuclear safety. They are at the basis of, inter alia, spent fuel criticality safety analyses, reactor physics calculations, burn-up credit methodologies, decay heat thermal analyses, radiation shielding, reprocessing, waste management, deep geological repository safety studies and safeguards. Experimentally determined nuclide compositions of well-characterised spent nuclear fuel (SNF) samples are used to validate the accuracy of depletion code predictions for a given burn-up. At the same time, the measured nuclide composition of the sample is used to determine the burn-up of the fuel. It is therefore essential to have a reliable and well-qualified database of measured nuclide concentrations and relevant reactor operational data that can be used as experimental benchmark data for depletion codes and associated nuclear data. The Spent Fuel Isotopic Composition Database (SFCOMPO) has been hosted by the NEA since 2001. In 2012, a collaborative effort led by the NEA Data Bank and Oak Ridge National Laboratory (ORNL) in the United States, under the guidance of the NEA Expert Group on Assay Data of Spent Nuclear Fuel (EGADSNF) of the Working Party on Nuclear Criticality Safety (WPNCS), has resulted in the creation of an enhanced relational database structure and a significant expansion of the SFCOMPO database, which now contains experimental assay data for a wider selection of international reactor designs. The new database was released online in 2014. This new SFCOMPO database aims to provide access to open experimental SNF assay data to ensure their preservation and to facilitate their qualification as evaluated assay data suitable for the validation of methodologies used to predict the composition of irradiated nuclear fuel. Having a centralised, internationally reviewed database that makes these data openly available for a large selection of international reactor designs is of

Ecological considerations of natural and depleted uranium

International Nuclear Information System (INIS)

Hanson, W.C.

1980-01-01

Depleted 238 U is a major by-product of the nuclear fuel cycle for which increasing use is being made in counterweights, radiation shielding, and ordnance applications. This paper (1) summarizes the pertinent literature on natural and depleted uranium in the environment, (2) integrates results of a series of ecological studies conducted at Los Alamos Scientific Laboratory (LASL) in New Mexico where 70,000 kg of depleted and natural uranium has been expended to the environment over the past 34 years, and (3) synthesizes the information into an assessment of the ecological consequences of natural and depleted uranium released to the environment by various means. Results of studies of soil, plant, and animal communities exposed to this radiation and chemical environment over a third of a century provide a means of evaluating the behavior and effects of uranium in many contexts

Constant strength fuel-fuel cell

International Nuclear Information System (INIS)

Vaseen, V.A.

1980-01-01

A fuel cell is an electrochemical apparatus composed of both a nonconsumable anode and cathode; and electrolyte, fuel oxidant and controls. This invention guarantees the constant transfer of hydrogen atoms and their respective electrons, thus a constant flow of power by submergence of the negative electrode in a constant strength hydrogen furnishing fuel; when said fuel is an aqueous absorbed hydrocarbon, such as and similar to ethanol or methnol. The objective is accomplished by recirculation of the liquid fuel, as depleted in the cell through specific type membranes which pass water molecules and reject the fuel molecules; thus concentrating them for recycle use

Depleted uranium hexafluoride: Waste or resource?

International Nuclear Information System (INIS)

Schwertz, N.; Zoller, J.; Rosen, R.; Patton, S.; Bradley, C.; Murray, A.

1995-07-01

The US Department of Energy is evaluating technologies for the storage, disposal, or re-use of depleted uranium hexafluoride (UF 6 ). This paper discusses the following options, and provides a technology assessment for each one: (1) conversion to UO 2 for use as mixed oxide duel, (2) conversion to UO 2 to make DUCRETE for a multi-purpose storage container, (3) conversion to depleted uranium metal for use as shielding, (4) conversion to uranium carbide for use as high-temperature gas-cooled reactor (HTGR) fuel. In addition, conversion to U 3 O 8 as an option for long-term storage is discussed

Evaluation of subcritical hybrid systems loaded with reprocessed fuel

International Nuclear Information System (INIS)

Velasquez, Carlos E.; Barros, Graiciany de P.; Pereira, Claubia; Veloso, Maria Auxiliadora F.; Costa, Antonella L.

2015-01-01

Highlights: • Accelerator driven systems (ADS) and fusion–fission systems are investigated for transmutation and fuel regeneration. • The calculations were performed using Monteburns code. • The results indicate the most suitable system for achieve transmutation. - Abstract: Two subcritical hybrid systems containing spent fuel reprocessed by Ganex technique and spiked with thorium were submitted to neutron irradiation of two different sources: ADS (Accelerator-driven subcritical) and Fusion. The aim is to investigate the nuclear fuel evolution using reprocessed fuel and the neutronic parameters under neutron irradiation. The source multiplication factor and fuel depletion for both systems were analysed during 10 years. The simulations were performed using MONTEBURNS code (MCNP/ORIGEN). The results indicate the main differences when irradiating the fuel with different neutron sources as well as the most suitable system for achieving transmutation

A Critical Review of Practice of Equating the Reactivity of Spent Fuel to Fresh Fuel in Burnup Credit Criticality Safety Analyses for PWR Spent Fuel Pool Storage

International Nuclear Information System (INIS)

Wagner, J.C.; Parks, C.V.

2000-01-01

This research examines the practice of equating the reactivity of spent fuel to that of fresh fuel for the purpose of performing burnup credit criticality safety analyses for PWR spent fuel pool (SFP) storage conditions. The investigation consists of comparing k inf estimates based on reactivity equivalent fresh fuel enrichment (REFFE) to k inf estimates using the actual spent fuel isotopics. Analyses of selected storage configurations common in PWR SFPs show that this practice yields nonconservative results (on the order of a few tenths of a percent) in configurations in which the spent fuel is adjacent to higher-reactivity assemblies (e.g., fresh or lower-burned assemblies) and yields conservative results in configurations in which spent fuel is adjacent to lower-reactivity assemblies (e.g., higher-burned fuel or empty cells). When the REFFE is determined based on unborated water moderation, analyses for storage conditions with soluble boron present reveal significant nonconservative results associated with the use of the REFFE. This observation is considered to be important, especially considering the recent allowance of credit for soluble boron up to 5% in reactivity. Finally, it is shown that the practice of equating the reactivity of spent fuel to fresh fuel is acceptable, provided the conditions for which the REFFE was determined remain unchanged. Determination of the REFFE for a reference configuration and subsequent use of the REFFE for different configurations violates the basis used for the determination of the REFFE and, thus, may lead to inaccurate, and possibly, nonconservative estimates of reactivity. A significant concentration (approx. 2000 ppm) of soluble boron is typically (but not necessarily required to be) present in PWR SFPs, of which only a portion (le 500 ppm) may be credited in safety analyses. Thus, a large subcritical margin currently exists that more than accounts for errors or uncertainties associated with the use of the REFFE

CRC DEPLETION CALCULATIONS FOR THE NON-RODDED ASSEMBLIES IN BATCHES 8 AND 9 CRYSTAL RIVER UNIT 3

International Nuclear Information System (INIS)

Wilson, Michael L.

2001-01-01

The purpose of this design analysis is to document the SAS2H depletion calculations of certain non-rodded fuel assemblies from batches 8 and 9 of the Crystal River Unit 3 pressurized water reactor (PWR) that are required for Commercial Reactor Critical (CRC) evaluations to support the development of the disposal criticality methodology. A non-rodded assembly is one which never contains a control rod assembly (CRA) or an axial power shaping rod assembly (APSRA) during its irradiation history. The objective of this analysis is to provide SAS2H generated isotopic compositions for each fuel assembly's depleted fuel and depleted burnable poison materials. These SAS2H generated isotopic compositions are acceptable for use in CRC benchmark reactivity calculations containing the various fuel assemblies

Sensitivity and parametric evaluations of significant aspects of burnup credit for PWR spent fuel packages

Energy Technology Data Exchange (ETDEWEB)

DeHart, M.D.

1996-05-01

Spent fuel transportation and storage cask designs based on a burnup credit approach must consider issues that are not relevant in casks designed under a fresh-fuel loading assumption. For example, the spent fuel composition must be adequately characterized and the criticality analysis model can be complicated by the need to consider axial burnup variations. Parametric analyses are needed to characterize the importance of fuel assembly and fuel cycle parameters on spent fuel composition and reactivity. Numerical models must be evaluated to determine the sensitivity of criticality safety calculations to modeling assumptions. The purpose of this report is to describe analyses and evaluations performed in order to demonstrate the effect physical parameters and modeling assumptions have on the criticality analysis of spent fuel. The analyses in this report include determination and ranking of the most important actinides and fission products; study of the effect of various depletion scenarios on subsequent criticality calculations; establishment of trends in neutron multiplication as a function of fuel enrichment, burnup, cooling time- and a parametric and modeling evaluation of three-dimensional effects (e.g., axially varying burnup and temperature/density effects) in a conceptual cask design. The sensitivity and parametric evaluations were performed with the consideration of two different burnup credit approaches: (1) only actinides in the fuel are considered in the criticality analysis, and (2) both actinides and fission products are considered. Calculations described in this report were performed using the criticality and depletion sequences available in the SCALE code system and the SCALE 27-group burnup library. Although the results described herein do not constitute a validation of SCALE for use in spent fuel analysis, independent validation efforts have been completed and are described in other reports.

Sensitivity and parametric evaluations of significant aspects of burnup credit for PWR spent fuel packages

International Nuclear Information System (INIS)

DeHart, M.D.

1996-05-01

Spent fuel transportation and storage cask designs based on a burnup credit approach must consider issues that are not relevant in casks designed under a fresh-fuel loading assumption. For example, the spent fuel composition must be adequately characterized and the criticality analysis model can be complicated by the need to consider axial burnup variations. Parametric analyses are needed to characterize the importance of fuel assembly and fuel cycle parameters on spent fuel composition and reactivity. Numerical models must be evaluated to determine the sensitivity of criticality safety calculations to modeling assumptions. The purpose of this report is to describe analyses and evaluations performed in order to demonstrate the effect physical parameters and modeling assumptions have on the criticality analysis of spent fuel. The analyses in this report include determination and ranking of the most important actinides and fission products; study of the effect of various depletion scenarios on subsequent criticality calculations; establishment of trends in neutron multiplication as a function of fuel enrichment, burnup, cooling time- and a parametric and modeling evaluation of three-dimensional effects (e.g., axially varying burnup and temperature/density effects) in a conceptual cask design. The sensitivity and parametric evaluations were performed with the consideration of two different burnup credit approaches: (1) only actinides in the fuel are considered in the criticality analysis, and (2) both actinides and fission products are considered. Calculations described in this report were performed using the criticality and depletion sequences available in the SCALE code system and the SCALE 27-group burnup library. Although the results described herein do not constitute a validation of SCALE for use in spent fuel analysis, independent validation efforts have been completed and are described in other reports

SCALE Continuous-Energy Monte Carlo Depletion with Parallel KENO in TRITON

International Nuclear Information System (INIS)

Goluoglu, Sedat; Bekar, Kursat B.; Wiarda, Dorothea

2012-01-01

The TRITON sequence of the SCALE code system is a powerful and robust tool for performing multigroup (MG) reactor physics analysis using either the 2-D deterministic solver NEWT or the 3-D Monte Carlo transport code KENO. However, as with all MG codes, the accuracy of the results depends on the accuracy of the MG cross sections that are generated and/or used. While SCALE resonance self-shielding modules provide rigorous resonance self-shielding, they are based on 1-D models and therefore 2-D or 3-D effects such as heterogeneity of the lattice structures may render final MG cross sections inaccurate. Another potential drawback to MG Monte Carlo depletion is the need to perform resonance self-shielding calculations at each depletion step for each fuel segment that is being depleted. The CPU time and memory required for self-shielding calculations can often eclipse the resources needed for the Monte Carlo transport. This summary presents the results of the new continuous-energy (CE) calculation mode in TRITON. With the new capability, accurate reactor physics analyses can be performed for all types of systems using the SCALE Monte Carlo code KENO as the CE transport solver. In addition, transport calculations can be performed in parallel mode on multiple processors.

'THERMOST' for analysing thermo-structural behaviour of LWR fuel rods under PCI conditions

International Nuclear Information System (INIS)

Nuno, H.; Ogawa, S.; Kobayashi, H.

1983-01-01

As a method for evaluating fuel rod performance under power ramping or load following operations, the combined FROST/ THERMOST system has been developed and brought into practical use. FROST was presented at the IAEA Blackpool Meeting in 1978, and THERMOST is the subject of this paper. The major purpose of THERMOST is to analyse very detailed thermal and structural fuel behaviour in a rather localised part of the fuel rod whereas FROST deals with whole rod general performance. The code handles two-dimensional thermal and structural analyses simultaneously by using a finite element method, in axial section or in lateral section. It consists of a fundamental FEM system of generalised constitution, and a surrounding subroutine system which characterises fuel behaviour, such as temperature distribution, thermal expansion, elastoplasticity, creep, cracking, swelling, growth, etc. Thermal analysis is handled by heat conduction and heat transfer element (six kinds), and structural analysis by axisymmetric ring and lateral plane element (six kinds). Boundary problems such as contact, friction and cracking are treated by gap and crack elements. A sample calculation of PCI performance on a PWR fuel rod under ramping conditions is presented with some in-pile test data. (author)

The role of ORIGEN-S in the design of burnup credit spent fuel casks

International Nuclear Information System (INIS)

Brady, M.C.

1991-01-01

Current licensing practices for spent fuel pools, storage facilities, and transportation casks require a conservative ''fresh fuel assumption'' be used in the criticality analysis. Burnup credit refers to a new approach in criticality analyses for spent fuel handling systems in which reactivity credit is allowed for the depleted state of the fuel. Studies have shown that the increased cask capacities that can be achieved with burnup credit offer both economic and risk incentives. The US Department of Energy is currently sponsoring a program to develop analysis methodologies and establish a new generation of spent fuel casks using the principle of burnup credit. The key difference in this new approach is the necessity to accurately predict the isotopic composition of the spent fuel. ORIGEN-S was selected to satisfy this requirement because of the flexibility and user-friendly input offered via its usage in the Standardized Computer Analyses for Licensing and Evaluation (SCALE) code system. Specifically, through the Shielding Analysis Sequence 2H (SAS2H), ORIGEN-S is linked with cross-section processing codes and one-dimensional transport analyses to produce problem-specific cross-section data for the point-depletion calculation. The utility code COUPLE facilitates updating basic cross-section and fission-yield data for the calculations. This paper describes the fundamental role fulfilled by ORIGEN-S in the development of the analysis methodology, validation of the methods, definition of criticality safety margins and other licensing considerations in the design of a new generation of spent fuel casks. Particular emphasis is given to the performance of ORIGEN-S in comparisons with measurements of irradiated fuel compositions and in predicting isotopics for use in the calculation of reactor restart critical configurations that are performed as a part of the validation process

High order depletion sensitivity analysis

International Nuclear Information System (INIS)

Naguib, K.; Adib, M.; Morcos, H.N.

2002-01-01

A high order depletion sensitivity method was applied to calculate the sensitivities of build-up of actinides in the irradiated fuel due to cross-section uncertainties. An iteration method based on Taylor series expansion was applied to construct stationary principle, from which all orders of perturbations were calculated. The irradiated EK-10 and MTR-20 fuels at their maximum burn-up of 25% and 65% respectively were considered for sensitivity analysis. The results of calculation show that, in case of EK-10 fuel (low burn-up), the first order sensitivity was found to be enough to perform an accuracy of 1%. While in case of MTR-20 (high burn-up) the fifth order was found to provide 3% accuracy. A computer code SENS was developed to provide the required calculations

Physical characterization of biomass-based pyrolysis liquids. Application of standard fuel oil analyses

Energy Technology Data Exchange (ETDEWEB)

Oasmaa, A; Leppaemaeki, E; Koponen, P; Levander, J; Tapola, E [VTT Energy, Espoo (Finland). Energy Production Technologies

1998-12-31

The main purpose of the study was to test the applicability of standard fuel oil methods developed for petroleum-based fuels to pyrolysis liquids. In addition, research on sampling, homogeneity, stability, miscibility and corrosivity was carried out. The standard methods have been tested for several different pyrolysis liquids. Recommendations on sampling, sample size and small modifications of standard methods are presented. In general, most of the methods can be used as such but the accuracy of the analysis can be improved by minor modifications. Fuel oil analyses not suitable for pyrolysis liquids have been identified. Homogeneity of the liquids is the most critical factor in accurate analysis. The presence of air bubbles may disturb in several analyses. Sample preheating and prefiltration should be avoided when possible. The former may cause changes in the composition and structure of the pyrolysis liquid. The latter may remove part of organic material with particles. The size of the sample should be determined on the basis of the homogeneity and the water content of the liquid. The basic analyses of the Technical Research Centre of Finland (VTT) include water, pH, solids, ash, Conradson carbon residue, heating value, CHN, density, viscosity, pourpoint, flash point, and stability. Additional analyses are carried out when needed. (orig.) 53 refs.

Depleted uranium hexafluoride: Waste or resource?

Energy Technology Data Exchange (ETDEWEB)

Schwertz, N.; Zoller, J.; Rosen, R.; Patton, S. [Lawrence Livermore National Lab., CA (United States); Bradley, C. [USDOE Office of Nuclear Energy, Science, Technology, Washington, DC (United States); Murray, A. [SAIC (United States)

1995-07-01

the US Department of Energy is evaluating technologies for the storage, disposal, or re-use of depleted uranium hexafluoride (UF{sub 6}). This paper discusses the following options, and provides a technology assessment for each one: (1) conversion to UO{sub 2} for use as mixed oxide duel, (2) conversion to UO{sub 2} to make DUCRETE for a multi-purpose storage container, (3) conversion to depleted uranium metal for use as shielding, (4) conversion to uranium carbide for use as high-temperature gas-cooled reactor (HTGR) fuel. In addition, conversion to U{sub 3}O{sub 8} as an option for long-term storage is discussed.

CRC DEPLETION CALCULATIONS FOR THE NON-RODDED ASSEMBLIES IN BATCHES 4 AND 5 OF CRYSTAL RIVER UNIT 3

International Nuclear Information System (INIS)

Wright, Kenneth D.

1997-01-01

The purpose of this design analysis is to document the SAS2H depletion calculations of certain non-rodded fuel assemblies from batches 4 and 5 of the Crystal River Unit 3 pressurized water reactor (PWR) that are required for commercial Reactor Critical (CRC) evaluations to support the development of the disposal criticality methodology. A non-rodded assembly is one which never contains a control rod assembly (CRA) or an axial power shaping rod assembly (APSRA) during its irradiation history. The objective of this analysis is to provide SAS2H generated isotopic compositions for each fuel assembly's depleted fuel and depleted burnable poison materials. These SAS2H generated isotopic compositions are acceptable for use in CRC benchmark reactivity calculations containing the various fuel assemblies

Revisit of analytical methods for the process and plant control analyses during reprocessing of fast reactor fuels

International Nuclear Information System (INIS)

Subba Rao, R.V.

2016-01-01

CORAL (COmpact facility for Reprocessing of Advanced fuels in Lead cell) is an experimental facility for demonstrating the reprocessing of irradiated fast reactor fuels discharged from the Fast Breeder Test Reactor (FBTR). The objective of the reprocessing plant is to achieve nuclear grade plutonium and uranium oxides with minimum process waste volumes. The process flow sheet for the reprocessing of spent Fast Reactor Fuel consists of Transport of spent fuel, Chopping, Dissolution, Feed conditioning, Solvent Extraction cycle, Partitioning Cycle and Re-conversion of Plutonium nitrate and uranium nitrate to respective oxides. The efficiency and performance of the plant to achieve desired objective depends on the analyses of various species in the different steps adopted during reprocessing of fuels. The analytical requirements in the plant can be broadly classified as 1. Process control Analyses (Analyses which effect the performance of the plant- PCA); 2. Plant control Analyses (Analyses which indicates efficiency of the plant-PLCA); 3. Nuclear Material Accounting samples (Analyses which has bearing on nuclear material accounting in the plant - NUMAC) and Quality control Analyses (Quality of the input bulk chemicals as well as products - QCA). The analytical methods selected are based on the duration of analyses, precision and accuracies required for each type analytical requirement classified earlier. The process and plant control analyses requires lower precision and accuracies as compared to NUMAC analyses, which requires very high precision accuracy. The time taken for analyses should be as lower as possible for process and plant control analyses as compared to NUMAC analyses. The analytical methods required for determining U and Pu in process and plant samples from FRFR will be different as compared to samples from TRFR (Thermal Reactor Fuel Reprocessing) due to higher Pu to U ratio in FRFR as compared TRFR and they should be such that they can be easily

Depleted uranium: A DOE management guide

International Nuclear Information System (INIS)

1995-10-01

The U.S. Department of Energy (DOE) has a management challenge and financial liability in the form of 50,000 cylinders containing 555,000 metric tons of depleted uranium hexafluoride (UF 6 ) that are stored at the gaseous diffusion plants. The annual storage and maintenance cost is approximately $10 million. This report summarizes several studies undertaken by the DOE Office of Technology Development (OTD) to evaluate options for long-term depleted uranium management. Based on studies conducted to date, the most likely use of the depleted uranium is for shielding of spent nuclear fuel (SNF) or vitrified high-level waste (HLW) containers. The alternative to finding a use for the depleted uranium is disposal as a radioactive waste. Estimated disposal costs, utilizing existing technologies, range between $3.8 and $11.3 billion, depending on factors such as applicability of the Resource Conservation and Recovery Act (RCRA) and the location of the disposal site. The cost of recycling the depleted uranium in a concrete based shielding in SNF/HLW containers, although substantial, is comparable to or less than the cost of disposal. Consequently, the case can be made that if DOE invests in developing depleted uranium shielded containers instead of disposal, a long-term solution to the UF 6 problem is attained at comparable or lower cost than disposal as a waste. Two concepts for depleted uranium storage casks were considered in these studies. The first is based on standard fabrication concepts previously developed for depleted uranium metal. The second converts the UF 6 to an oxide aggregate that is used in concrete to make dry storage casks

Technical Basis for Peak Reactivity Burnup Credit for BWR Spent Nuclear Fuel in Storage and Transportation Systems

Energy Technology Data Exchange (ETDEWEB)

Marshall, William BJ J [ORNL; Ade, Brian J [ORNL; Bowman, Stephen M [ORNL; Gauld, Ian C [ORNL; Ilas, Germina [ORNL; Mertyurek, Ugur [ORNL; Radulescu, Georgeta [ORNL

2015-01-01

Oak Ridge National Laboratory and the United States Nuclear Regulatory Commission have initiated a multiyear project to investigate application of burnup credit for boiling-water reactor (BWR) fuel in storage and transportation casks. This project includes two phases. The first phase (1) investigates applicability of peak reactivity methods currently used in spent fuel pools (SFPs) to storage and transportation systems and (2) evaluates validation of both reactivity (k_eff) calculations and burnup credit nuclide concentrations within these methods. The second phase will focus on extending burnup credit beyond peak reactivity. This paper documents the first phase, including an analysis of lattice design parameters and depletion effects, as well as both validation components. Initial efforts related to extended burnup credit are discussed in a companion paper. Peak reactivity analyses have been used in criticality analyses for licensing of BWR fuel in SFPs over the last 20 years. These analyses typically combine credit for the gadolinium burnable absorber present in the fuel with a modest amount of burnup credit. Gadolinium burnable absorbers are used in BWR assemblies to control core reactivity. The burnable absorber significantly reduces assembly reactivity at beginning of life, potentially leading to significant increases in assembly reactivity for burnups less than 15–20 GWd/MTU. The reactivity of each fuel lattice is dependent on gadolinium loading. The number of gadolinium-bearing fuel pins lowers initial lattice reactivity, but it has a small impact on the burnup and reactivity of the peak. The gadolinium concentration in each pin has a small impact on initial lattice reactivity but a significant effect on the reactivity of the peak and the burnup at which the peak occurs. The importance of the lattice parameters and depletion conditions are primarily determined by their impact on the gadolinium depletion. Criticality code validation for BWR burnup

Potential benefits and impacts on the CRWMS transportation system of filling spent fuel shipping casks with depleted uranium silicate glass

International Nuclear Information System (INIS)

Pope, R.B.; Forsberg, C.W.; DeHart, M.D.; Childs, K.W.; Tang, J.S.

1996-01-01

A new technology, the Depleted Uranium Silicate COntainer Fill System (DUSCOFS), is proposed to improve the performance and reduce the uncertainties of geological disposal of spent nuclear fuel (SNF), thus reducing both radionuclide release rates from the waste package and the potential for repository nuclear criticality events. DUSCOFS may also provide benefits for SNF storage and transport if it is loaded into the container early in the waste management cycle. Assessments have been made of the benefits to be derived by placing depleted uranium silicate (DUS) glass into SNF containers for enhancing repository performance assessment and controlling criticality over geologic times in the repository. Also, the performance, benefits, and impacts which can be derived if the SNF is loaded into a multi-purpose canister with DUS glass at a reactor site have been assessed. The DUSCOFS concept and the benefits to the waste management cycle of implementing DUSCOFS early in the cycle are discussed in this paper

Core Designs and Economic Analyses of Homogeneous Thoria-Urania Fuel in Light Water Reactors

International Nuclear Information System (INIS)

Saglam, Mehmet; Sapyta, Joe J.; Spetz, Stewart W.; Hassler, Lawrence A.

2004-01-01

The objective is to develop equilibrium fuel cycle designs for a typical pressurized water reactor (PWR) loaded with homogeneously mixed uranium-thorium dioxide (ThO 2 -UO 2 ) fuel and compare those designs with more conventional UO 2 designs.The fuel cycle analyses indicate that ThO 2 -UO 2 fuel cycles are technically feasible in modern PWRs. Both power peaking and soluble boron concentrations tend to be lower than in conventional UO 2 fuel cycles, and the burnable poison requirements are less.However, the additional costs associated with the use of homogeneous ThO 2 -UO 2 fuel in a PWR are significant, and extrapolation of the results gives no indication that further increases in burnup will make thoria-urania fuel economically competitive with the current UO 2 fuel used in light water reactors

CRC DEPLETION CALCULATIONS FOR THE NON-RODDED ASSEMBLIES IN BATCHES 4 AND 5 OF CRYSTAL RIVER UNIT 3

Energy Technology Data Exchange (ETDEWEB)

Kenneth D. Wright

1997-07-30

The purpose of this design analysis is to document the SAS2H depletion calculations of certain non-rodded fuel assemblies from batches 4 and 5 of the Crystal River Unit 3 pressurized water reactor (PWR) that are required for commercial Reactor Critical (CRC) evaluations to support the development of the disposal criticality methodology. A non-rodded assembly is one which never contains a control rod assembly (CRA) or an axial power shaping rod assembly (APSRA) during its irradiation history. The objective of this analysis is to provide SAS2H generated isotopic compositions for each fuel assembly's depleted fuel and depleted burnable poison materials. These SAS2H generated isotopic compositions are acceptable for use in CRC benchmark reactivity calculations containing the various fuel assemblies.

CRDIAC: Coupled Reactor Depletion Instrument with Automated Control

International Nuclear Information System (INIS)

Logan, Steven K.

2012-01-01

When modeling the behavior of a nuclear reactor over time, it is important to understand how the isotopes in the reactor will change, or transmute, over that time. This is especially important in the reactor fuel itself. Many nuclear physics modeling codes model how particles interact in the system, but do not model this over time. Thus, another code is used in conjunction with the nuclear physics code to accomplish this. In our code, Monte Carlo N-Particle (MCNP) codes and the Multi Reactor Transmutation Analysis Utility (MRTAU) were chosen as the codes to use. In this way, MCNP would produce the reaction rates in the different isotopes present and MRTAU would use cross sections generated from these reaction rates to determine how the mass of each isotope is lost or gained. Between these two codes, the information must be altered and edited for use. For this, a Python 2.7 script was developed to aid the user in getting the information in the correct forms. This newly developed methodology was called the Coupled Reactor Depletion Instrument with Automated Controls (CRDIAC). As is the case in any newly developed methodology for modeling of physical phenomena, CRDIAC needed to be verified against similar methodology and validated against data taken from an experiment, in our case AFIP-3. AFIP-3 was a reduced enrichment plate type fuel tested in the ATR. We verified our methodology against the MCNP Coupled with ORIGEN2 (MCWO) method and validated our work against the Post Irradiation Examination (PIE) data. When compared to MCWO, the difference in concentration of U-235 throughout Cycle 144A was about 1%. When compared to the PIE data, the average bias for end of life U-235 concentration was about 2%. These results from CRDIAC therefore agree with the MCWO and PIE data, validating and verifying CRDIAC. CRDIAC provides an alternative to using ORIGEN-based methodology, which is useful because CRDIAC's depletion code, MRTAU, uses every available isotope in its depletion

CRC DEPLETION CALCULATIONS FOR THE NON-RODDED ASSEMBLIES IN BATCHES 1, 2, AND 3 OF CRYSTAL RIVER UNIT 3

International Nuclear Information System (INIS)

Wright, Kenneth D.

1997-01-01

The purpose of this design analysis is to document the SAS2H depletion calculations of certain non-rodded fuel assemblies from batches 1, 2, and 3 of the Crystal River Unit 3 pressurized water reactor (PWR) that are required for Commercial Reactor Critical (CRC) evaluations to support development of the disposal criticality methodology. A non-rodded assembly is one which never contains a control rod assembly (CRA) or an axial power shaping rod assembly (APSRA) during its irradiation history. The objective of this analysis is to provide SAS2H generated isotopic compositions for each fuel assembly's depleted fuel and depleted burnable poison materials. These SAS2H generated isotopic compositions are acceptable for use in CRC benchmark reactivity calculations containing the various fuel assemblies

Life cycle analyses applied to first generation bio-fuels consumed in France

International Nuclear Information System (INIS)

2010-01-01

This rather voluminous publication reports detailed life cycle analyses for the different present bio-fuels channels also named first-generation bio-fuels: bio-ethanol, bio-diesel, pure vegetal oils, and oil. After a recall of the general principles adopted for this life-cycle analysis, it reports the modelling of the different channels (agricultural steps, bio-fuel production steps, Ethyl tert-butyl ether or ETBE steps, vehicles, animal fats and used vegetal oils, soil assignment change). It gives synthetic descriptions of the different production ways (methyl ester from different plants, ethanol from different plants). It reports and compares the results obtained in terms of performance

Optimization to reduce fuel consumption in charge depleting mode

Science.gov (United States)

Roos, Bryan Nathaniel; Martini, Ryan D.

2014-08-26

A powertrain includes an internal combustion engine, a motor utilizing electrical energy from an energy storage device, and a plug-in connection. A Method for controlling the powertrain includes monitoring a fuel cut mode, ceasing a fuel flow to the engine based upon the fuel cut mode, and through a period of operation including acceleration of the powertrain, providing an entirety of propelling torque to the powertrain with the electrical energy from the energy storage device based upon the fuel cut mode.

Riddle of depleted uranium

International Nuclear Information System (INIS)

Hussein, A.S.

2005-01-01

Depleted Uranium (DU) is the waste product of uranium enrichment from the manufacturing of fuel rods for nuclear reactors in nuclear power plants and nuclear power ships. DU may also results from the reprocessing of spent nuclear reactor fuel. Potentially DU has both chemical and radiological toxicity with two important targets organs being the kidney and the lungs. DU is made into a metal and, due to its availability, low price, high specific weight, density and melting point as well as its pyrophoricity; it has a wide range of civilian and military applications. Due to the use of DU over the recent years, there appeared in some press on health hazards that are alleged to be due to DU. In these paper properties, applications, potential environmental and health effects of DU are briefly reviewed

Comparison of DUPIC fuel composition heterogeneity control methods

International Nuclear Information System (INIS)

Choi, Hang Bok; Ko, Won Il

1999-08-01

A method to reduce the fuel composition heterogeneity effect on the core performance parameters has been studied for the DUPIC fuel which is made of spent pressurized water reactor (PWR) fuels by a dry refabrication process. This study focuses on the reactivity control method which uses either slightly enriched, depleted, or natural uranium to minimize the cost rise effect on the manufacturing of DUPIC fuel, when adjusting the excess reactivity control by slightly enriched and depleted uranium, reactivity control by natural uranium for high reactivity spent PWR fuels, and reactivity control by natural uranium for linear reactivity spent PWR fuels. The results of this study have shown that the reactivity control by slightly enriched and depleted uranium, all the spent PWR fuels can be utilized as the DUPIC fuel and the fraction of fresh uranium feed is 3.4% on an average. For the reactivity control by natural uranium, about 88% of spent PWR fuel can be utilized as the DUPIC fuel when the linear reactivity spent PWR fuels are used, and the amount of natural uranium feed needed to control the DUPIC fuel reactivity is negligible. (author). 13 refs., 16 tabs., 6 figs

Reverse depletion method for PWR core reload design

International Nuclear Information System (INIS)

Downar, T.J.; Kim, Y.J.

1985-01-01

Low-leakage fuel management is currently practiced in over half of all pressurized water reactor (PWR) cores. Prospects for even greater use of in-board fresh fuel loading are good as utilities seek to reduce core vessel fluence, mitigate pressurized thermal shock concerns, and extend vessel lifetime. Consequently, large numbers of burnable poison (BP) pins are being used to control the power peaking at the in-board fresh fuel positions. This has presented an additional complexity to the core reload design problem. In addition to determining the best location of each assembly in the core, the designer must concurrently determine the distribution of BP pins in the fresh fuel. A procedure was developed that utilizes the well-known Haling depletion to achieve an end-of-cycle (EOC) core state where the assembly pattern is configured in the absence of all control poison. This effectively separates the assembly assignment and BP distribution problems. Once an acceptable pattern at EOC is configured, the burnable and soluble poison required to control the power and core excess reactivity are solved for as unknown variables while depleting the cycle in reverse from the EOC exposure distribution to the beginning of cycle. The methods developed were implemented in an approved light water reactor licensing code to ensure the validity of the results obtained and provide for the maximum utility to PWR core reload design

Fuel tank integrity research : fuel tank analyses and test plans

Science.gov (United States)

2013-04-15

The Federal Railroad Administrations Office of Research : and Development is conducting research into fuel tank : crashworthiness. Fuel tank research is being performed to : determine strategies for increasing the fuel tank impact : resistance to ...

CRC DEPLETION CALCULATIONS FOR THE RODDED ASSEMBLIES IN BATCHES 1, 2, 3, AND 1X OF CRYSTAL RIVER UNIT 3

Energy Technology Data Exchange (ETDEWEB)

Kenneth D. Wright

1997-09-03

The purpose of this design analysis is to document the SAS2H depletion calculations of certain rodded fuel assemblies from batches 1, 2, 3, and 1X of the Crystal River Unit 3 pressurized water reactor (PWR) that are required for Commercial Reactor Critical (CRC) evaluations to support the development of the disposal criticality methodology. A rodded assembly is one that contains a control rod assembly (CRA) or an axial power shaping rod assembly (APSRA) for some period of time during its irradiation history. The objective of this analysis is to provide SAS2H calculated isotopic compositions of depleted fuel and depleted burnable poison for each fuel assembly to be used in subsequent CRC reactivity calculations containing the fuel assemblies.

CRC DEPLETION CALCULATIONS FOR THE RODDED ASSEMBLIES IN BATCHES 1, 2, 3, AND 1X OF CRYSTAL RIVER UNIT 3

International Nuclear Information System (INIS)

Wright, Kenneth D.

1997-01-01

The purpose of this design analysis is to document the SAS2H depletion calculations of certain rodded fuel assemblies from batches 1, 2, 3, and 1X of the Crystal River Unit 3 pressurized water reactor (PWR) that are required for Commercial Reactor Critical (CRC) evaluations to support the development of the disposal criticality methodology. A rodded assembly is one that contains a control rod assembly (CRA) or an axial power shaping rod assembly (APSRA) for some period of time during its irradiation history. The objective of this analysis is to provide SAS2H calculated isotopic compositions of depleted fuel and depleted burnable poison for each fuel assembly to be used in subsequent CRC reactivity calculations containing the fuel assemblies

Studies on the safety and transmutation behaviour of innovative fuels for light water reactors

International Nuclear Information System (INIS)

Schitthelm, Oliver

2012-01-01

Nuclear power plants contribute a substantial part to the energy demand in industry. Today the most common fuel cycle uses enriched uranium which produces plutonium due to its 238 U content. With respect to the long-term waste disposal Plutonium is an issue due to its heat production and radiotoxicity. This thesis consists of three main parts. In the first part the development and validation of a new code package MCBURN for spatial high resolution burnup simulations is presented. In the second part several innovative uranium-free and plutonium-burning fuels are evaluated on assembly level. Candidates for these fuels are a thorium/plutonium fuel and an inert matrix fuel consisting of plutonium dispersed in an enriched molybdenum matrix. The performance of these fuels is evaluated against existing MOX and enriched uranium fuels considering the safety and transmutation behaviour. The evaluation contains the boron efficiency, the void coefficient, the doppler coefficient and the net balances of every radionuclide. In the third part these innovative fuels are introduced into a German KONVOI reactor core. Considering todays approved usage of MOX fuels a partial loading of one third of innovative fuels and two third of classical uranium fuels was analysed. The efficiency of the plutonium depletion is determined by the ratio of the production of higher isotopes compared to the plutonium depletion. Todays MOX-fuels transmutate about 25% to 30% into higher actinides as Americium or Curium. In uranium-free fuels this ratio is about 10% due to the lack of additional plutonium production. The analyses of the reactor core have shown that one third of MOX fuel is not capable of a net reduction of plutonium. On the other hand a partial loading with thorium/plutonium fuel incinerates about half the amount of plutonium produced by an uranium only core. If IMF is used the ratio increases to about 75%. Considering the safety behavior all fuels have shown comparable results.

ISODEP, A Fuel Depletion Analysis Code for Predicting Isotopic ...

African Journals Online (AJOL)

The trend of results was found to be consistent with those obtained by analytical and other numerical methods. Discovery and Innovation Vol. 13 no. 3/4 December (2001) pp. 184-195. KEY WORDS: depletion analysis, code, research reactor, simultaneous equations, decay of nuclides, radionuclitides, isotope. Résumé

Bullet scintigraphy: can gamma camera be used for depleted uranium accident measurements?

International Nuclear Information System (INIS)

Spaic, R.; Markovic, S.; Pavlovic, S.; Radic, Z.; Pavlovic, R.; Ajdinovic, B.; Baskot, B.; Djurovic, B.

2002-01-01

The aim of this study was to see could gamma cameras be used for measurement of internal contamination with depleted uranium. Radioactive waste depleted uranium, which is by-product from the production of enriched fuel for nuclear rectors and weapons now, is used for manufacture bullets, which are used in Iraq, Republic of Srpska and Yugoslavia. In this paper is measured minimum detectable activity (MDA) of gamma cameras for depleted uranium, iodine and technetium. For detection of the depleted uranium are used low energy X-rays, energy of 100 keV with 20% windows width. About 40% of gamma emissions of the depleted uranium are in these limits. Measured MDA activities 50-100 Bq for depleted uranium, iodine and technetium are about then times more then same for WBC (5 Bq). Gamma cameras can be used for relatively measurement of depleted uranium activity, what can be used for absorbed dose estimation. Detection of low level internal contamination with depleted uranium can be done with gamma cameras. (authors)

SPOTS, Library Generator for Program LEOPARD from Cross-Sections Data. LEOPARD, Fast and Thermal Neutron Spectra from Temperature and Geometry with Depletion Calculation

International Nuclear Information System (INIS)

Barry, R.F.; Krug, H.E P. Jr.

1981-01-01

1 - Description of problem or function: LEOPARD is a unit cell homogenization and spectrum generation (MUFT-SOFOCATE type) program with a fuel depletion option. 2 - Method of solution: The MUFT-SOFOCATE homogeneous medium spectrum analyses with heterogeneous corrections are used. The monoenergetic Amouyal-Benoist thermal disadvantage factor is applied at each of 172 SOFOCATE energy levels up to 0.625 eV. The U-238 resonance integral is forced to agree with a generalized Hellstrand correlation. 3 - Restrictions on the complexity of the problem: LEOPARD works with nuclides commonly used in water reactors. Thorium and U-238 fuel chains are allowed

TURTLE 24.0 diffusion depletion code

International Nuclear Information System (INIS)

Altomare, S.; Barry, R.F.

1971-09-01

TURTLE is a two-group, two-dimensional (x-y, x-z, r-z) neutron diffusion code featuring a direct treatment of the nonlinear effects of xenon, enthalpy, and Doppler. Fuel depletion is allowed. TURTLE was written for the study of azimuthal xenon oscillations, but the code is useful for general analysis. The input is simple, fuel management is handled directly, and a boron criticality search is allowed. Ten thousand space points are allowed (over 20,000 with diagonal symmetry). TURTLE is written in FORTRAN IV and is tailored for the present CDC-6600. The program is core-contained. Provision is made to save data on tape for future reference. (auth)

Economic analyses of LWR fuel cycles

International Nuclear Information System (INIS)

Field, F.R.

1977-05-01

An economic comparison was made of three options for handling irradiated light-water reactor (LWR) fuel. These options are reprocessing of spent reactor fuel and subsequent recycle of both uranium and plutonium, reprocessing and recycle of uranium only, and direct terminal storage of spent fuel not reprocessed. The comparison was based on a peak-installed nuclear capacity of 507 GWe by CY 2000 and retirement of reactors after 30 years of service. Results of the study indicate that: Through the year 2000, recycle of uranium and plutonium in LWRs saves about $12 billion (FY 1977 dollars) compared with the throwaway cycle, but this amounts to only about 1.3% of the total cost of generating electricity by nuclear power. If deferred costs are included for fuel that has been discharged from reactors but not reprocessed, the economic advantage increases to $17.7 billion. Recycle of uranium only (storage of plutonium) is approximately $7 billion more expensive than the throwaway fuel cycle and is, therefore, not considered an economically viable option. The throwaway fuel cycle ultimately requires >40% more uranium resources (U 3 O 8 ) than does reprocessing spent fuel where both uranium and plutonium are recycled

A radiochemical analyses of metastudtite and leachates from spent fuel

International Nuclear Information System (INIS)

McNamara, Bruce K.; Hanson, Brady D.; Buck, Edgar C.; Soderquist, Chuck Z.

2004-01-01

Immersion of commercial spent nuclear fuel (CSNF) in deionized water produced two novel corrosion products after a two-year contact period. Another unexpected result was that suspensions of aggregates were observed to form at the air-water interface for each of five samples. These solids were characterized, by SEM and XRD to be nearly pure metastudtite (UO4-2H2O); while the corrosion present on the surface of the fuel itself was determined to be studtite (UO4-2H2O). The occurrence of the floating phase prompted a radiochemical analysis of these solids. This chemical analysis was a unique opportunity to study the relatively pure corrosion phase for incorporation of radionuclides. The analysis indicated that high concentration of 90Sr, 137Cs, 99Tc, and that lower concentrations 237Np, 238, 239Pu and 243, 244Cm had partitioned with the air-water interface aggregates. The concentrations of 241Am were two orders of magnitude lower than the expected inventory in the suspended solids. The radiochemical analyses of the several leachate samples provide preliminary solubility data for the hydrogen peroxide leaching of CSNF and these data are compared to leaching of the same fuel in J-13 and deionized waters. The extent of fuel dissolution in these media are discussed

Analyses of the double-layed repository concepts for spent nuclear fuels

Energy Technology Data Exchange (ETDEWEB)

Lee, Jong Youl; Kim, Hyeona; Lee, Min Soo; Choi, Heui Joo; Kim, Kyung Su [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2017-06-15

A deep geological disposal at a depth of 500 m in stable host rock is considered to be the safest method with current technologies for disposal of spent fuels classified as high-level radioactive waste. The most important requirement is that the temperature of the bentonite buffer, which is a component of the engineered barrier, should not exceed 100℃. In Korea, the amount of spent fuel generated by nuclear power generation, which accounts for about 30% of the total electricity, is continuously increasing and accumulating. Accordingly, the area required to dispose of it is also increasing. In this study, various duplex disposal concepts were derived for the purpose of improving the disposal efficiency by reducing the disposal area. Based on these concepts, thermal analyses were carried out to confirm whether the critical disposal system requirements were met, and the thermal stability of the disposal system was evaluated by analyzing the results. The results showed that upward 75 m or downward 75 m apart from the reference disposal system location of 500 m depth would qualify for the double layered disposal concept. The results of this study can be applied to the establishment of spent fuel management policy and the design of practical commercial disposal system. Detailed analyses with data of a real disposal site are necessary.

The ''THERMOST'' for analysing thermo-structural behaviour of LWR fuel rod under PCI conditions

International Nuclear Information System (INIS)

Nuno, H.; Ogawa, S.; Kobayashi, H.

1983-01-01

As one of the methods for evaluating the fuel rod performances under power ramping or load following operations, the combined ''FROST'' and ''THERMOST'' system has been developed and being brought into practical use. The former had already been presented at Blackpool Meeting in 1978, and the latter is going to be presented in this paper. The major purpose of the THERMOST is to analyse very detailed thermal and structural fuel behaviours in a rather localized part of fuel rod whereas the FROST deals with whole-rod-wide general performances. The code handles 2-dimensional thermal and structural analyses simultaneously by using finite element method, in axial section wide or in lateral section wide. It consists of a fundamental FEM system of generalized constitution and its surrounding subroutine system which characterizes fuel behaviours such as temperature distribution, thermal expansion, elastoplasticity, creep, cracking, swelling, growth, etc. Thermal analysis is handled by heat conduction and heat transfer elements (6 kinds) and structural analysis by axisymmetric ring and lateral plane elements (6 kinds). Boundary problems such as contact, friction and cracking are treated by gap and crack elements. A sample calculation of PCI performance on a PWR fuel rod under ramping condition is presented with some inpile test data. (author)

Impact of modeling Choices on Inventory and In-Cask Criticality Calculations for Forsmark 3 BWR Spent Fuel

International Nuclear Information System (INIS)

Martinez-Gonzalez, Jesus S.; Ade, Brian J.; Bowman, Stephen M.; Gauld, Ian C.; Ilas, Germina; Marshall, William BJ J.

2015-01-01

Simulation of boiling water reactor (BWR) fuel depletion poses a challenge for nuclide inventory validation and nuclear criticality safety analyses. This challenge is due to the complex operating conditions and assembly design heterogeneities that characterize these nuclear systems. Fuel depletion simulations and in-cask criticality calculations are affected by (1) completeness of design information, (2) variability of operating conditions needed for modeling purposes, and (3) possible modeling choices. These effects must be identified, quantified, and ranked according to their significance. This paper presents an investigation of BWR fuel depletion using a complete set of actual design specifications and detailed operational data available for five operating cycles of the Swedish BWR Forsmark 3 reactor. The data includes detailed axial profiles of power, burnup, and void fraction in a very fine temporal mesh for a GE14 (10x10) fuel assembly. The specifications of this case can be used to assess the impacts of different modeling choices on inventory prediction and in-cask criticality, specifically regarding the key parameters that drive inventory and reactivity throughout fuel burnup. This study focused on the effects of the fidelity with which power history and void fraction distributions are modeled. The corresponding sensitivity of the reactivity in storage configurations is assessed, and the impacts of modeling choices on decay heat and inventory are addressed.

An Overview of Stationary Fuel Cell Technology

Energy Technology Data Exchange (ETDEWEB)

DR Brown; R Jones

1999-03-23

Technology developments occurring in the past few years have resulted in the initial commercialization of phosphoric acid (PA) fuel cells. Ongoing research and development (R and D) promises further improvement in PA fuel cell technology, as well as the development of proton exchange membrane (PEM), molten carbonate (MC), and solid oxide (SO) fuel cell technologies. In the long run, this collection of fuel cell options will be able to serve a wide range of electric power and cogeneration applications. A fuel cell converts the chemical energy of a fuel into electrical energy without the use of a thermal cycle or rotating equipment. In contrast, most electrical generating devices (e.g., steam and gas turbine cycles, reciprocating engines) first convert chemical energy into thermal energy and then mechanical energy before finally generating electricity. Like a battery, a fuel cell is an electrochemical device, but there are important differences. Batteries store chemical energy and convert it into electrical energy on demand, until the chemical energy has been depleted. Depleted secondary batteries may be recharged by applying an external power source, while depleted primary batteries must be replaced. Fuel cells, on the other hand, will operate continuously, as long as they are externally supplied with a fuel and an oxidant.

Thermoelastic analyses of spent fuel repositories in bedded and dome salt. Technical memorandum report RSI-0054

International Nuclear Information System (INIS)

Callahan, G.D.; Ratigan, J.L.

1978-01-01

Global thermoelastic analyses of bedded and dome salt models showed a slight preference for the bedded salt model through the range of thermal loading conditions. Spent fuel thermal loadings should be less than 75 kW/acre of the repository pending more accurate material modeling. One should first limit the study to one or two spent fuel thermal loading (i.e. 75 kW/acre and/or 50 kW/acre) analyses up to a maximum time of approximately 2000 years. Parametric thermoelastic type analyses could then be readily obtained to determine the influence of the thermomechanical properties. Recommendations for further study include parametric analyses, plasticity analyses, consideration of the material interfaces as joints, and possibly consideration of a global joint pattern (i.e. jointed at the same orientation everywhere) for the non-salt materials. Subsequently, the viscoelastic analyses could be performed

Cooking fuel use patterns in India: 1983-2000

International Nuclear Information System (INIS)

Viswanathan, Brinda; Kavi Kumar, K.S.

2005-01-01

This study analyses the expenditure share of 'clean' and 'dirty' fuels in total cooking fuel consumption for the rural and urban households across 16 major states in India, using household level data from national sample surveys conducted during the period 1983-2000. The results show wide disparity between rural and urban households and also across states. Analysis to identify the determinants of fuel choice reveals that affordability plays a major role, while the pro-rich and pro-urban bias of kerosene supply through public distribution system also has influenced the observed variation in consumption patterns across states and over rural and urban areas. The study discusses the policies that could facilitate switch towards 'clean' fuels and argues that enabling policies should pay attention among other things to the gender issues and trade-offs that exist between say, local and global pollution, deforestation and resource depletion, and disease and subsidy burden

Enhancement of MARS with an Advanced Fuel Model by Coupling FRAPTRAN

Energy Technology Data Exchange (ETDEWEB)

Kim, Hyong Chol; Lee, Young Jin; Han, Sam Hee [KAERI, Daejeon (Korea, Republic of)

2016-05-15

FRAPTRAN calculates heat conduction, heat transfer from cladding to coolant, elastic-plastic fuel and cladding deformation, cladding oxidation, fission gas release, and fuel rod gas pressure. FRAPTRAN is used for analyzing the fuel response under postulated accidents such as reactivity-initiated accidents (RIAs) and loss-of-coolant accidents (LOCAs), and also for analyzing and interpreting experimental results. Burnup dependent variables such as fuel densification and swelling, and cladding creep and irradiation growth may be considered by incorporating FRAPCON steady state depletion calculation results as the initial conditions. FRAPTRAN-DLL has been successfully verified and the coupled calculations have shown to provide reasonable results. An EOC core loaded with irradiated fuels was analyzed with the integrated code system. The coupled code system has demonstrated its applicability to variety of applications such as assessing the effects of fuel thermal conductivity degradation with burnup. MARS has been enhanced with the advanced fuel model of FRAPTRAN so that users can use the fuel rod performance evaluation capability in the transient analyses.

Plutonium in depleted uranium penetrators

International Nuclear Information System (INIS)

McLaughlin, J.P.; Leon-Vintro, L.; Smith, K.; Mitchell, P.I.; Zunic, Z.S.

2002-01-01

Depleted Uranium (DU) penetrators used in the recent Balkan conflicts have been found to be contaminated with trace amounts of transuranic materials such as plutonium. This contamination is usually a consequence of DU fabrication being carried out in facilities also using uranium recycled from spent military and civilian nuclear reactor fuel. Specific activities of 239+240 Plutonium generally in the range 1 to 12 Bq/kg have been found to be present in DU penetrators recovered from the attack sites of the 1999 NATO bombardment of Kosovo. A DU penetrator recovered from a May 1999 attack site at Bratoselce in southern Serbia and analysed by University College Dublin was found to contain 43.7 +/- 1.9 Bq/kg of 239+240 Plutonium. This analysis is described. An account is also given of the general population radiation dose implications arising from both the DU itself and from the presence of plutonium in the penetrators. According to current dosimetric models, in all scenarios considered likely ,the dose from the plutonium is estimated to be much smaller than that due to the uranium isotopes present in the penetrators. (author)

Sensitivity of nuclear fuel cycle cost to uncertainties in nuclear data

International Nuclear Information System (INIS)

Harris, D.R.; Becker, M.; Parvez, A.; Ryskamp, J.M.

1979-01-01

A sensitivity analysis system is developed for assessing the economic implications of uncertainties in nuclear data and related computational methods for light water power reactors. Results of the sensitivity analysis indicate directions for worthwhile improvements in data and methods. Benefits from improvements in data and methods are related to reduction of margins provided by designers to ensure meeting reactor and fuel objectives. Sensitivity analyses are carried out using the batch depletion code FASTCELL, the core analysis code FASTCORE, and the reactor cost code COSTR. FASTCELL depletes a cell using methods comparable to industry cell codes except for a few-group treatment of cell flux distribution. FASTCORE is used with the Haling strategy of fixed power sharing among batches in the core. COSTR computes costs using components and techniques as in industry costing codes, except that COSTR uses fixed payment schedules. Sensitivity analyses are carried out for large commercial boiling and pressurized water reactors. Each few-group nuclear parameter is changed, and initial enrichment is also changed so as to keep the end-of-cycle core multiplication factor unchanged, i.e., to preserve cycle time at the demand power. Sensitivities of equilibrium fuel cycle cost are determined with respect to approx. 300 few-group nuclear parameters, both for a normal fuel cycle and for a throwaway fuel cycle. Particularly large dollar implications are found for thermal and resonance range cross sections in fissile and fertile materials. Sensitivities constrained by adjustment of fission neutron yield so as to preserve agreement with zero exposure integral data also are computed

Fuel cycle in Japanese Fugen - HWR

International Nuclear Information System (INIS)

1979-04-01

This paper describes the use of plutonium-bearing fuel in the Japanese Fugen-HWR. The Fugen-HWR is a pressure tube type, boiling light water cooled, and heavy water moderated reactor, which by using plutonium fuel (MOX) achieves the advantage of high neutron economy. The characteristics of the reactor are discussed, particularly its ability to operate with several different types of fuel - Pu-natural U MOX, Pu-Depleted U (from spent LWR fuel) MOX, Pu-Depleted U (from enrichment tails) MOX, and enriched UO 2 . The natural U and separative work units saved are given and the fuel management and control of the reactor discussed. Non-proliferation and safety considerations are given. The Fugen-HWR achieved 100% power rating in the autumn of 1979

Development of an Application Programming Interface for Depletion Analysis (APIDA)

International Nuclear Information System (INIS)

Lago, Daniel; Rahnema, Farzad

2017-01-01

Highlights: • APIDA an Application Programming Interface tool for Depletion Analysis. • APIDA employs a matrix exponential method and a linear chain method. • A burnup solver to couple to neutron transport solvers in lattice depletion or reactor core analysis codes. - Abstract: A new utility has been developed with extensive capabilities in identifying nuclide decay and transmutation characteristics, allowing for accurate and efficient tracking of the change in isotopic concentrations in nuclear reactor fuel over time when coupled with a transport solution method. This tool, named the Application Programming Interface for Depletion Analysis (APIDA), employs both a matrix exponential method and a linear chain method to solve for the end-of-time-step nuclide concentrations for all isotopes relevant to nuclear reactors. The Chebyshev Rational Approximation Method (CRAM) was utilized to deal with the ill-conditioned matrices generated during lattice depletion calculations, and a complex linear chain solver was developed to handle isotopes reduced from the burnup matrix due to either radioactive stability or a sufficiently low neutron-induced reaction cross section. The entire tool is housed in a robust but simple application programming interface (API). The development of this API allows other codes, particularly numerical neutron transport solvers, to incorporate APIDA as the burnup solver in a lattice depletion code or reactor core analysis code in memory, without the need to write or read from the hard disk. The APIDA code was benchmarked using several decay and transmutation chains. Burnup solutions produced by APIDA were shown to provide material concentrations comparable to the analytically solved Bateman equations – well below 0.01% relative error for even the most extreme cases using isotopes with vastly different decay constants. As a first order demonstration of the API, APIDA was coupled with the transport solver in the SERPENT code for a fuel pin

Development of a micro-depletion model to us WIMS properties in history-based local-parameter calculations in RFSP

International Nuclear Information System (INIS)

Shen, W.

2004-01-01

A micro-depletion model has been developed and implemented in the *SIMULATE module of RFSP to use WIMS-calculated lattice properties in history-based local-parameter calculations. A comparison between the micro-depletion and WIMS results for each type of lattice cross section and for the infinite-lattice multiplication factor was also performed for a fuel similar to that which may be used in the ACR fuel. The comparison shows that the micro-depletion calculation agrees well with the WIMS-IST calculation. The relative differences in k-infinity are within ±0.5 mk and ±0.9 mk for perturbation and depletion calculations, respectively. The micro-depletion model gives the *SIMULATE module of RFSP the capability to use WIMS-calculated lattice properties in history-based local-parameter calculations without resorting to the Simple-Cell-Methodology (SCM) surrogate for CANDU core-tracking simulations. (author)

Comparison of DUPIC fuel composition heterogeneity control methods

Energy Technology Data Exchange (ETDEWEB)

Choi, Hang Bok; Ko, Won Il [Korea Atomic Energy Research Institute, Taejon (Korea)

1999-08-01

A method to reduce the fuel composition heterogeneity effect on the core performance parameters has been studied for the DUPIC fuel which is made of spent pressurized water reactor (PWR) fuels by a dry refabrication process. This study focuses on the reactivity control method which uses either slightly enriched, depleted, or natural uranium to minimize the cost rise effect on the manufacturing of DUPIC fuel, when adjusting the excess reactivity of the spent PWR fuel. In order to reduce the variation of isotopic composition of the DUPIC fuel, the inter-assembly mixing operation was taken three times. Then, three options have been considered: reactivity control by slightly enriched and depleted uranium, reactivity control by natural uranium for high reactivity spent PWR fuels, and reactivity control by natural uranium for linear reactivity spent PWR fuels. The results of this study have shown that the reactivity of DUPIC fuel can be tightly controlled with the minimum amount of fresh uranium feed. For the reactivity control by slightly enriched and depleted uranium, all the spent PWR fuels can be utilized as the DUPIC fuel and the fraction of fresh uranium feed is 3.4% on an average. For the reactivity control by natural uranium, about 88% of spent PWR fuel can be utilized as the DUPIC fuel when the linear reactivity spent PWR fuels are used, and the amount of natural uranium feed needed to control the DUPIC fuel reactivity is negligible. 13 refs., 6 figs., 16 tabs. (Author)

Assessing fuel spill risks in polar waters: Temporal dynamics and behaviour of hydrocarbons from Antarctic diesel, marine gas oil and residual fuel oil.

Science.gov (United States)

Brown, Kathryn E; King, Catherine K; Kotzakoulakis, Konstantinos; George, Simon C; Harrison, Peter L

2016-09-15

As part of risk assessment of fuel oil spills in Antarctic and subantarctic waters, this study describes partitioning of hydrocarbons from three fuels (Special Antarctic Blend diesel, SAB; marine gas oil, MGO; and intermediate grade fuel oil, IFO 180) into seawater at 0 and 5°C and subsequent depletion over 7days. Initial total hydrocarbon content (THC) of water accommodated fraction (WAF) in seawater was highest for SAB. Rates of THC loss and proportions in equivalent carbon number fractions differed between fuels and over time. THC was most persistent in IFO 180 WAFs and most rapidly depleted in MGO WAF, with depletion for SAB WAF strongly affected by temperature. Concentration and composition remained proportionate in dilution series over time. This study significantly enhances our understanding of fuel behaviour in Antarctic and subantarctic waters, enabling improved predictions for estimates of sensitivities of marine organisms to toxic contaminants from fuels in the region. Copyright © 2016 Elsevier Ltd. All rights reserved.

Electron probe microanalysis of a METAPHIX UPuZr metallic alloy fuel irradiated to 7.0 at.% burn-up

Energy Technology Data Exchange (ETDEWEB)

Brémier, S., E-mail: stephan.bremier@ec.europa.eu [European Commission, Joint Research Centre, Institute for Transuranium Elements, P.O. Box 2340, D-76125 Karlsruhe (Germany); Inagaki, K. [Central Research Institute of Electric Power Industry, Nuclear Technology Research Laboratory, 2-11-1 Iwado-kita, Komae-shi, Tokyo 201-8511 (Japan); Capriotti, L.; Poeml, P. [European Commission, Joint Research Centre, Institute for Transuranium Elements, P.O. Box 2340, D-76125 Karlsruhe (Germany); Ogata, T.; Ohta, H. [Central Research Institute of Electric Power Industry, Nuclear Technology Research Laboratory, 2-11-1 Iwado-kita, Komae-shi, Tokyo 201-8511 (Japan); Rondinella, V.V. [European Commission, Joint Research Centre, Institute for Transuranium Elements, P.O. Box 2340, D-76125 Karlsruhe (Germany)

2016-11-15

The METAPHIX project is a collaboration between CRIEPI and JRC-ITU investigating safety and performance of a closed fuel cycle option based on fast reactor metal alloy fuels containing Minor Actinides (MA). The aim of the project is to investigate the behaviour of this type of fuel and demonstrate the transmutation of MA under irradiation. A UPuZr metallic fuel sample irradiated to a burn-up of 7 at.% was examined by electron probe microanalysis. The fuel sample was extensively characterised qualitatively and quantitatively using elemental X-ray imaging and point analysis techniques. The analyses reveal a significant redistribution of the fuel components along the fuel radius highlighting a nearly complete depletion of Zr in the central part of the fuel. Numerous rare earth and fission products secondary phases are present in various compositions. Fuel cladding chemical interaction was observed with creation of a number of intermediary layers affecting a cladding depth of 15–20 μm and migration of cladding elements to the fuel. - Highlights: • Electron Probe MicroAnalysis of a UPuZr metallic fuel alloy irradiated to 7.0 at.% burn-up. • Significant redistribution of the fuel components along the fuel radius, nearly complete depletion of Zr in the central part of the fuel. • Interactions between the fuel and the cladding with occurrence of a number of intermediary layers and migration of cladding elements to the fuel. • Safe irradiation behaviour of the base alloy fuel.

Department of Energy depleted uranium recycle

International Nuclear Information System (INIS)

Kosinski, F.E.; Butturini, W.G.; Kurtz, J.J.

1994-01-01

With its strategic supply of depleted uranium, the Department of Energy is studying reuse of the material in nuclear radiation shields, military hardware, and commercial applications. the study is expected to warrant a more detailed uranium recycle plan which would include consideration of a demonstration program and a program implementation decision. Such a program, if implemented, would become the largest nuclear material recycle program in the history of the Department of Energy. The bulk of the current inventory of depleted uranium is stored in 14-ton cylinders in the form of solid uranium hexafluoride (UF 6 ). The radioactive 235 U content has been reduced to a concentration of 0.2% to 0.4%. Present estimates indicate there are about 55,000 UF 6 -filled cylinders in inventory and planned operations will provide another 2,500 cylinders of depleted uranium each year. The United States government, under the auspices of the Department of Energy, considers the depleted uranium a highly-refined strategic resource of significant value. A possible utilization of a large portion of the depleted uranium inventory is as radiation shielding for spent reactor fuels and high-level radioactive waste. To this end, the Department of Energy study to-date has included a preliminary technical review to ascertain DOE chemical forms useful for commercial products. The presentation summarized the information including preliminary cost estimates. The status of commercial uranium processing is discussed. With a shrinking market, the number of chemical conversion and fabrication plants is reduced; however, the commercial capability does exist for chemical conversion of the UF 6 to the metal form and for the fabrication of uranium radiation shields and other uranium products. Department of Energy facilities no longer possess a capability for depleted uranium chemical conversion

Analytical out-of-pile and in-pile experiments on gadolinia bearing fuels

International Nuclear Information System (INIS)

Bruet, M.; Francois, B.; Do, Q.; Bergeron, J.; Trotabas, M.

1986-06-01

New fuel management schemes in PWRs can be achieved through the use of burnable poisons like gadolinia bearing fuel rods. However, the introduction of such a design has required a qualification program, which has been performed in collaboration between CEA, FRAGEMA and/or FRAMATOME by specialized teams in CEA facilities. The main scoops of this program concern: the fabrication process; the out of pile physical properties determination: the in pile thermomechanical behaviour and fission product release; the neutronic studies in view to validate the Computed Gd efficiency and the LBP depletion calculation schemes and to analyse and assess various schemes of core calculations

Bioethanol: fuel or feedstock?

DEFF Research Database (Denmark)

Rass-Hansen, Jeppe; Falsig, Hanne; Jørgensen, Betina

2007-01-01

Increasing amounts of bioethanol are being produced from fermentation of biomass, mainly to counteract the continuing depletion of fossil resources and the consequential escalation of oil prices. Today, bioethanol is mainly utilized as a fuel or fuel additive in motor vehicles, but it could also...

Uncertainty Propagation in Monte Carlo Depletion Analysis

International Nuclear Information System (INIS)

Shim, Hyung Jin; Kim, Yeong-il; Park, Ho Jin; Joo, Han Gyu; Kim, Chang Hyo

2008-01-01

A new formulation aimed at quantifying uncertainties of Monte Carlo (MC) tallies such as k eff and the microscopic reaction rates of nuclides and nuclide number densities in MC depletion analysis and examining their propagation behaviour as a function of depletion time step (DTS) is presented. It is shown that the variance of a given MC tally used as a measure of its uncertainty in this formulation arises from four sources; the statistical uncertainty of the MC tally, uncertainties of microscopic cross sections and nuclide number densities, and the cross correlations between them and the contribution of the latter three sources can be determined by computing the correlation coefficients between the uncertain variables. It is also shown that the variance of any given nuclide number density at the end of each DTS stems from uncertainties of the nuclide number densities (NND) and microscopic reaction rates (MRR) of nuclides at the beginning of each DTS and they are determined by computing correlation coefficients between these two uncertain variables. To test the viability of the formulation, we conducted MC depletion analysis for two sample depletion problems involving a simplified 7x7 fuel assembly (FA) and a 17x17 PWR FA, determined number densities of uranium and plutonium isotopes and their variances as well as k ∞ and its variance as a function of DTS, and demonstrated the applicability of the new formulation for uncertainty propagation analysis that need be followed in MC depletion computations. (authors)

Contribution to fuel depletion study in PWR type reactors, reactor core with three and four regions of enrichment; Contribuicao ao estudo da evolucao da composicao do combustivel em reatores tipo PWR nucleos a tres e a quatro regioes de enriquecimento

Energy Technology Data Exchange (ETDEWEB)

Teixeira, M C.C.

1977-03-01

The main methods for calculation of fuel depletion are studied and some approaches to do it are mentioned; the LEOPARD Code is described and full details are given for each subroutine, flow charts are included; the method given by the code for calculation of fuel depletion is described; some imperfections from the IPR`s version are listed, and corrected, for instance: the method for burn-up calculation of heavy isotopes; the results of calculations for a reference reactor based on data of the Preliminary Safety Analysis Report (PSAR) for Angra I Nuclear Power Plant are presented and discussed. (author).

The manufacturing of depleted uranium biological shield components

International Nuclear Information System (INIS)

Metelkin, J.A.

1998-01-01

The unique combination of the physical and mechanical properties of uranium made it possible to manufacture biological shield components of transport package container (TPC) for transportation nuclear power plant irradiated fuel and radionuclides of radiation diagnostic instruments. Protective properties are substantially dependent on the nature radionuclide composition of uranium, that why I recommended depleted uranium after radiation chemical processing. Depleted uranium biological shield (DUBS) has improved specific mass-size characteristics compared to a shield made of lead, steel or tungsten. Technological achievements in uranium casting and machining made it possible to manufacture DUBS components of TPC up to 3 tons of mass and up to 2 metres of the maximum size. (authors)

Safety demonstration analyses for severe accident of fresh nuclear fuel transport packages at JAERI

International Nuclear Information System (INIS)

Yamada, K.; Watanabe, K.; Nomura, Y.; Okuno, H.; Miyoshi, Y.

2004-01-01

It is expected in the near future that more and more fresh nuclear fuel will be transported in a variety of transport packages to cope with increasing demand from nuclear fuel cycle facilities. Accordingly, safety demonstration analyses of these methods are planned and conducted at JAERI under contract with the Ministry of Economy, Trade and Industry of Japan. These analyses are conducted part of a four year plan from 2001 to 2004 to verify integrity of packaging against leakage of radioactive material in the case of a severe accident envisioned to occur during transportation, for the purpose of gaining public acceptance of such nuclear fuel activities. In order to create the accident scenarios, actual transportation routes were surveyed, accident or incident records were tracked, international radioactive material transport regulations such as IAEA rules were investigated and, thus, accident conditions leading to mechanical damage and thermal failure were selected for inclusion in the scenario. As a result, the worst-case conditions of run-off-the-road accidents were incorporated, where there is impact against a concrete or asphalt surface. Fire accidents were assumed to occur after collision with a tank truck carrying lots of inflammable material or destruction by fire after collision inside a tunnel. The impact analyses were performed by using three-dimensional elements according to the general purpose impact analysis code LS-DYNA. Leak-tightness of the package was maintained even in the severe impact accident scenario. In addition, the thermal analyses were performed by using two-dimensional elements according to the general purpose finite element method computer code ABAQUS. As a result of these analyses, the integrity of the inside packaging component was found to be sufficient to maintain a leak-tight state, confirming its safety

A reverse depletion method for pressurized water reactor core reload design

International Nuclear Information System (INIS)

Downar, T.J.; Kin, Y.J.

1986-01-01

Low-leakage fuel management is currently practiced in over half of all pressurized water reactor (PWR) cores. The large numbers of burnable poison pins used to control the power peaking at the in-board fresh fuel positions have introduced an additional complexity to the core reload design problem. In addition to determining the best location of each assembly in the core, the designer must concurrently determine the distribution of burnable poison pins in the fresh fuel. A new method for performing core design more suitable for low-leakage fuel management is reported. A procedure was developed that uses the wellknown ''Haling depletion'' to achieve an end-of-cycle (EOC) core state where the assembly pattern is configured in the absence of all control poison. This effectively separates the assembly assignment and burnable poison distribution problems. Once an acceptable pattern at EOC is configured, the burnable and soluble poison required to control the power and core excess reactivity are solved for as unknown variables while depleting the cycle in reverse from the EOC exposure distribution to the beginning of cycle. The methods developed were implemented in an approved light water reactor licensing code to ensure the validity of the results obtained and provided for the maximum utility to PWR core reload design

Stochastic methods for the quantification of sensitivities and uncertainties in criticality analyses; Stochastische Methoden zur Quantifizierung von Sensitivitaeten und Unsicherheiten in Kritikalitaetsanalysen

Energy Technology Data Exchange (ETDEWEB)

Behler, Matthias; Bock, Matthias; Stuke, Maik; Wagner, Markus

2014-06-15

This work describes statistical analyses based on Monte Carlo sampling methods for criticality safety analyses. The methods analyse a large number of calculations of a given problem with statistically varied model parameters to determine uncertainties and sensitivities of the computed results. The GRS development SUnCISTT (Sensitivities and Uncertainties in Criticality Inventory and Source Term Tool) is a modular, easily extensible abstract interface program, designed to perform such Monte Carlo sampling based uncertainty and sensitivity analyses in the field of criticality safety. It couples different criticality and depletion codes commonly used in nuclear criticality safety assessments to the well-established GRS tool SUSA for sensitivity and uncertainty analyses. For uncertainty analyses of criticality calculations, SunCISTT couples various SCALE sequences developed at Oak Ridge National Laboratory and the general Monte Carlo N-particle transport code MCNP from Los Alamos National Laboratory to SUSA. The impact of manufacturing tolerances of a fuel assembly configuration on the neutron multiplication factor for the various sequences is shown. Uncertainties in nuclear inventories, dose rates, or decay heat can be investigated via the coupling of the GRS depletion system OREST to SUSA. Some results for a simplified irradiated Pressurized Water Reactor (PWR) UO{sub 2} fuel assembly are shown. SUnCISTT also combines the two aforementioned modules for burnup credit criticality analysis of spent nuclear fuel to ensures an uncertainty and sensitivity analysis using the variations of manufacturing tolerances in the burn-up code and criticality code simultaneously. Calculations and results for a storage cask loaded with typical irradiated PWR UO{sub 2} fuel are shown, including Monte Carlo sampled axial burn-up profiles. The application of SUnCISTT in the field of code validation, specifically, how it is applied to compare a simulation model to available benchmark

Castor-1C spent fuel storage cask decay heat, heat transfer, and shielding analyses

International Nuclear Information System (INIS)

Rector, D.R.; McCann, R.A.; Jenquin, U.P.; Heeb, C.M.; Creer, J.M.; Wheeler, C.L.

1986-12-01

This report documents the decay heat, heat transfer, and shielding analyses of the Gesellschaft fuer Nuklear Services (GNS) CASTOR-1C cask used in a spent fuel storage demonstration performed at Preussen Elektra's Wurgassen nuclear power plant. The demonstration was performed between March 1982 and January 1984, and resulted in cask and fuel temperature data and cask exterior surface gamma-ray and neutron radiation dose rate measurements. The purpose of the analyses reported here was to evaluate decay heat, heat transfer, and shielding computer codes. The analyses consisted of (1) performing pre-look predictions (predictions performed before the analysts were provided the test data), (2) comparing ORIGEN2 (decay heat), COBRA-SFS and HYDRA (heat transfer), and QAD and DOT (shielding) results to data, and (3) performing post-test analyses if appropriate. Even though two heat transfer codes were used to predict CASTOR-1C cask test data, no attempt was made to compare the two codes. The codes are being evaluated with other test data (single-assembly data and other cask data), and to compare the codes based on one set of data may be premature and lead to erroneous conclusions

The impact of spent fuel reprocessing facilities deployment rate on transuranics inventory in alternative fuel cycle strategies

International Nuclear Information System (INIS)

Aquien, A.; Kazimi, M.; Hejzlar, P.

2007-01-01

The depletion rate of transuranic inventories from spent fuel depends on both the deployment of advanced reactors that can be loaded with recycled transuranics, and on the deployment of the facilities that separate and reprocess spent fuel. In addition to tracking the mass allocation of TRU in the system and calculating a system cost, the fuel cycle simulation tool CAFCA includes a flexible recycling plant deployment model. This study analyses the impact of different recycling deployment schemes for various fuel cycle strategies in the US over the next hundred years under the assumption of a demand for nuclear energy growing at a rate of 2,4%. Recycling strategies explored in this study fall under two categories: recycling in thermal light water reactors using combined non-fertile and UO 2 fuel (CONFU) and recycling in fast reactors (either fertile-free actinide burner reactors, or self-sustaining gas-cooled fast reactors). Preliminary results show that the earlier deployment of recycling in the thermal reactors will limit the stored levels of TRU below those of fast reactors. However, the avoided accumulation of spent fuel interim storage depends on the deployment rate of the recycling facilities. In addition, by the end of the mid century, the TRU in cooling storage will exceed that in interim storage. (authors)

Hybrid microscopic depletion model in nodal code DYN3D

International Nuclear Information System (INIS)

Bilodid, Y.; Kotlyar, D.; Shwageraus, E.; Fridman, E.; Kliem, S.

2016-01-01

Highlights: • A new hybrid method of accounting for spectral history effects is proposed. • Local concentrations of over 1000 nuclides are calculated using micro depletion. • The new method is implemented in nodal code DYN3D and verified. - Abstract: The paper presents a general hybrid method that combines the micro-depletion technique with correction of micro- and macro-diffusion parameters to account for the spectral history effects. The fuel in a core is subjected to time- and space-dependent operational conditions (e.g. coolant density), which cannot be predicted in advance. However, lattice codes assume some average conditions to generate cross sections (XS) for nodal diffusion codes such as DYN3D. Deviation of local operational history from average conditions leads to accumulation of errors in XS, which is referred as spectral history effects. Various methods to account for the spectral history effects, such as spectral index, burnup-averaged operational parameters and micro-depletion, were implemented in some nodal codes. Recently, an alternative method, which characterizes fuel depletion state by burnup and 239 Pu concentration (denoted as Pu-correction) was proposed, implemented in nodal code DYN3D and verified for a wide range of history effects. The method is computationally efficient, however, it has applicability limitations. The current study seeks to improve the accuracy and applicability range of Pu-correction method. The proposed hybrid method combines the micro-depletion method with a XS characterization technique similar to the Pu-correction method. The method was implemented in DYN3D and verified on multiple test cases. The results obtained with DYN3D were compared to those obtained with Monte Carlo code Serpent, which was also used to generate the XS. The observed differences are within the statistical uncertainties.

Depleted uranium hexafluoride: The source material for advanced shielding systems

Energy Technology Data Exchange (ETDEWEB)

Quapp, W.J.; Lessing, P.A. [Idaho National Engineering Lab., Idaho Falls, ID (United States); Cooley, C.R. [Department of Technology, Germantown, MD (United States)

1997-02-01

The U.S. Department of Energy (DOE) has a management challenge and financial liability problem in the form of 50,000 cylinders containing 555,000 metric tons of depleted uranium hexafluoride (UF{sub 6}) that are stored at the gaseous diffusion plants. DOE is evaluating several options for the disposition of this UF{sub 6}, including continued storage, disposal, and recycle into a product. Based on studies conducted to date, the most feasible recycle option for the depleted uranium is shielding in low-level waste, spent nuclear fuel, or vitrified high-level waste containers. Estimates for the cost of disposal, using existing technologies, range between $3.8 and $11.3 billion depending on factors such as the disposal site and the applicability of the Resource Conservation and Recovery Act (RCRA). Advanced technologies can reduce these costs, but UF{sub 6} disposal still represents large future costs. This paper describes an application for depleted uranium in which depleted uranium hexafluoride is converted into an oxide and then into a heavy aggregate. The heavy uranium aggregate is combined with conventional concrete materials to form an ultra high density concrete, DUCRETE, weighing more than 400 lb/ft{sup 3}. DUCRETE can be used as shielding in spent nuclear fuel/high-level waste casks at a cost comparable to the lower of the disposal cost estimates. Consequently, the case can be made that DUCRETE shielded casks are an alternative to disposal. In this case, a beneficial long term solution is attained for much less than the combined cost of independently providing shielded casks and disposing of the depleted uranium. Furthermore, if disposal is avoided, the political problems associated with selection of a disposal location are also avoided. Other studies have also shown cost benefits for low level waste shielded disposal containers.

Safety analyses for an in-pile SCWR fuel qualification test loop

Energy Technology Data Exchange (ETDEWEB)

Schulenberg, T.; Raque, M. [Karlsruhe Inst. of Tech., Karlsruhe (Germany)

2014-07-01

As a nuclear facility cooled with supercritical water has never been built nor operated in the past, the planned SCWR fuel qualification test will give the first experience with supercritical water-cooled nuclear systems in general. With a fuel inventory of almost 1 kg of UO{sub 2} with almost 20% enrichment, the supercritical pressure test section inside a low pressure, pool type research reactor needs to be cooled properly even in case of a number of postulated design basis accidents. Depressurization systems and emergency cooling systems will need to be designed with similar reliability as for a prototype reactor to ensure the integrity of barriers retaining the radioactive material. The paper reports about the safety concept and summarizes the safety analyses which have been performed in this context. (author)

Energetic and exergetic analyses of T56 turboprop engine

International Nuclear Information System (INIS)

Balli, Ozgur; Hepbasli, Arif

2013-01-01

Highlights: • Performing comprehensive energy and exergy analyses of T56 turboprop engine at various operation modes. • Proposing two new parameters, energetic and exergetic fuel-production ratios. • Calculating maximum energy efficiency values of 25.4% for Case A and 28.1% for Case B at Takeoff mode. • Accounting maximum exergy efficiency values of 23.8% for Case A and 26.3% for Case B at Takeoff mode. - Abstract: This study presents the results of energetic and exergetic analyses of T56 turboprop engine at various power loading operation modes (75%, 100%, Military and Takeoff). The energetic and exergetic performance evaluations were made for both the shaft power (Case A) and the shaft power plus the kinetic energy of exhaust gaseous (Case B). The energetic efficiency was calculated to be maximum at 25.4% for Case A and 28.1% for Case B while the exergy efficiency was obtained to be maximum at 23.8% for Case A and 26.3% for Case B at Takeoff mode, respectively. The maximum exergy destruction rate occurred within the combustion chamber. It increased from 4846.3 kW to 6234.1 kW depending on operation modes. The exergetic performance parameters, such as the relative exergy consumption, the fuel depletion ratio, the productivity lack ratio, the improvement potential and the fuel-production ratio, were also investigated. The fuel energy-production ratio decreased from 4.6 to 3.9 while the fuel exergy-production ratio decreased from 4.9 to 4.2 by increasing the produced shaft power and residual thrust. The results provided here can be helpful to regulate and select operation modes for these engine users

Development, implementation, and verification of multicycle depletion perturbation theory for reactor burnup analysis

Energy Technology Data Exchange (ETDEWEB)

White, J.R.

1980-08-01

A generalized depletion perturbation formulation based on the quasi-static method for solving realistic multicycle reactor depletion problems is developed and implemented within the VENTURE/BURNER modular code system. The present development extends the original formulation derived by M.L. Williams to include nuclide discontinuities such as fuel shuffling and discharge. This theory is first described in detail with particular emphasis given to the similarity of the forward and adjoint quasi-static burnup equations. The specific algorithm and computational methods utilized to solve the adjoint problem within the newly developed DEPTH (Depletion Perturbation Theory) module are then briefly discussed. Finally, the main features and computational accuracy of this new method are illustrated through its application to several representative reactor depletion problems.

Depleted uranium management alternatives

Energy Technology Data Exchange (ETDEWEB)

Hertzler, T.J.; Nishimoto, D.D.

1994-08-01

This report evaluates two management alternatives for Department of Energy depleted uranium: continued storage as uranium hexafluoride, and conversion to uranium metal and fabrication to shielding for spent nuclear fuel containers. The results will be used to compare the costs with other alternatives, such as disposal. Cost estimates for the continued storage alternative are based on a life-cycle of 27 years through the year 2020. Cost estimates for the recycle alternative are based on existing conversion process costs and Capital costs for fabricating the containers. Additionally, the recycle alternative accounts for costs associated with intermediate product resale and secondary waste disposal for materials generated during the conversion process.

Depleted uranium management alternatives

International Nuclear Information System (INIS)

Hertzler, T.J.; Nishimoto, D.D.

1994-08-01

This report evaluates two management alternatives for Department of Energy depleted uranium: continued storage as uranium hexafluoride, and conversion to uranium metal and fabrication to shielding for spent nuclear fuel containers. The results will be used to compare the costs with other alternatives, such as disposal. Cost estimates for the continued storage alternative are based on a life-cycle of 27 years through the year 2020. Cost estimates for the recycle alternative are based on existing conversion process costs and Capital costs for fabricating the containers. Additionally, the recycle alternative accounts for costs associated with intermediate product resale and secondary waste disposal for materials generated during the conversion process

Analyses for inserting fresh LEU fuel assemblies instead of fresh HEU fuel assemblies in the Dalat Nuclear Research Reactor in Vietnam

International Nuclear Information System (INIS)

Hanan, N. A.; Deen, J.R.; Matos, J.E.

2005-01-01

Analyses were performed by the RERTR Program to replace 36 burned HEU (36%) fuel assemblies in the Dalat Nuclear Research Reactor in Vietnam with either 36 fresh fuel assemblies currently on-hand at the reactor or with LEU fuel assemblies to be procured. The study concludes that the current HEU (36%) WWR-M2 fuel assemblies can be replaced with LEU WWR-M2 fuel assemblies that are fully-qualified and have been commercially available since 2001 from the Novosibirsk Chemical Concentrates Plant in Russia. The current reactor configuration using re-shuffled HEU fuel began in June 2004 and is expected to allow normal operation until around August 2006. If 36 HEU assemblies each with 40.2 g 235 U are inserted without fuel shuffling over the next five operating cycles, the core could operate for an additional 10 years until June 2016. Alternatively, inserting 36 LEU fuel assemblies each containing 49.7 g 235 U without fuel shuffling over five operating cycles would allow normal operation for about 14 years from August 2006 until October 2020. The main reason for the longer service life of the LEU fuel is that its 235 U content is higher than the 235 U content needed simply to match the service life of the HEU fuel. Fast neutron fluxes in the experiment regions would be very nearly the same in both the HEU and LEU cores. Thermal neutron fluxes in the experiment regions would be lower by 1-5%, depending on the experiment type and location. (author)

Safety demonstration analyses at JAERI for severe accident during overland transport of fresh nuclear fuel

International Nuclear Information System (INIS)

Nomura, Yasushi; Kitao, Kohichi; Karasawa, Kiyonori; Yamada, Kenji; Takahashi, Satoshi; Watanabe, Kohji; Okuno, Hiroshi; Miyoshi, Yoshinori

2005-01-01

It is expected in the near future that more and more fresh nuclear fuel will be transported in a variety of transport packages to cope with increasing demand from nuclear fuel cycle facilities. Accordingly, safety demonstration analyses are planned and conducted at JAERI under contract with the Ministry of Economy, Trade and Industry of Japan. These analyses are conducted in a four year plan from 2001 to 2004 to verify integrity of packaging against leakage of radioactive material in the case of a severe accident postulated to occur during transportation, for the purpose of gaining acceptance of such nuclear fuel activities. In order to create the accident scenarios, actual transportation routes were surveyed, accident or incident records were tracked, international radioactive material transport regulations such as IAEA rules were investigated and thus, accident conditions leading to mechanical damages and thermal failure were determined to characterize the scenarios. As a result, the worst-case conditions of run-off-the-road accidents were set up to define the impact against a concrete or asphalt surface. For fire accident scenarios to be set up, collisions were assumed to occur with an oil tanker carrying lots of inflammable material in open air, or with a commonly used two-ton-truck inside a tunnel without ventilation. Then the cask models were determined for these safety demonstration analyses to represent those commonly used for fresh nuclear fuel transported throughout Japan. Following the postulated accident scenarios, the mechanical damages were analyzed by using the general-purpose finite element code LS-DYNA with three-dimensional elements. It was found that leak tightness of the package be maintained even in the severe impact scenario. Then the thermal safety was analyzed by using the general-purpose finite element code ABAOUS with three-dimensional elements to describe cask geometry. As a result of the thermal analyses, the integrity of the containment

Monte carlo depletion analysis of SMART core by MCNAP code

International Nuclear Information System (INIS)

Jung, Jong Sung; Sim, Hyung Jin; Kim, Chang Hyo; Lee, Jung Chan; Ji, Sung Kyun

2001-01-01

Depletion an analysis of SMART, a small-sized advanced integral PWR under development by KAERI, is conducted using the Monte Carlo (MC) depletion analysis program, MCNAP. The results are compared with those of the CASMO-3/ MASTER nuclear analysis. The difference between MASTER and MCNAP on k eff prediction is observed about 600pcm at BOC, and becomes smaller as the core burnup increases. The maximum difference bet ween two predict ions on fuel assembly (FA) normalized power distribution is about 6.6% radially , and 14.5% axially but the differences are observed to lie within standard deviation of MC estimations

Uncertainty Analyses of Advanced Fuel Cycles

International Nuclear Information System (INIS)

Miller, Laurence F.; Preston, J.; Sweder, G.; Anderson, T.; Janson, S.; Humberstone, M.; MConn, J.; Clark, J.

2008-01-01

The Department of Energy is developing technology, experimental protocols, computational methods, systems analysis software, and many other capabilities in order to advance the nuclear power infrastructure through the Advanced Fuel Cycle Initiative (AFDI). Our project, is intended to facilitate will-informed decision making for the selection of fuel cycle options and facilities for development

Uncertainty Analyses of Advanced Fuel Cycles

Energy Technology Data Exchange (ETDEWEB)

Laurence F. Miller; J. Preston; G. Sweder; T. Anderson; S. Janson; M. Humberstone; J. MConn; J. Clark

2008-12-12

The Department of Energy is developing technology, experimental protocols, computational methods, systems analysis software, and many other capabilities in order to advance the nuclear power infrastructure through the Advanced Fuel Cycle Initiative (AFDI). Our project, is intended to facilitate will-informed decision making for the selection of fuel cycle options and facilities for development.

Synthetic liquid fuels development: assessment of critical factors. Volume III. Coal resource depletion

Energy Technology Data Exchange (ETDEWEB)

Dickson, E.M.; Yabroff, I.W.; Kroll, C.A.; White, R.K.; Walton, B.L.; Ivory, M.E.; Fullen, R.E.; Weisbecker, L.W.; Hays, R.L.

1977-01-01

While US coal resources are known to be vast, their rate of depletion in a future based predominantly on coal has not been examined analytically heretofore. The Coal Depletion Model inventories the coal resource on a regional basis and calculates the cost of coal extraction by three technologies - strip and underground mining and in-situ combustion. A plausible coal demand scenario extending from 1975 to the year 2050 is used as a basis in applying the model. In the year 2050, plants in operation include 285 syncrude plants, each producing 100,000 B/D; 312 SNG plants, each producing 250 million SCF/D and 722 coal-fired electric power plants, each of 1000 MW capacity. In addition, there is 890 million tons per year of industrial coal consumption. Such a high level of coal use would deplete US coal resources much more rapidly than most people appreciate. Of course, the actual amount of US coal is unknown, and if the coal in the hypothetical reliability category is included, depletion is delayed. Coal in this category, however, has not been mapped; it is only presumed to exist on the basis of geological theory. The coal resource depletion model shows that unilateral imposition of a severance tax by a state tends to shift production to other coal producing regions. Boom and bust cycles are both delayed and reduced in their magnitude. When several states simultaneously impose severance taxes, the effect of each is weakened.Key policy issues that emerge from this analysis concern the need to reduce the uncertainty of the magnitude and geographic distribution of the US coal resource and the need to stimulate interaction among the parties at interest to work out equitable and acceptable coal conversion plant location strategies capable of coping with the challenges of a high-coal future.

Recent activity on the post-irradiation analyses of nuclear fuels and actinide samples at JAERI

International Nuclear Information System (INIS)

Shinohara, Nobuo; Nakahara, Yoshinori; Kohno, Nobuaki; Tsujimoto, Kazufumi

2003-01-01

Radiochemical analyses of spent fuels have been carried out at JAERI for contributing to the development of nuclear technologies, where several samples from research reactors and nuclear power plants were analyzed to obtain isotopic compositions and burnups. The history and procedures of the radiochemical analyses are depicted and some recent results are given in this paper. (author)

Metallographic analysis of irradiated RERTR-3 fuel test specimens

International Nuclear Information System (INIS)

Meyer, M. K.; Hofman, G. L.; Strain, R. V.; Clark, C. R.; Stuart, J. R.

2000-01-01

The RERTR-3 irradiation test was designed to investigate the irradiation behavior of aluminum matrix U-MO alloy dispersion fuels under high-temperature, high-fission-rate conditions. Initial postirradiation examination of RERTR-3 fuel specimens has concentrated on binary U-MO atomized fuels. The rate of matrix aluminum depletion was found to be higher than predictions based on low temperature irradiation data. Wavelength Dispersive X-ray Spectroscopy (WDS) indicates that aluminum is present in the interior of the fuel particles. WDS data is supported by a mass and volume balance calculation performed on the basis of image analysis results. The depletion of matrix aluminum seems to have no detrimental effects on fuel performance under the conditions tested to date

Depletion methodology in the 3-D whole core transport code DeCART

Energy Technology Data Exchange (ETDEWEB)

Kim, Kang Seog; Cho, Jin Young; Zee, Sung Quun

2005-02-01

Three dimensional whole-core transport code DeCART has been developed to include a characteristics of the numerical reactor to replace partly the experiment. This code adopts the deterministic method in simulating the neutron behavior with the least assumption and approximation. This neutronic code is also coupled with the thermal hydraulic code CFD and the thermo mechanical code to simulate the combined effects. Depletion module has been implemented in DeCART code to predict the depleted composition in the fuel. The exponential matrix method of ORIGEN-2 has been used for the depletion calculation. The library of including decay constants, yield matrix and others has been used and greatly simplified for the calculation efficiency. This report summarizes the theoretical backgrounds and includes the verification of the depletion module in DeCART by performing the benchmark calculations.

Spray and atomization of diesel fuel and its alternatives from a single-hole injector using a common rail fuel injection system

KAUST Repository

Chen, PinChia; Wang, Weicheng; Roberts, William L.; Fang, Tiegang

2013-01-01

Fuel spray and atomization characteristics play an important role in the performance of internal combustion engines. As the reserves of petroleum fuel are expected to be depleted within a few decades, finding alternative fuels that are economically

Greenhouse gas impacts of declining hydrocarbon resource quality: Depletion, dynamics, and process emissions

Science.gov (United States)

Brandt, Adam Robert

This dissertation explores the environmental and economic impacts of the transition to hydrocarbon substitutes for conventional petroleum (SCPs). First, mathematical models of oil depletion are reviewed, including the Hubbert model, curve-fitting methods, simulation models, and economic models. The benefits and drawbacks of each method are outlined. I discuss the predictive value of the models and our ability to determine if one model type works best. I argue that forecasting oil depletion without also including substitution with SCPs results in unrealistic projections of future energy supply. I next use information theoretic techniques to test the Hubbert model of oil depletion against five other asymmetric and symmetric curve-fitting models using data from 139 oil producing regions. I also test the assumptions that production curves are symmetric and that production is more bell-shaped in larger regions. Results show that if symmetry is enforced, Gaussian production curves perform best, while if asymmetry is allowed, asymmetric exponential models prove most useful. I also find strong evidence for asymmetry: production declines are consistently less steep than inclines. In order to understand the impacts of oil depletion on GHG emissions, I developed the Regional Optimization Model for Emissions from Oil Substitutes (ROMEO). ROMEO is an economic optimization model of investment and production of fuels. Results indicate that incremental emissions (with demand held constant) from SCPs could be 5-20 GtC over the next 50 years. These results are sensitive to the endowment of conventional oil and not sensitive to a carbon tax. If demand can vary, total emissions could decline under a transition because the higher cost of SCPs lessens overall fuel consumption. Lastly, I study the energetic and environmental characteristics of the in situ conversion process, which utilizes electricity to generate liquid hydrocarbons from oil shale. I model the energy inputs and outputs

Kr ion irradiation study of the depleted-uranium alloys

Science.gov (United States)

Gan, J.; Keiser, D. D.; Miller, B. D.; Kirk, M. A.; Rest, J.; Allen, T. R.; Wachs, D. M.

2010-12-01

Fuel development for the reduced enrichment research and test reactor (RERTR) program is tasked with the development of new low enrichment uranium nuclear fuels that can be employed to replace existing high enrichment uranium fuels currently used in some research reactors throughout the world. For dispersion type fuels, radiation stability of the fuel-cladding interaction product has a strong impact on fuel performance. Three depleted-uranium alloys are cast for the radiation stability studies of the fuel-cladding interaction product using Kr ion irradiation to investigate radiation damage from fission products. SEM analysis indicates the presence of the phases of interest: U(Al, Si) 3, (U, Mo)(Al, Si) 3, UMo 2Al 20, U 6Mo 4Al 43 and UAl 4. Irradiations of TEM disc samples were conducted with 500 keV Kr ions at 200 °C to ion doses up to 2.5 × 10 19 ions/m 2 (˜10 dpa) with an Kr ion flux of 10 16 ions/m 2/s (˜4.0 × 10 -3 dpa/s). Microstructural evolution of the phases relevant to fuel-cladding interaction products was investigated using transmission electron microscopy.

Time-on-task effects in children with and without ADHD: depletion of executive resources or depletion of motivation?

Science.gov (United States)

Dekkers, Tycho J; Agelink van Rentergem, Joost A; Koole, Alette; van den Wildenberg, Wery P M; Popma, Arne; Bexkens, Anika; Stoffelsen, Reino; Diekmann, Anouk; Huizenga, Hilde M

2017-12-01

Children with attention-deficit/hyperactivity disorder (ADHD) are characterized by deficits in their executive functioning and motivation. In addition, these children are characterized by a decline in performance as time-on-task increases (i.e., time-on-task effects). However, it is unknown whether these time-on-task effects should be attributed to deficits in executive functioning or to deficits in motivation. Some studies in typically developing (TD) adults indicated that time-on-task effects should be interpreted as depletion of executive resources, but other studies suggested that they represent depletion of motivation. We, therefore, investigated, in children with and without ADHD, whether there were time-on-task effects on executive functions, such as inhibition and (in)attention, and whether these were best explained by depletion of executive resources or depletion of motivation. The stop-signal task (SST), which generates both indices of inhibition (stop-signal reaction time) and attention (reaction time variability and errors), was administered in 96 children (42 ADHD, 54 TD controls; aged 9-13). To differentiate between depletion of resources and depletion of motivation, the SST was administered twice. Half of the participants was reinforced during second task performance, potentially counteracting depletion of motivation. Multilevel analyses indicated that children with ADHD were more affected by time-on-task than controls on two measures of inattention, but not on inhibition. In the ADHD group, reinforcement only improved performance on one index of attention (i.e., reaction time variability). The current findings suggest that time-on-task effects in children with ADHD occur specifically in the attentional domain, and seem to originate in both depletion of executive resources and depletion of motivation. Clinical implications for diagnostics, psycho-education, and intervention are discussed.

A semi-empirical model for the formation and depletion of the high burnup structure in UO{sub 2}

Energy Technology Data Exchange (ETDEWEB)

Pizzocri, D. [European Commission, Joint Research Centre, Directorate for Nuclear Safety and Security, PO Box 2340, 76125, Karlsruhe (Germany); Politecnico di Milano, Department of Energy, Nuclear Engineering Division, Via La Masa 34, 20156, Milan (Italy); Cappia, F. [European Commission, Joint Research Centre, Directorate for Nuclear Safety and Security, PO Box 2340, 76125, Karlsruhe (Germany); Technische Universität München, Boltzmannstraße 15, 85747, Garching bei München (Germany); Luzzi, L., E-mail: lelio.luzzi@polimi.it [Politecnico di Milano, Department of Energy, Nuclear Engineering Division, Via La Masa 34, 20156, Milan (Italy); Pastore, G. [Idaho National Laboratory, Fuel Modeling and Simulation Department, 2525 Fremont Avenue, 83415, Idaho Falls (United States); Rondinella, V.V.; Van Uffelen, P. [European Commission, Joint Research Centre, Directorate for Nuclear Safety and Security, PO Box 2340, 76125, Karlsruhe (Germany)

2017-04-15

In the rim zone of UO{sub 2} nuclear fuel pellets, the combination of high burnup and low temperature drives a microstructural change, leading to the formation of the high burnup structure (HBS). In this work, we propose a semi-empirical model to describe the formation of the HBS, which embraces the polygonisation/recrystallization process and the depletion of intra-granular fission gas, describing them as inherently related. For this purpose, we performed grain-size measurements on samples at radial positions in which the restructuring was incomplete. Based on these new experimental data, we infer an exponential reduction of the average grain size with local effective burnup, paired with a simultaneous depletion of intra-granular fission gas driven by diffusion. The comparison with currently used models indicates the applicability of the herein developed model within integral fuel performance codes. - Highlights: •Development of a new model for the formation and depletion of the high burnup structure. •New average grain-size measurements to support model development. •Formation threshold of the high burnup structure based on the concept of effective burnup. •Coupled description of grain recrystallization/polygonisation and depletion of intra-granular fission gas. •Model suitable for application in fuel performance codes.

The enhancements and testing for the MCNPX depletion capability

International Nuclear Information System (INIS)

Fensin, M. L.; Hendricks, J. S.; Anghaie, S.

2008-01-01

Monte Carlo-linked depletion methods have gained recent interest due to the ability to more accurately model true system physics and better track the evolution of temporal nuclide inventory by simulating the actual physical process. The integration of INDER90 into the MCNPX Monte Carlo radiation transport code provides a completely self-contained Monte- Carlo-linked depletion capability in a single Monte Carlo code that is compatible with most nuclear criticality (KCODE) particle tracking features in MCNPX. MCNPX depletion tracks all necessary reaction rates and follows as many isotopes as cross section data permits in order to achieve a highly accurate temporal nuclide inventory solution. We describe here the depletion methodology dating from the original linking of MONTEBURNS and MCNP to the first public release of the integrated capability (MCNPX 2. 6.B, June, 2006) that has been reported previously. Then we further detail the many new depletion capability enhancements since then leading to the present capability. The H.B. Robinson benchmark calculation results are also reported. The new MCNPX depletion capability enhancements include: (1) allowing the modeling of as large a system as computer memory capacity permits; (2) tracking every fission product available in ENDF/B VII. 0; (3) enabling depletion in repeated structures geometries such as repeated arrays of fuel pins; (4) including metastable isotopes in burnup; and (5) manually changing the concentrations of key isotopes during different time steps to simulate changing reactor control conditions such as dilution of poisons to maintain criticality during burnup. These enhancements allow better detail to model the true system physics and also improve the robustness of the capability. The H.B. Robinson benchmark calculation was completed in order to determine the accuracy of the depletion solution. Temporal nuclide computations of key actinide and fission products are compared to the results of other

Depletion of energy or depletion of knowledge alternative use of energy resources

International Nuclear Information System (INIS)

Arslan, M.

2011-01-01

This research paper is about the depletion of Energy resources being a huge problem facing the world at this time. As available energy sources are coming to a shortage and measures are be taken in order to conserve the irreplaceable energy resources that leads to sustainability and fair use of energy sources for future generations. Alternative energy sources are being sought; however no other energy source is able to provide even a fraction of energy as that of fossil fuels. Use of the alternative energy resources like wind corridors (Sindh and Baluchistan), fair use of Hydro energy (past monsoon flooding can produce enough energy that may available for next century). Uranium Resources which are enough for centuries energy production in Pakistan (Dhok Pathan Formation) lying in Siwalick series from Pliocene to Pleistocene. Among all of these, my focus is about energy from mineral fuels like Uranium from Sandstone hosted deposits in Pakistan (Siwalik Series in Pakistan). A number of uranium bearing mineralized horizons are present in the upper part of the Dhok Pathan Formation. These horizons have secondary uranium mineral carnotite and other ores. Uranium mineralization is widely distributed throughout the Siwaliks The purpose of this paper was to introduce the use of alternative energy sources in Pakistan which are present in enough amounts by nature. Pakistan is blessed with wealth of natural resources. Unfortunately, Pakistan is totally depending on non renewable energy resource. There are three main types of fossil fuels: coal, oil and natural gas. After food, fossil fuel is humanity's most important source of energy. Pakistan is among the most gas dependent economies of the world. Use of fossil fuel for energy will not only increase the demand of more fossils but it has also extreme effects on climate as well as direct and indirect effects to humans. These entire remedial thinking can only be possible if you try to use alternative energy resources rather than

Analyses of the Anticipated Operational Occurrences for the HANARO Fuel Test Loop

International Nuclear Information System (INIS)

Park, S. K.; Sim, B. S.; Chi, D. Y.; Lee, C. Y.; Ahn, S. H.

2007-12-01

The analyses of anticipated operational occurrences of the HANARO fuel test loop have been carried out by using the MARS/FTL A code, which is a modified version of the MARS code. A critical heat flux correlation on the three rods with triangular array was implemented in the MARS/FTL A code. The correlation was obtained from the critical heat fluxes measured at a test section, which is the same geometry of the in-pile test section of the HANARO fuel test loop. The anticipated operational occurrences of the HANARO fuel test loop are the inadvertent closure of the isolation valves, the over-power transient of the HANARO, the stuck open of the safety valves, and the loss of HANARO class IV power. A minimum DNBR (Departure from Nucleate Boiling Ratio) was predicted in the inadvertent closure of the isolation valves. It is indicated that the minimum DNBR of 1.85 is greater than the design limit DNBR of 1.39. The maximum coolant pressure calculated in the anticipated operational occurrences is also less than the 110 percents of the design pressure

Fuel surveillance and accountability at the Texas A and M NSCR

International Nuclear Information System (INIS)

Feltz, D.E.; Schumache, R.F.; Bates, E.F.

1978-01-01

The record keeping requirements established by CFR 70.51 (SNM balance, inventory, and records requirements), CFR 70.53 (SNM status report) and fuel surveillance as per technical specifications have been simplified for the NSCR by use of a small computer (DEC PDP 11/05). Prior to its utilization, fuel records and calculation of fuel depletion were performed by hand and required several man-hours. A fuel inventory program for the determination of fuel depletion, beginning and ending inventory, and records of fuel location, steady state energy and pulse energy on an element by element basis is presented. The requirement of plutonium production calculations and reporting for the NSCR is also discussed. (author)

Research on calculation of mixing fraction for natural uranium equivalent fuel

International Nuclear Information System (INIS)

Huang Shien; Wang Lianjie; Wei Yanqin; Li Qing; Zheng Jiye

2013-01-01

Based on the first-order perturbation theory and reasonable approximations, the calculation method of recycled uranium (RU) and depleted uranium (DU) mixing fraction for natural uranium equivalent (NUE) fuel was studied, so the equivalence between NUE fuel and natural uranium (NU) fuel was assured. The adopted calculation method accurately takes the variation of micro cross sections alone with fuel depletion into account. A computer code named ALPHA was programmed to execute the calculation procedure. Then the ALPHA code and the WIMS-AECL code compose a processing system, which is applicable to the mixing fraction calculation for heavy water reactor NUE fuel. The validation shows that the processing system can accurately calculate the mixing fraction for NUE fuel. (authors)

Evaluating and Addressing Potential Hazards of Fuel Tanks Surviving Atmospheric Reentry

Science.gov (United States)

Kelley, Robert L.; Johnson, Nicholas L.

2011-01-01

In order to ensure reentering spacecraft do not pose an undue risk to the Earth's population it is important to design satellites and rocket bodies with end of life considerations in mind. In addition to considering the possible consequences of deorbiting a vehicle, consideration must also be given to the possible risks associated with a vehicle failing to become operational or reach its intended orbit. Based on recovered space debris and numerous reentry survivability analyses, fuel tanks are of particular concern in both of these considerations. Most spacecraft utilize some type of fuel tank as part of their propulsion system. These fuel tanks are most often constructed using stainless steel or titanium and are filled with potentially hazardous substances such as hydrazine and nitrogen tetroxide. For a vehicle which has reached its scheduled end of mission the contents of the tanks are typically depleted. In this scenario the use of stainless steel and titanium results in the tanks posing a risk to people and property do to the high melting point and large heat of ablation of these materials leading to likely survival of the tank during reentry. If a large portion of the fuel is not depleted prior to reentry, there is the added risk of hazardous substance being released when the tank impact the ground. This paper presents a discussion of proactive methods which have been utilized by NASA satellite projects to address the risks associated with fuel tanks reentering the atmosphere. In particular it will address the design of a demiseable fuel tank as well as the evaluation of off the shelf designs which are selected to burst during reentry.

Comparison based on energy and exergy analyses of the potential cogeneration efficiencies for fuel cells and other electricity generation devices

Energy Technology Data Exchange (ETDEWEB)

Rosen, M A [Ryerson Polytechnical Inst., Toronto, (CA). Dept. of Mechanical Engineering

1990-01-01

Comparisons of the potential cogeneration efficiencies are made, based on energy and exergy analyses, for several devices for electricity generation. The investigation considers several types of fuel cell system (Phosphoric Acid, Alkaline, Solid Polymer Electrolyte, Molten Carbonate and Solid Oxide), and several fossil-fuel and nuclear cogeneration systems based on steam power plants. In the analysis, each system is modelled as a device for which fuel and air enter, and electrical- and thermal-energy products and material and thermal-energy wastes exit. The results for all systems considered indicate that exergy analyses should be used when analysing the cogeneration potential of systems for electricity generation, because they weigh the usefulnesses of heat and electricity on equivalent bases. Energy analyses tend to present overly optimistic views of performance. These findings are particularly significant when large fractions of the heat output from a system are utilized for cogeneration. (author).

MOSRA-SRAC. Lattice calculation module of the modular code system for nuclear reactor analyses MOSRA

International Nuclear Information System (INIS)

Okumura, Keisuke

2015-10-01

MOSRA-SRAC is a lattice calculation module of the Modular code System for nuclear Reactor Analyses (MOSRA). This module performs the neutron transport calculation for various types of fuel elements including existing light water reactors, research reactors, etc. based on the collision probability method with a set of the 200-group cross-sections generated from the Japanese Evaluated Nuclear Data Library JENDL-4.0. It has also a function of the isotope generation and depletion calculation for up to 234 nuclides in each fuel material in the lattice. In these ways, MOSRA-SRAC prepares the burn-up dependent effective microscopic and macroscopic cross-section data to be used in core calculations. A CD-ROM is attached as an appendix. (J.P.N.)

Depleted uranium oxides as spent-nuclear-fuel waste-package fill materials

International Nuclear Information System (INIS)

Forsberg, C.W.

1997-01-01

Depleted uranium dioxide fill inside the waste package creates the potential for significant improvements in package performance based on uranium geochemistry, reduces the potential for criticality in a repository, and consumes DU inventory. As a new concept, significant uncertainties exist: fill properties, impacts on package design, post- closure performance

Irradiation behavior of uranium-silicide dispersion fuels

International Nuclear Information System (INIS)

Hofman, G.L.; Neimark, L.A.

1984-01-01

This paper describes and analyzes the irradiation behavior of experimental fuel plates containing U 3 Si, U 3 Si-1.5 w/o Al, and U 3 Si 2 particulate fuel dispersed and clad in aluminum. The fuel is nominally 19.9%-enriched 235 U and the fuel volume fraction in the central ''meat'' section of the plates is approximately 33%. Sets of fuel plates were removed from the Oak Ridge Research reactor at burnup levels of 35, 83, and 94% 235 U depletion and examined at the Alpha-Gamma Hot-Cell Facility at Argonne National Laboratory. The results of the examination may be summarized as follows. The dimensional stability of the U 3 Si 2 and pure U 3 Si fuel was excellent throughout the entire burnup range, with uniform plate thickness increases up to a maximum of 4 mils at the highest burnup level (94% 235 U depletion). This corresponds to a meat volume increase of 11%. The swelling was partially due to solid fission products but to a larger extent to fission gas bubbles. The fission gas bubbles in U 3 Si 2 were small (submicrometer size) and very uniformly distributed, indicating great stability. To a large extent this was also the case for U 3 Si; however, larger bubbles ( 3 Si-1.5 w/o Al fuel became unstable at the higher burnup levels. Fission gas bubbles were larger than in the other two fuels and were present throughout the fuel particles. At 94% 235 U depletion, the formation of fission gas bubbles with diameters up to 20 mils caused the plates to pillow. It is proposed that aluminum in U 3 Si destabilizes fission gas bubble formation to the point of severe breakaway swelling in the prealloyed silicide fuel. (author)

Depleted uranium

International Nuclear Information System (INIS)

Huffer, E.; Nifenecker, H.

2001-02-01

This document deals with the physical, chemical and radiological properties of the depleted uranium. What is the depleted uranium? Why do the military use depleted uranium and what are the risk for the health? (A.L.B.)

CANDU flexible and economical fuel technology in China

Energy Technology Data Exchange (ETDEWEB)

Mingjun, C. [CNNC Nuclear Power Operation Management Co., Zhejiang (China); Zhenhua, Z.; Zhiliang, M. [CNNC Third Qinshan Nuclear Power Co., Zhejiang (China); Cottrell, C.M.; Kuran, S. [Candu Energy Inc., Mississauga, ON (Canada)

2014-07-01

Use in CANDU reactor is one good option of recycled uranium (RU) and thorium (Th) resource. It is also good economy to CANDU fuel. Since 2008 Qinshan CANDU Plant and our partners (Candu Energy and CNNC and NPIC) have made great efforts to develop the engineering technologies of Flexible and Economical Fuel (RU and Th) in CANDU type reactor and finding the CANDU's position in Chinese closed fuel cycle (CFC) system. This paper presents a proposal of developing strategy and implementation plan. Qinshan CANDU reactors will be converted to use recycled and depleted uranium based fuels, a first-of-its-kind. The fuel is composed of both recycled and depleted uranium and simulating natural uranium behavior. This paper discusses its development, design, manufacture and verification tested with success and the full core implementation plan by the end of 2014. (author)

Kr ion irradiation study of the depleted-uranium alloys

Energy Technology Data Exchange (ETDEWEB)

Gan, J., E-mail: Jian.Gan@inl.go [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-6188 (United States); Keiser, D.D. [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-6188 (United States); Miller, B.D. [University of Wisconsin, 1500 Engineering Drive, Madison, WI 53706 (United States); Kirk, M.A.; Rest, J. [Argonne National Laboratory, 9700 South Cass Ave., Argonne, IL 60439 (United States); Allen, T.R. [University of Wisconsin, 1500 Engineering Drive, Madison, WI 53706 (United States); Wachs, D.M. [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-6188 (United States)

2010-12-01

Fuel development for the reduced enrichment research and test reactor (RERTR) program is tasked with the development of new low enrichment uranium nuclear fuels that can be employed to replace existing high enrichment uranium fuels currently used in some research reactors throughout the world. For dispersion type fuels, radiation stability of the fuel-cladding interaction product has a strong impact on fuel performance. Three depleted-uranium alloys are cast for the radiation stability studies of the fuel-cladding interaction product using Kr ion irradiation to investigate radiation damage from fission products. SEM analysis indicates the presence of the phases of interest: U(Al, Si){sub 3}, (U, Mo)(Al, Si){sub 3}, UMo{sub 2}Al{sub 20}, U{sub 6}Mo{sub 4}Al{sub 43} and UAl{sub 4}. Irradiations of TEM disc samples were conducted with 500 keV Kr ions at 200 {sup o}C to ion doses up to 2.5 x 10{sup 19} ions/m{sup 2} ({approx}10 dpa) with an Kr ion flux of 10{sup 16} ions/m{sup 2}/s ({approx}4.0 x 10{sup -3} dpa/s). Microstructural evolution of the phases relevant to fuel-cladding interaction products was investigated using transmission electron microscopy.

Safety analyses for sodium-cooled fast reactors with pelletized and sphere-pac oxide fuels within the FP-7 European project PELGRIMM - 15386

International Nuclear Information System (INIS)

Maschek, W.; Andriolo, L.; Matzerath-Boccaccini, C.; Delage, F.; Parisi, C.; Del Nevo, A.; Abbate, G.; Schmitt, D.

2015-01-01

The European FP-7 project PELGRIMM addresses the development of Minor-Actinide (MA) bearing oxide fuel for Sodium-cooled Fast Reactors. Optionally, both MA homogeneous recycling and heterogeneous recycling is investigated with pellet and sphere-pac fuel. A first safety assessment of sphere-pac fuelled cores should be given in the Work Package 4 of the project. This assessment is in continuity with the former FP-7 CP-ESFR project. Within the CP-ESFR project the CONF2 core design has been developed characterized by a core with a large upper sodium plenum to reduce the coolant void worth. This optimized core has been chosen for the safety analyses in PELGRIMM. The task within the PELGRIMM project is thus a safety assessment of the CONF2 core loaded either with pellets or with sphere-pac fuel. The investigations started with the design of the CONF2 core with sphere-pac fuel and the determination of core safety parameters and burn-up behavior. The neutronic analyses have been performed with the MCNPX code. Variants of the CONF2 core contain up to 4% Am in the fuel. The results revealed an extended void worth (core + upper plenum) for an Am free core of 1 up to 3 dollars for the 4% Am core. Thermal-hydraulic design analyses have been performed by RELAP5-3D. The accident simulations should be performed by different codes, some of which focus on the initiation phase of the accident, as SAS4A, BELLA and the MAT5DYN code, whereas the SIMMER-III code will also deal with the later accident phases and a potential whole core melting. The codes had to be adapted to the specifics of the sphere-pac fuel, in particular to the thermal conductivity and gap conditions. Analyses showed that the safety assessment has to take into account two main phases. Starting up the core, the green fuel shows a reduced fuel thermal conductivity. After restructuring within a couple of hours, the thermal conductivity recovers and the fuel temperature decreases. The main objective of the safety analyses

Studies on the safety and transmutation behaviour of innovative fuels for light water reactors; Untersuchungen zum Sicherheits- und Transmutationsverhalten innovativer Brennstoffe fuer Leichtwasserreaktoren

Energy Technology Data Exchange (ETDEWEB)

Schitthelm, Oliver

2012-07-01

Nuclear power plants contribute a substantial part to the energy demand in industry. Today the most common fuel cycle uses enriched uranium which produces plutonium due to its {sup 238}U content. With respect to the long-term waste disposal Plutonium is an issue due to its heat production and radiotoxicity. This thesis consists of three main parts. In the first part the development and validation of a new code package MCBURN for spatial high resolution burnup simulations is presented. In the second part several innovative uranium-free and plutonium-burning fuels are evaluated on assembly level. Candidates for these fuels are a thorium/plutonium fuel and an inert matrix fuel consisting of plutonium dispersed in an enriched molybdenum matrix. The performance of these fuels is evaluated against existing MOX and enriched uranium fuels considering the safety and transmutation behaviour. The evaluation contains the boron efficiency, the void coefficient, the doppler coefficient and the net balances of every radionuclide. In the third part these innovative fuels are introduced into a German KONVOI reactor core. Considering todays approved usage of MOX fuels a partial loading of one third of innovative fuels and two third of classical uranium fuels was analysed. The efficiency of the plutonium depletion is determined by the ratio of the production of higher isotopes compared to the plutonium depletion. Todays MOX-fuels transmutate about 25% to 30% into higher actinides as Americium or Curium. In uranium-free fuels this ratio is about 10% due to the lack of additional plutonium production. The analyses of the reactor core have shown that one third of MOX fuel is not capable of a net reduction of plutonium. On the other hand a partial loading with thorium/plutonium fuel incinerates about half the amount of plutonium produced by an uranium only core. If IMF is used the ratio increases to about 75%. Considering the safety behavior all fuels have shown comparable results.

Code package to analyse behavior of the WWER fuel rods in normal operation: TOPRA's code

International Nuclear Information System (INIS)

Scheglov, A.; Proselkov, V.

2001-01-01

This paper briefly describes the code package intended for analysis of WWER fuel rod characteristics. The package includes two computer codes: TOPRA-1 and TOPRA-2 for full-scale fuel rod analyses; MRZ and MKK codes for analyzing the separate sections of fuel rods in r-z and r-j geometry. The TOPRA's codes are developed on the base of PIN-mod2 version and verified against experimental results obtained in MR, MIR and Halden research reactors (in the framework of SOFIT, FGR-2 and FUMEX experimental programs). Comparative analysis of calculation results and results from post-reactor examination of the WWER-440 and WWER-1000 fuel rod are also made as additional verification of these codes. To avoid the enlarging of uncertainties in fuel behavior prediction as a result of simplifying of the fuel geometry, MKK and MRZ codes are developed on the basis of the finite element method with use of the three nodal finite elements. Results obtained in the course of the code verification indicate the possibility for application of the method and TOPRA's code for simplified engineering calculations of WWER fuel rods thermal-physical parameters. An analysis of maximum relative errors for predicting of the fuel rod characteristics in the range of the accepted parameter values is also presented in the paper

Nuclear fuel production

International Nuclear Information System (INIS)

Randol, A.G.

1985-01-01

The production of new fuel for a power plant reactor and its disposition following discharge from the power plant is usually referred to as the ''nuclear fuel cycle.'' The processing of fuel is cyclic in nature since sometime during a power plant's operation old or ''depleted'' fuel must be removed and new fuel inserted. For light water reactors this step typically occurs once every 12-18 months. Since the time required for mining of the raw ore to recovery of reusable fuel materials from discharged materials can span up to 8 years, the management of fuel to assure continuous power plant operation requires simultaneous handling of various aspects of several fuel cycles, for example, material is being mined for fuel to be inserted in a power plant 2 years into the future at the same time fuel is being reprocessed from a discharge 5 years prior. Important aspects of each step in the fuel production process are discussed

Ego Depletion Does Not Interfere With Working Memory Performance.

Science.gov (United States)

Singh, Ranjit K; Göritz, Anja S

2018-01-01

Ego depletion happens if exerting self-control reduces a person's capacity to subsequently control themselves. Previous research has suggested that ego depletion not only interferes with subsequent self-control but also with working memory. However, recent meta-analytical evidence casts doubt onto this. The present study tackles the question if ego depletion does interfere with working memory performance. We induced ego depletion in two ways: using an e-crossing task and using a Stroop task. We then measured working memory performance using the letter-number sequencing task. There was no evidence of ego depletion interfering with working memory performance. Several aspects of our study render this null finding highly robust. We had a large and heterogeneous sample of N = 1,385, which provided sufficient power. We deployed established depletion tasks from two task families (e-crossing task and Stroop), thus making it less likely that the null finding is due to a specific depletion paradigm. We derived several performance scores from the working memory task and ran different analyses to maximize the chances of finding an effect. Lastly, we controlled for two potential moderators, the implicit theories about willpower and dispositional self-control capacity, to ensure that a possible effect on working memory is not obscured by an interaction effect. In sum, this experiment strengthens the position that ego depletion works but does not affect working memory performance.

Analyses of the Anticipated Operational Occurrences for the HANARO Fuel Test Loop

Energy Technology Data Exchange (ETDEWEB)

Park, S. K.; Sim, B. S.; Chi, D. Y.; Lee, C. Y.; Ahn, S. H

2007-12-15

The analyses of anticipated operational occurrences of the HANARO fuel test loop have been carried out by using the MARS/FTL{sub A} code, which is a modified version of the MARS code. A critical heat flux correlation on the three rods with triangular array was implemented in the MARS/FTL{sub A} code. The correlation was obtained from the critical heat fluxes measured at a test section, which is the same geometry of the in-pile test section of the HANARO fuel test loop. The anticipated operational occurrences of the HANARO fuel test loop are the inadvertent closure of the isolation valves, the over-power transient of the HANARO, the stuck open of the safety valves, and the loss of HANARO class IV power. A minimum DNBR (Departure from Nucleate Boiling Ratio) was predicted in the inadvertent closure of the isolation valves. It is indicated that the minimum DNBR of 1.85 is greater than the design limit DNBR of 1.39. The maximum coolant pressure calculated in the anticipated operational occurrences is also less than the 110 percents of the design pressure.

Fuel utilization potential in light water reactors with once-through fuel irradiation (AWBA Development Program)

International Nuclear Information System (INIS)

Rampolla, D.S.; Conley, G.H.; Candelore, N.R.; Cowell, G.K.; Estes, G.P.; Flanery, B.K.; Duncombe, E.; Dunyak, J.; Satterwhite, D.G.

1979-07-01

Current commercial light water reactor cores operate without recylce of fuel, on a once-through fuel cycle. To help conserve the limited nuclear fuel resources, there is interest in increasing the energy yield and, hence, fuel utilization from once-through fuel irradiation. This report evaluates the potential increase in fuel utilization of light water reactor cores operating on a once-through cycle assuming 0.2% enrichment plant tails assay. This evaluation is based on a large number of survey calculations using techniques which were verified by more detailed calculations of several core concepts. It is concluded that the maximum fuel utilization which could be achieved by practical once-through pressurized light water reactor cores with either uranium or thorium is about 17 MWYth/ST U 3 O 8 (Megawatt Years Thermal per Short Ton of U 3 O 8 ). This is about 50% higher than that of current commercial light water reactor cores. Achievement of this increased fuel utilization would require average fuel burnup beyond 50,000 MWD/MT and incorporation of the following design features to reduce parasitic losses of neutrons: reflector blankets to utilize neutrons that would otherwise leak out of the core; fuel management practices in which a smaller fraction of the core is replaced at each refueling; and neutron economic reactivity control, such as movable fuel control rather than soluble boron control. For a hypothetical situation in which all neutron leakage and parasitic losses are eliminated and fuel depletion is not limited by design considerations, a maximum fuel utilization of about 20 MWYth/ST U 3 O 8 is calculated for either uranium or thorium. It is concluded that fuel utilization for comparable reactor designs is better with uranium fuel than with thorium fuel for average fuel depletions of 30,000 to 35,000 MWD/MT which are characteristic of present light water reactor cores

Burnup credit feasibility for BWR spent fuel shipments

International Nuclear Information System (INIS)

Broadhead, B.L.

1990-01-01

Considerable interest in the allowance of reactivity credit for the exposure history of power reactor fuel currently exists. This ''burnup credit'' issue has the potential to greatly reduce risk and cost when applied to the design and certification of spent of fuel casks used for transportation and storage. Analyses 1 have shown the feasibility estimated the risk and economic incentives for allowing burnup credit in pressurized water reactor (PWR) spent fuel shipping cask applications. This paper summarizes the extension of the previous PWR feasibility assessments to boiling water reactor (BWR) fuel. As with the PWR analysis, the purpose was not verification of burnup credit (see ref. 2 for ongoing work in this area) but a reasonable assessment of the feasibility and potential gains from its use in BWR applications. This feasibility analysis aims to apply simple methods that adequately characterize the time-dependent isotopic compositions of typical BWR fuel. An initial analysis objective was to identify a simple and reliable method for characterizing BWR spent fuel. The method includes characterization of a typical pin-cell spectrum, using a one-dimensional (1-D) model of a BWR assembly. The calculated spectrum allows burnup-dependent few-group material constants to be generated. Point depletion methods were then used to obtain the time-varying characteristics of the fuel. These simple methods were validated, where practical, with multidimensional methods. 6 refs., 1 tab

Update and evaluation of decay data for spent nuclear fuel analyses

Science.gov (United States)

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.

Update and evaluation of decay data for spent nuclear fuel analyses

Directory of Open Access Journals (Sweden)

Simeonov Teodosi

2017-01-01

Full Text Available 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.

Assessment of microalgae biodiesel fuels using a fuel property estimation methodology

Energy Technology Data Exchange (ETDEWEB)

Torrens, Jonas Colen Ladeia; Vargas, Jose Viriato Coelho; Mariano, Andre Bellin [Center for Research and Development of Sustainable Energy. Universidade Federal do Parana, Curitiba, PR (Brazil)

2010-07-01

Recently, depleting supplies of petroleum and the concerns about global warming are drawing attention to alternative sources of energy. In this context, advanced biofuels, derived from non edible superior plants and microorganisms, are presented as promising options for the transportation sector. Biodiesel, which is the most prominent alternative fuel for compression ignition engines, have a large number as potential feedstock, such as plants (e.g., soybean, canola, palm) and microorganism (i.e., microalgae, yeast, fungi and bacterium). In order to determine their potential, most studies focus on the economic viability, but few discuss the technical viability of producing high quality fuels from such feedstock. Since the fuel properties depend on the composition of the parent oil, and considering the variability of the fatty acid profile found in these organisms, it is clear that the fuels derived may present undesirable properties, e.g., high viscosity, low cetane number, low oxidative stability and poor cold flow properties. Therefore, it is very important to develop ways of analysing the fuel quality prior to production, specially considering the high cost of producing and testing several varieties of plants and microorganisms. In this aim, this work presents the use of fuel properties estimation methods on the assessment of the density, viscosity, cetane number and cold filter plugging point of several microalgae derived biofuels, comparing then to more conventional biodiesel fuels. The information gathered with these methods helps on the selection of species and cultivation parameters, which have a high impact on the derived fuel quality, and have been successfully employed on the Center for Research and Development of Sustainable Energy. The results demonstrate that some species of microalgae have the potential to produce high quality biodiesel if cultivated with optimised conditions, associated with the possibility of obtaining valuable long chain

Depleted uranium hexafluoride management program : data compilation for the Paducah site

International Nuclear Information System (INIS)

Hartmann, H.

2001-01-01

This report is a compilation of data and analyses for the Paducah site, near Paducah, Kentucky. The data were collected and the analyses were done in support of the U.S. Department of Energy (DOE) 1999 Programmatic Environmental Impact Statement for Alternative Strategies for the Long-Term Management and Use of Depleted Uranium Hexafluoride (DOE/EIS-0269). The report describes the affected environment at the Paducah site and summarizes potential environmental impacts that could result from conducting the following depleted uranium hexafluoride (UF 6 ) activities at the site: continued cylinder storage, preparation of cylinders for shipment, conversion, and long-term storage. DOE's preferred alternative is to begin converting the depleted UF 6 inventory as soon as possible to either uranium oxide, uranium metal, or a combination of both, while allowing for use of as much of this inventory as possible

Depleted uranium hexafluoride management program : data compilation for the Portsmouth site

International Nuclear Information System (INIS)

Hartmann, H. M.

2001-01-01

This report is a compilation of data and analyses for the Portsmouth site, near Portsmouth, Ohio. The data were collected and the analyses were done in support of the U.S. Department of Energy (DOE) 1999 Programmatic Environmental Impact Statement for Alternative Strategies for the Long-Term Management and Use of Depleted Uranium Hexafluoride (DOE/EIS-0269). The report describes the affected environment at the Portsmouth site and summarizes potential environmental impacts that could result from conducting the following depleted uranium hexafluoride (UF 6 ) management activities at the site: continued cylinder storage, preparation of cylinders for shipment, conversion, and long-term storage. DOE's preferred alternative is to begin converting the depleted UF 6 inventory as soon as possible to either uranium oxide, uranium metal, or a combination of both, while allowing for use of as much of this inventory as possible

Analyses of the internal structure of the oscillating vibro-packed fuels by the micro focus X-rays CT method

International Nuclear Information System (INIS)

Mizuta, Yasutoshi

2003-02-01

The purpose of this study is to support the development of vibro-packed fuel technology at Japan Nuclear Cycle Development Institute. 3-dimensional (3-D) data was built from the multi-cross sectional images obtained by the micro focus X-rays CT method in the vibro-packed fuel models. The structural analyses were carried out about the obtained 3-D CT images. The packing-rate distribution and the density distribution were measured as well as the number distribution of particles, etc. Consequently, it is obtained that vibrate conditions and a vibrating state have strong correlation, and it is also shown that the 3-D analyses of the internal structure by the micro focus X-rays CT method are effective in performance evaluation of vibro-packed fuels. (author)

Depleted uranium concrete container feasibility study

International Nuclear Information System (INIS)

Haelsig, R.T.

1994-09-01

The purpose of this report is to consider the feasibility of using containers constructed of depleted uranium aggregate concrete (DUCRETE) to store and transport radioactive materials. The method for this study was to review the advantages and disadvantages of DUCRETE containers considering design requirements for potential applications. The author found that DUCRETE is a promising material for onsite storage containers, provided DUCRETE vessels can be certified for one-way transport to disposal sites. The author also found that DUCRETE multipurpose spent nuclear fuel storage/transport packages are technically viable, provided altered temperature acceptance limits can be developed for DUCRETE

Processing and Applications of Depleted Uranium Alloy Products

Science.gov (United States)

1976-09-01

ammunition, weapons, gyrorotors, and ballast. Depleted uranium used in fly- wheel devices, nuclear fuel casks, and ammunition could consume a significant...from straight in the range of 0,002 to 0.060-inch TIR (total indicated runout ) with an average of 0.025-inch TIR.* Solution heat treatment of the as-cast...an envelope thickness of 0.050 inch to allow for runout and to clean up surface imperfections. The runout resulting from heat treatment was in the

Efficiency analyses of the CANDU spent fuel repository using modified disposal canisters for a deep geological disposal system design

International Nuclear Information System (INIS)

Lee, J.Y.; Cho, D.K.; Lee, M.S.; Kook, D.H.; Choi, H.J.; Choi, J.W.; Wang, L.M.

2012-01-01

Highlights: ► A reference disposal concept for spent nuclear fuels in Korea has been reviewed. ► To enhance the disposal efficiency, alternative disposal concepts were developed. ► Thermal analyses for alternative disposal concepts were performed. ► From the result of the analyses, the disposal efficiency of the concepts was reviewed. ► The most effective concept was suggested. - Abstract: Deep geological disposal concept is considered to be the most preferable for isolating high-level radioactive waste (HLW), including nuclear spent fuels, from the biosphere in a safe manner. The purpose of deep geological disposal of HLW is to isolate radioactive waste and to inhibit its release of for a long time, so that its toxicity does not affect the human beings and the biosphere. One of the most important requirements of HLW repository design for a deep geological disposal system is to keep the buffer temperature below 100 °C in order to maintain the integrity of the engineered barrier system. In this study, a reference disposal concept for spent nuclear fuels in Korea has been reviewed, and based on this concept, efficient alternative concepts that consider modified CANDU spent fuels disposal canister, were developed. To meet the thermal requirement of the disposal system, the spacing of the disposal tunnels and that of the disposal pits for each alternative concept, were drawn following heat transfer analyses. From the result of the thermal analyses, the disposal efficiency of the alternative concepts was reviewed and the most effective concept suggested. The results of these analyses can be used for a deep geological repository design and detailed analyses, based on exact site characteristics data, will reduce the uncertainty of the results.

Depletion of liver glutathione levels in rats: a potential confound of nose-only inhalation.

Science.gov (United States)

Fechter, Laurence D; Nelson-Miller, Alisa; Gearhart, Caroline

2008-07-01

Nose-only inhalation exposure chambers offer key advantages to whole-body systems, particularly when aerosol or mixed aerosol-vapor exposures are used. Specifically, nose-only chambers provide enhanced control over the route of exposure and dose by minimizing the deposition of particles either on the subjects skin/fur or on surfaces of a whole-body exposure system. In the current series of experiments, liver, brain, and lung total glutathione (GSH) levels were assessed following either nose-only or whole-body exposures to either jet fuel or to clean, filtered air. The data were compared to untreated control subjects. Acute nose-only inhalation exposures of rats resulted in a significant depletion of liver GSH levels both in subjects that were exposed to clean, filtered air as well as those exposed to JP-8 jet fuel and to a synthetic jet fuel. Glutathione levels were not altered in lung or brain tissue. Whole-body inhalation exposure had no effect on GSH levels in any tissue for any of the treatment groups. A second experiment demonstrated that the loss of GSH did not occur if rats were anaesthetized prior to and during nose-only exposure to clean, filtered air or to mixed hydrocarbons. These data appear to be consistent with studies demonstrating depletion in liver GSH levels among rats subjected to restraint stress. Finally, the depletion of GSH that was observed in liver following a single acute exposure was reduced following five daily exposures to clean, filtered air, suggesting the possibility of habituation to restraint in the nose-only exposure chamber. The finding that placement in a nose-only exposure chamber per se yields liver GSH depletion raises the possibility of an interaction between this mode of toxicant exposure and the toxicological effects of certain inhaled test substances.

Improved analysis on multiple recycling of fuel in prototype fast ...

Indian Academy of Sciences (India)

2011-08-03

Aug 3, 2011 ... ENDF/B-VII.0, and with the most recent specification of the fuel composition ... Fast breeder reactors; closed fuel cycle; fuel production and depletion; reprocessing .... set including self-shielding factors (SSF, i.e. self-shielded to ...

Bio fuels. A comparative analysis; Biokraftstoffe. Eine vergleichende Analyse

Energy Technology Data Exchange (ETDEWEB)

Schmitz, Norbert; Henke, Jan; Klepper, Gernot

2009-07-01

The market for bio fuels is subject to very high dynamics worldwide. Due to the extreme rise of the prices of raw materials as well as due to the retrogressive tax reductions for bio fuels in Germany one hardly invests in bio fuels. Substantial changes are experienced in the markets for fossil raw materials. The prices for agrarian raw material used in this contribution originate from the years 2006 and 2007. The effects of clearly higher oil prices on the bio fuel market are described. The investigation under consideration also deals with criteria of sustainability. The contribution of the individual bio fuels to the reduction of greenhouse gases is analyzed. The costs resulting from this are numerated. This enables a well-established comparison in which less representative bio fuels such as bio methane, BtL fuels and cellulose ethanol also are included.

COMRAD96, Nuclear Fuel Burnup and Depletion Calculation System

International Nuclear Information System (INIS)

Suyama, K.; Masukawa, F.; Ido, M.; Enomoto, M.; Takyu, S.; Hara, T.

2002-01-01

1 - Description of program or function: Burn-up calculation of nuclear fuel. 2 - Methods: Matrix exponential method, Bateman Equation. 3 - Restrictions on the complexity of the problem: a) One-grouped cross section library should be prepared for the fuel system to be analyzed using UNITBURN. However, UNITBURN is not available now for UNIX systems. b) Gamma ray spectrometry calculation will fail using the attached piflib routine. This problem has already been rectified in the internal version. 4 - Typical running time: Two minutes for standard burn-up calculation on Sun ULTRA 30. 5 - Unusual features - a) Selection of Matrix exponential method, or Bateman Equation. b) JDDL, a detailed decay chain data based on ENSDF. 6 - Related or auxiliary programs: UNITBURN: Burnup calculation code unit cell system

chemical determination of burnup ratio in nuclear fuels

International Nuclear Information System (INIS)

Guereli, L.

1997-01-01

Measurements of the extent of fission are important to determine the irradiation performance of a nuclear fuel. The energy released per unit mass of uranium (burnup) can be determined from measurement of the percent of heavy atoms that have fissioned during irradiation.The preferred method for this determination is choosing a suitable fission monitor (usually ''1''4''8Nd) and its determination after separation from the fuel matrix. In thermal reactor fuels where the only heavy element in the starting material is uranium, uranium depletion can be used for burnup determination. ''2''3''5U depletion method requires measurement of uranium isotopic ratios of both irradiated and unirradiated fuel. Isotopic ratios can be determined by thermal ionization mass spectrometer following separation of uranium from the fuel matrix. Separation procedures include solvent extraction, ion exchange and anion exchange chromatography. Another fission monitor used is ''1''3''9La determination by HPLC. Because La is monoisotopic (''1''3''9La) in the fuel, it can be determined by chemical analysis techniques

Code Analyses Supporting PIE of Weapons-Grade MOX Fuel

International Nuclear Information System (INIS)

Ott, Larry J.; Bevard, Bruce Balkcom; Spellman, Donald J.; McCoy, Kevin

2010-01-01

The U.S. Department of energy has decided to dispose of a portion of the nation's surplus weapons-grade plutonium by reconstituting it into mixed oxide (MOX) fuel and irradiating the fuel in commercial power reactors. Four lead test assemblies (LTAs) were manufactured with weapons-grade mixed oxide (WG-MOX) fuel and irradiated in the Catawba Nuclear Station Unit 1, to a maximum fuel rod burnup of ∼47.3 GWd/MTHM. As part of the fuel qualification process, five rods with varying burnups and initial plutonium contents were selected from one assembly and shipped to the Oak Ridge National Laboratory (ORNL) for hot cell examination. ORNL has provided analytical support for the post-irradiation examination (PIE) of these rods via extensive fuel performance modeling which has aided in instrument settings and PIE data interpretation. The results of these fuel performance simulations are compared in this paper with available PIE data.

Performance Analyses of Renewable and Fuel Power Supply Systems for Different Base Station Sites

Directory of Open Access Journals (Sweden)

Josip Lorincz

2014-11-01

Full Text Available Base station sites (BSSs powered with renewable energy sources have gained the attention of cellular operators during the last few years. This is because such “green” BSSs impose significant reductions in the operational expenditures (OPEX of telecom operators due to the possibility of on-site renewable energy harvesting. In this paper, the green BSSs power supply system parameters detected through remote and centralized real time sensing are presented. An implemented sensing system based on a wireless sensor network enables reliable collection and post-processing analyses of many parameters, such as: total charging/discharging current of power supply system, battery voltage and temperature, wind speed, etc. As an example, yearly sensing results for three different BSS configurations powered by solar and/or wind energy are discussed in terms of renewable energy supply (RES system performance. In the case of powering those BSS with standalone systems based on a fuel generator, the fuel consumption models expressing interdependence among the generator load and fuel consumption are proposed. This has allowed energy-efficiency comparison of the fuel powered and RES systems, which is presented in terms of the OPEX and carbon dioxide (CO2 reductions. Additionally, approaches based on different BSS air-conditioning systems and the on/off regulation of a daily fuel generator activity are proposed and validated in terms of energy and capital expenditure (CAPEX savings.

VOC composition of current motor vehicle fuels and vapors, and collinearity analyses for receptor modeling.

Science.gov (United States)

Chin, Jo-Yu; Batterman, Stuart A

2012-03-01

The formulation of motor vehicle fuels can alter the magnitude and composition of evaporative and exhaust emissions occurring throughout the fuel cycle. Information regarding the volatile organic compound (VOC) composition of motor fuels other than gasoline is scarce, especially for bioethanol and biodiesel blends. This study examines the liquid and vapor (headspace) composition of four contemporary and commercially available fuels: gasoline (gasoline), ultra-low sulfur diesel (ULSD), and B20 (20% soy-biodiesel and 80% ULSD). The composition of gasoline and E85 in both neat fuel and headspace vapor was dominated by aromatics and n-heptane. Despite its low gasoline content, E85 vapor contained higher concentrations of several VOCs than those in gasoline vapor, likely due to adjustments in its formulation. Temperature changes produced greater changes in the partial pressures of 17 VOCs in E85 than in gasoline, and large shifts in the VOC composition. B20 and ULSD were dominated by C(9) to C(16)n-alkanes and low levels of the aromatics, and the two fuels had similar headspace vapor composition and concentrations. While the headspace composition predicted using vapor-liquid equilibrium theory was closely correlated to measurements, E85 vapor concentrations were underpredicted. Based on variance decomposition analyses, gasoline and diesel fuels and their vapors VOC were distinct, but B20 and ULSD fuels and vapors were highly collinear. These results can be used to estimate fuel related emissions and exposures, particularly in receptor models that apportion emission sources, and the collinearity analysis suggests that gasoline- and diesel-related emissions can be distinguished. Copyright © 2011 Elsevier Ltd. All rights reserved.

Analysis of uncertainty propagation in nuclear fuel cycle scenarios

International Nuclear Information System (INIS)

Krivtchik, Guillaume

2014-01-01

Nuclear scenario studies model nuclear fleet over a given period. They enable the comparison of different options for the reactor fleet evolution, and the management of the future fuel cycle materials, from mining to disposal, based on criteria such as installed capacity per reactor technology, mass inventories and flows, in the fuel cycle and in the waste. Uncertainties associated with nuclear data and scenario parameters (fuel, reactors and facilities characteristics) propagate along the isotopic chains in depletion calculations, and through out the scenario history, which reduces the precision of the results. The aim of this work is to develop, implement and use a stochastic uncertainty propagation methodology adapted to scenario studies. The method chosen is based on development of depletion computation surrogate models, which reduce the scenario studies computation time, and whose parameters include perturbations of the depletion model; and fabrication of equivalence model which take into account cross-sections perturbations for computation of fresh fuel enrichment. Then the uncertainty propagation methodology is applied to different scenarios of interest, considering different options of evolution for the French PWR fleet with SFR deployment. (author) [fr

CESAR5.3: Isotopic depletion for Research and Testing Reactor decommissioning

Science.gov (United States)

Ritter, Guillaume; Eschbach, Romain; Girieud, Richard; Soulard, Maxime

2018-05-01

CESAR stands in French for "simplified depletion applied to reprocessing". The current version is now number 5.3 as it started 30 years ago from a long lasting cooperation with ORANO, co-owner of the code with CEA. This computer code can characterize several types of nuclear fuel assemblies, from the most regular PWR power plants to the most unexpected gas cooled and graphite moderated old timer research facility. Each type of fuel can also include numerous ranges of compositions like UOX, MOX, LEU or HEU. Such versatility comes from a broad catalog of cross section libraries, each corresponding to a specific reactor and fuel matrix design. CESAR goes beyond fuel characterization and can also provide an evaluation of structural materials activation. The cross-sections libraries are generated using the most refined assembly or core level transport code calculation schemes (CEA APOLLO2 or ERANOS), based on the European JEFF3.1.1 nuclear data base. Each new CESAR self shielded cross section library benefits all most recent CEA recommendations as for deterministic physics options. Resulting cross sections are organized as a function of burn up and initial fuel enrichment which allows to condensate this costly process into a series of Legendre polynomials. The final outcome is a fast, accurate and compact CESAR cross section library. Each library is fully validated, against a stochastic transport code (CEA TRIPOLI 4) if needed and against a reference depletion code (CEA DARWIN). Using CESAR does not require any of the neutron physics expertise implemented into cross section libraries generation. It is based on top quality nuclear data (JEFF3.1.1 for ˜400 isotopes) and includes up to date Bateman equation solving algorithms. However, defining a CESAR computation case can be very straightforward. Most results are only 3 steps away from any beginner's ambition: Initial composition, in core depletion and pool decay scenario. On top of a simple utilization architecture

Infinite fuel element simulation of pin power distributions and control blade history in a BWR fuel assembly

Energy Technology Data Exchange (ETDEWEB)

Li, J.; Nuenighoff, K.; Allelein, H.J. [Forschungszentrum Juelich GmbH (DE). Inst. fuer Energie- und Klimaforschung (IEK), Sicherheitsforschung und Reaktortechnik (IEK-6)

2011-07-01

Pellet-Cladding Interaction (PCI) is a well known effect in fuel pins. One possible reason for PCI-effects could be local power excursions in the fuel pins, which can led to a rupture of the fuel cladding tube. From a reactor safety point of view this has to be considered as a violence of the barrier principal in order to retain fission products in the fuel pins. This paper focuses on the pin power distributions in a 2D infinite lattice of a BWR fuel element. Lots of studies related PCI effect can be found in the literature. In this compact, coupled neutronic depletion calculations taking the control history effect into account are described. Depletion calculations of an infinite fuel element of a BWR were carried out with controlled, uncontrolled and temporarily controlled scenarios. Later ones are needed to describe the control blade history (CBH) effect. A Monte-Carlo approach is mandatory to simulate the neutron physics. The VESTA code was applied to couple the Monte-Carlo-Code MCNP(X) with the burnup code ORIGEN. Additionally, CASMO-4 is also employed to verify the method of simulation results from VESTA. The cross sections for Monte Carlo and burn-up calculations are derived from ENDF/B-VII.0. (orig.)

Depleted Reactor Analysis With MCNP-4B

International Nuclear Information System (INIS)

Caner, M.; Silverman, L.; Bettan, M.

2004-01-01

Monte Carlo neutronics calculations are mostly done for fresh reactor cores. There is today an ongoing activity in the development of Monte Carlo plus burnup code systems made possible by the fast gains in computer processor speeds. In this work we investigate the use of MCNP-4B for the calculation of a depleted core of the Soreq reactor (IRR-1). The number densities as function of burnup were taken from the WIMS-D/4 cell code calculations. This particular code coupling has been implemented before. The Monte Carlo code MCNP-4B calculates the coupled transport of neutrons and photons for complicated geometries. We have done neutronics calculations of the IRR-1 core with the WIMS and CITATION codes in the past Also, we have developed an MCNP model of the IRR-1 standard fuel for a criticality safety calculation of a spent fuel storage pool

Fuel conditioning facility electrorefiner start-up results

International Nuclear Information System (INIS)

Goff, K.M.; Mariani, R.D.; Vaden, D.; Bonomo, N.L.; Cunningham, S.S.

1996-01-01

At ANL-West, there are several thousand kilograms of metallic spent nuclear fuel containing bond sodium. This fuel will be treated in the Fuel Conditioning Facility (FCF) at ANL-West to produce stable waste forms for storage and disposal. The treatment operations will make use of an electrometallurgical process employing molten salts and liquid metals. The treatment equipment is presently undergoing testing with depleted uranium. Operations with irradiated fuel will commence when the environmental evaluation for FCF is complete

"When the going gets tough, who keeps going?" Depletion sensitivity moderates the ego-depletion effect.

Science.gov (United States)

Salmon, Stefanie J; Adriaanse, Marieke A; De Vet, Emely; Fennis, Bob M; De Ridder, Denise T D

2014-01-01

Self-control relies on a limited resource that can get depleted, a phenomenon that has been labeled ego-depletion. We argue that individuals may differ in their sensitivity to depleting tasks, and that consequently some people deplete their self-control resource at a faster rate than others. In three studies, we assessed individual differences in depletion sensitivity, and demonstrate that depletion sensitivity moderates ego-depletion effects. The Depletion Sensitivity Scale (DSS) was employed to assess depletion sensitivity. Study 1 employs the DSS to demonstrate that individual differences in sensitivity to ego-depletion exist. Study 2 shows moderate correlations of depletion sensitivity with related self-control concepts, indicating that these scales measure conceptually distinct constructs. Study 3 demonstrates that depletion sensitivity moderates the ego-depletion effect. Specifically, participants who are sensitive to depletion performed worse on a second self-control task, indicating a stronger ego-depletion effect, compared to participants less sensitive to depletion.

When the Going Gets Tough, Who Keeps Going? Depletion Sensitivity Moderates the Ego-Depletion Effect

Directory of Open Access Journals (Sweden)

Stefanie J. Salmon

2014-06-01

Full Text Available Self-control relies on a limited resource that can get depleted, a phenomenon that has been labeled ego-depletion. We argue that individuals may differ in their sensitivity to depleting tasks, and that consequently some people deplete their self-control resource at a faster rate than others. In three studies, we assessed individual differences in depletion sensitivity, and demonstrate that depletion sensitivity moderates ego-depletion effects. The Depletion Sensitivity Scale (DSS was employed to assess depletion sensitivity. Study 1 employs the DSS to demonstrate that individual differences in sensitivity to ego-depletion exist. Study 2 shows moderate correlations of depletion sensitivity with related self-control concepts, indicating that these scales measure conceptually distinct constructs. Study 3 demonstrates that depletion sensitivity moderates the ego-depletion effect. Specifically, participants who are sensitive to depletion performed worse on a second self-control task, indicating a stronger ego-depletion effect, compared to participants less sensitive to depletion.

Extension of hybrid micro-depletion model for decay heat calculation in the DYN3D code

International Nuclear Information System (INIS)

Bilodid, Yurii; Fridman, Emil; Shwageraus, E.

2017-01-01

This work extends the hybrid micro-depletion methodology, recently implemented in DYN3D, to the decay heat calculation by accounting explicitly for the heat contribution from the decay of each nuclide in the fuel.

Extension of hybrid micro-depletion model for decay heat calculation in the DYN3D code

Energy Technology Data Exchange (ETDEWEB)

Bilodid, Yurii; Fridman, Emil [Helmholtz-Zentrum Dresden-Rossendorf e.V., Dresden (Germany). Reactor Safety; Kotlyar, D. [Georgia Institute of Technology, Atlanta, GA (United States); Shwageraus, E. [Cambridge Univ. (United Kingdom)

2017-06-01

This work extends the hybrid micro-depletion methodology, recently implemented in DYN3D, to the decay heat calculation by accounting explicitly for the heat contribution from the decay of each nuclide in the fuel.

Fuel reprocessing data validation using the isotope correlation technique

International Nuclear Information System (INIS)

Persiani, P.J.; Bucher, R.G.; Pond, R.B.; Cornella, R.J.

1990-01-01

The Isotope Correlation Technique (ICT), in conjunction with the gravimetric (Pu/U ratio) method for mass determination, provides an independent verification of the input accountancy at the dissolver or accountancy stage of the reprocessing plant. The Isotope Correlation Technique has been applied to many classes of domestic and international reactor systems (light-water, heavy-water, and graphite reactors) operating in a variety of modes (power, research, and production reactors), and for a variety of reprocessing fuel cycle management strategies. Analysis of reprocessing operations data based on isotopic correlations derived for assemblies in a PWR environment and fuel management scheme, yielded differences between the measurement-derived and ICT-derived plutonium mass determinations of (- 0.02 ± 0.23)% for the measured U-235 and (+ 0.50 ± 0.31)% for the measured Pu-239, for a core campaign. The ICT analyses has been implemented for the plutonium isotopics in a depleted uranium assembly in a heavy-water, enriched uranium system and for the uranium isotopes in the fuel assemblies in light-water, highly-enriched systems

Analyses for licensing of new fuel types at Paks NPP

International Nuclear Information System (INIS)

Kereszturi, A.; Bogatyr, S.; Miko, S.; Nemes, I.

2003-01-01

In the last years Paks NPP initiated several projects aiming at the introduction of new fuel types and resulting in more economic fuel cycles. The motivations, the reasons, and the economic consequences of the above modifications are detailed. The application of a new fuel type requires the renewal of the relevant chapters of the Safety Analysis Report. The fulfilment of fuel design basis requirements, to be summarised briefly also in the paper, must be investigated during normal and accidental conditions. The characteristics of the different codes, the data transfer between them are detailed. After, the cases of the Normal Operation, Anticipated Operation Occurrence, and the Postulated Accidents, judged as the most relevant ones in case of fuel modifications, are overviewed. In the last part, selected examples of the licensing calculations, performed by the above tools are presented. In conclusion, modifications of the WWER fuel, namely increased enrichment, application of burnable fuel pins, modified geometry make more economic fuel cycles (larger discharge burnup, power up-rate, reduced pressure vessel fluence) are possible. The further step (increased enrichment, burnable poison) of the fuel modernisation at NPP Paks is necessary for more economic fuel cycles and fuel consuming. A sound basis of licensing methodology, safety analysis, and necessary computer codes for the WWER fuel modernisation is available

Depleted uranium disposal options evaluation

International Nuclear Information System (INIS)

Hertzler, T.J.; Nishimoto, D.D.; Otis, M.D.

1994-05-01

The Department of Energy (DOE), Office of Environmental Restoration and Waste Management, has chartered a study to evaluate alternative management strategies for depleted uranium (DU) currently stored throughout the DOE complex. Historically, DU has been maintained as a strategic resource because of uses for DU metal and potential uses for further enrichment or for uranium oxide as breeder reactor blanket fuel. This study has focused on evaluating the disposal options for DU if it were considered a waste. This report is in no way declaring these DU reserves a ''waste,'' but is intended to provide baseline data for comparison with other management options for use of DU. To PICS considered in this report include: Retrievable disposal; permanent disposal; health hazards; radiation toxicity and chemical toxicity

Review of the SIMMER-II analyses of liquid-metal-cooled fast breeder reactor core-disruptive accident fuel escape

International Nuclear Information System (INIS)

DeVault, G.P.; Bell, C.R.

1985-01-01

Early fuel removal from the active core of a liquid-metal-cooled fast breeder reactor undergoing a core-disruptive accident may reduce the potential for large energetics resulting from recriticalities. This paper presents a review of analyses with the SIMMER-II computer program of the effectiveness of possible fuel escape paths. Where possible, how SIMMER-II compares with or is validated against experiments that simulated the escape paths also is discussed

IFBA credit in the Shearon Harris fuel racks with Vantage 5 fuel

International Nuclear Information System (INIS)

Boyd, W.A.; Schmidt, R.F.; Erwin, R.D.

1989-01-01

At the Shearon Harris nuclear plant, fuel management strategies are being considered which will result in feed fuel enrichments approaching 5.0 w/o U-235. These types of enrichments require a new criticality analysis to raise the existing fuel rack enrichment limit. It is receiving Westinghouse Vantage 5 fuel with integral fuel burnable absorber (IFBA) rods providing the depletable neutron absorber. An analysis was performed on the fuel racks which demonstrates that fuel enriched up to 5.0 w/o U-235 can be stored by taking credit for the IFBA rods present in the high enriched fuel assemblies. This is done by calculating the maximum Vantage 5 fuel assembly reactivity that can be placed in the fuel racks and meet the criticality K-eff limit. A methodology is also developed which conservatively calculates the minimum number of IFBA rods needed per assembly to meet the fuel rack storage limits. This eliminates the need for core designers to determine assembly K-inf terms for every different enrichment/IFBA combination

Analyses on Cost Reduction and CO2 Mitigation by Penetration of Fuel Cells to Residential Houses

Science.gov (United States)

Aki, Hirohisa; Yamamoto, Shigeo; Kondoh, Junji; Murata, Akinobu; Ishii, Itaru; Maeda, Tetsuhiko

This paper presents analyses on the penetration of polymer electrolyte fuel cells (PEFC) into a group of 10 residential houses and its effects of CO2 emission mitigation and consumers’ cost reduction in next 30 years. The price is considered to be reduced as the penetration progress which is expected to begin in near future. An experimental curve is assumed to express the decrease of the price. Installation of energy interchange systems which involve electricity, gas and hydrogen between a house which has a FC and contiguous houses is assumed to utilize both electricity and heat more efficiently, and to avoid start-stop operation of fuel processor (reformer) as much as possible. A multi-objective model which considers CO2 mitigation and consumers’ cost reduction is constructed and provided a Pareto optimum solution. A solution which simultaneously realizes both CO2 mitigation and consumers’ cost reduction appeared in the Pareto optimum solution. Strategies to reduce CO2 emission and consumers’ cost are suggested from the results of the analyses. The analyses also revealed that the energy interchange systems are effective especially in the early stage of the penetration.

Dose rate analyses for fast reactor fuel manufacturers

International Nuclear Information System (INIS)

Smith, R.C.; Strode, J.N.; Brackenbush, L.W.; Faust, L.G.

1976-01-01

An early appraisal of the radiation exposure situation in the fabrication of plutonium enriched mixed-oxide fuels for fast reactors is presented. Radiation data are presented on fuel process operations measured under actual operating conditions using plutonium containing up to 19 wt percent 240 Pu

The optimum fuel and power distribution for a PWR burnup cycle

International Nuclear Information System (INIS)

Stillman, J.A.

1989-01-01

A method was developed to determine the optimum fuel and power distributions for a PWR burnup cycle. The backward diffusion calculation [1] and the Core-wise Green's Function [2] method were used for the core model which provided analytic derivatives for solving the nonlinear optimization problem using successive linear programming [3] methods. The solution algorithm consisted of a reverse depletion strategy which begins at the end of cycle and solves simultaneously for the optimal fuel and burnable absorber distributions while the core is depleted to the beginning of cycle. The resulting optimal solutions minimize the required fissile fuel inventory and burnable absorber loading for a PWR

Bio-fuels for the gas turbine: A review

International Nuclear Information System (INIS)

Gupta, K.K.; Rehman, A.; Sarviya, R.M.

2010-01-01

Due to depletion of fossil fuel, bio-fuels have generated a significant interest as an alternative fuel for the future. The use of bio-fuels to fuel gas turbine seems a viable solution for the problems of decreasing fossil-fuel reserves and environmental concerns. Bio-fuels are alternative fuels, made from renewable sources and having environmental benefit. In recent years, the desire for energy independence, foreseen depletion of nonrenewable fuel resources, fluctuating petroleum fuel costs, the necessity of stimulating agriculture based economy, and the reality of climate change have created an interest in the development of bio-fuels. The application of bio-fuels in automobiles and heating applications is increasing day by day. Therefore the use of these fuels in gas turbines would extend this application to aviation field. The impact of costly petroleum-based aviation fuel on the environment is harmful. So the development of alternative fuels in aviation is important and useful. The use of liquid and gaseous fuels from biomass will help to fulfill the Kyoto targets concerning global warming emissions. In addition, to reduce exhaust emission waste gases and syngas, etc., could be used as a potential gas turbine fuel. The term bio-fuel is referred to alternative fuel which is produced from biomass. Such fuels include bio-diesel, bio-ethanol, bio-methanol, pyrolysis oil, biogas, synthetic gas (dimethyl ether), hydrogen, etc. The bio-ethanol and bio-methanol are petrol additive/substitute. Bio-diesel is an environment friendly alternative liquid fuel for the diesel/aviation fuel. The gas turbine develops steady flame during its combustion; this feature gives a flexibility to use alternative fuels. Therefore so the use of different bio-fuels in gas turbine has been investigated by a good number of researchers. The suitability and modifications in the existing systems are also recommended. (author)

A Preliminary Assessment of the Adjuster Rod Depletion Effect in the CANDU Reactor

International Nuclear Information System (INIS)

Kim, Yonghee; Roh, Gyuhong; Kim, Won Young; Kim, Hak Sung; Park, Joo Hwan

2008-01-01

Lifetime of the Wolsong-1 CANDU reactor, which will be shutdown in April, 2009. Major reactor components such as the pressure tube are to be replaced and it is expected that the CANDU reactor can be operated for additional 25-30 years. Meanwhile, all the reactivity devices including the adjuster rods (ADJ) are supposed to be continuously used without any change. In the CANDU reactor, 21 stainless steel (SS) ADJs are used to control the core power distribution and compensate for some reactivity loss during several abnormal cases. The ADJs are normally fully inserted and the SS absorber should undergo a slow depletion through neutron irradiation for a long time. In April, 2009, the accumulated FPY (Full Power Day) of Wolsong-1 is about 23 years. Depletion of ADJs should result in a smaller ADJ worth and a higher fuel burnup and the core power distribution should also be affected by the ADJ depletion. In this work, the effects of the ADJ depletion have been assessed in terms of ADJ worth, time-average core characteristics

Description of Transmutation Library for Fuel Cycle System Analyses

International Nuclear Information System (INIS)

Piet, Steven J.; Bays, Samuel E.; Hoffman, Edward A.

2010-01-01

This report documents the Transmutation Library that is used in Fuel Cycle System Analyses. This version replaces the 2008 version.(Piet2008) The Transmutation Library has the following objectives: (1) Assemble past and future transmutation cases for system analyses. (2) For each case, assemble descriptive information such as where the case was documented, the purpose of the calculation, the codes used, source of feed material, transmutation parameters, and the name of files that contain raw or source data. (3) Group chemical elements so that masses in separation and waste processes as calculated in dynamic simulations or spreadsheets reflect current thinking of those processes. For example, the CsSr waste form option actually includes all Group 1A and 2A elements. (4) Provide mass fractions at input (charge) and output (discharge) for each case. (5) Eliminate the need for either ''fission product other'' or ''actinide other'' while conserving mass. Assessments of waste and separation cannot use ''fission product other'' or ''actinide other'' as their chemical behavior is undefined. (6) Catalog other isotope-specific information in one place, e.g., heat and dose conversion factors for individual isotopes. (7) Describe the correlations for how input and output compositions change as a function of UOX burnup (for LWR UOX fuel) or fast reactor (FR) transuranic (TRU) conversion ratio (CR) for either FR-metal or FR-oxide. This document therefore includes the following sections: (1) Explanation of the data set information, i.e., the data that describes each case. In no case are all of the data presented in the Library included in previous documents. In assembling the Library, we return to raw data files to extract the case and isotopic data, into the specified format. (2) Explanation of which isotopes and elements are tracked. For example, the transition metals are tracked via the following: two Zr isotopes, Zr-other, Tc99, Tc-other, two Mo-Ru-Rh-Pd isotopes, Mo

Optimal initial fuel distribution in a thermal reactor for maximum energy production

International Nuclear Information System (INIS)

Moran-Lopez, J.M.

1983-01-01

Using the fuel burnup as objective function, it is desired to determine the initial distribution of the fuel in a reactor in order to obtain the maximum energy possible, for which, without changing a fixed initial fuel mass, the results for different initial fuel and control poison configurations are analyzed and the corresponding running times compared. One-dimensional, two energy-group theory is applied to a reflected cylindrical reactor using U-235 as fuel and light water as moderator and reflector. Fissions in both fast and thermal groups are considered. The reactor is divided into several annular regions, and the constant flux approximation in each depletion step is then used to solve the fuel and fission-product poisons differential equations in each region. The computer code OPTIME was developed to determine the time variation of core properties during the fuel cycle. At each depletion step, OPTIME calls ODMUG, [12] a criticality search program, from which the spatially-averaged neutron fluxes and control poison cross sections are obtained

Monte Carlo Depletion with Critical Spectrum for Assembly Group Constant Generation

International Nuclear Information System (INIS)

Park, Ho Jin; Joo, Han Gyu; Shim, Hyung Jin; Kim, Chang Hyo

2010-01-01

The conventional two-step procedure has been used in practical nuclear reactor analysis. In this procedure, a deterministic assembly transport code such as HELIOS and CASMO is normally to generate multigroup flux distribution to be used in few-group cross section generation. Recently there are accuracy issues related with the resonance treatment or the double heterogeneity (DH) treatment for VHTR fuel blocks. In order to mitigate the accuracy issues, Monte Carlo (MC) methods can be used as an alternative way to generate few-group cross sections because the accuracy of the MC calculations benefits from its ability to use continuous energy nuclear data and detailed geometric information. In an earlier work, the conventional methods of obtaining multigroup cross sections and the critical spectrum are implemented into the McCARD Monte Carlo code. However, it was not complete in that the critical spectrum is not reflected in the depletion calculation. The purpose of this study is to develop a method to apply the critical spectrum to MC depletion calculations to correct for the leakage effect in the depletion calculation and then to examine the MC based group constants within the two-step procedure by comparing the two-step solution with the direct whole core MC depletion result

Physicochemical Characterization of Capstone Depleted Uranium Aerosols III: Morphologic and Chemical Oxide Analyses

International Nuclear Information System (INIS)

Krupka, Kenneth M.; Parkhurst, MaryAnn; Gold, Kenneth; Arey, Bruce W.; Jenson, Evan D.; Guilmette, Raymond A.

2009-01-01

The impact of depleted uranium (DU) penetrators against an armored target causes erosion and fragmentation of the penetrators, the extent of which is dependent on the thickness and material composition of the target. Vigorous oxidation of the DU particles and fragments creates an aerosol of DU oxide particles and DU particle agglomerations combined with target materials. Aerosols from the Capstone DU aerosol study, in which vehicles were perforated by DU penetrators, were evaluated for their oxidation states using X-ray diffraction (XRD) and particle morphologies using scanning electron microscopy/energy dispersive spectrometry (SEM/EDS). The oxidation state of a DU aerosol is important as it offers a clue to its solubility in lung fluids. The XRD analysis showed that the aerosols evaluated were a combination primarily of U3O8 (insoluble) and UO3 (relatively more soluble) phases, though intermediate phases resembling U4O9 and other oxides were prominent in some samples. Analysis of particle residues in the micrometer-size range by SEM/EDS provided microstructural information such as phase composition and distribution, fracture morphology, size distribution, and material homogeneity. Observations from SEM analysis show a wide variability in the shapes of the DU particles. Some of the larger particles appear to have been fractured (perhaps as a result of abrasion and comminution); others were spherical, occasionally with dendritic or lobed surface structures. Amorphous conglomerates containing metals other than uranium were also common, especially with the smallest particle sizes. A few samples seemed to contain small chunks of nearly pure uranium metal, which were verified by EDS to have a higher uranium content exceeding that expected for uranium oxides. Results of the XRD and SEM/EDS analyses were used in other studies described in this issue of The Journal of Health Physics to interpret the results of lung solubility studies and in selecting input parameters for

Depletion Calculations for MTR Core Using MCNPX and Multi-Group Nodal Diffusion Methods

International Nuclear Information System (INIS)

Jaradata, Mustafa K.; Park, Chang Je; Lee, Byungchul

2013-01-01

In order to maintain a self-sustaining steady-state chain reaction, more fuel than is necessary in order to maintain a steady state chain reaction must be loaded. The introduction of this excess fuel increases the net multiplication capability of the system. In this paper MCNPX and multi-group nodal diffusion theory will be used for depletion calculations for MTR core. The eigenvalue and power distribution in the core will be compared for different burnup. Multi-group nodal diffusion theory with combination of NEWT-TRITON system was used to perform depletion calculations for 3Χ3 MTR core. 2G and 6G approximations were used and compared with MCNPX results for 2G approximation the maximum difference from MCNPX was 40 mk and for 6G approximation was 6 mk which is comparable to the MCNPX results. The calculated power using nodal code was almost the same MCNPX results. Finally the results of the multi-group nodal theory were acceptable and comparable to the calculated using MCNPX

Deuterium-depleted water

International Nuclear Information System (INIS)

Stefanescu, Ion; Steflea, Dumitru; Saros-Rogobete, Irina; Titescu, Gheorghe; Tamaian, Radu

2001-01-01

Deuterium-depleted water represents water that has an isotopic content smaller than 145 ppm D/(D+H) which is the natural isotopic content of water. Deuterium depleted water is produced by vacuum distillation in columns equipped with structured packing made from phosphor bronze or stainless steel. Deuterium-depleted water, the production technique and structured packing are patents of National Institute of Research - Development for Cryogenics and Isotopic Technologies at Rm. Valcea. Researches made in the last few years showed the deuterium-depleted water is a biological active product that could have many applications in medicine and agriculture. (authors)

A model for fuel shuffling and burnable absorbers optimization in low leakage PWRs

International Nuclear Information System (INIS)

Zavaljevski, N.

1990-01-01

A nonlinear model for the simultaneous optimization of fuel shuffling and burnable absorbers in PWRs is formulated using the depletion perturbation theory. The sensitivity coefficients are defined in a new way, using a macroscopic burnup model coupled with the explicit burnable absorbers depletion equation. Since first-order perturbation theory is limited to small changes in burnable absorber concentration, the associated control variable is continuous, with a constraint on maximal increment. Fuel shuffling is described by Boolean variables. Thus a special case of a mixed-integer quadratic programming problem is obtained, since the interaction of fuel and absorber optimization is considered. (author)

Pollution and exhaustibility of fossil fuels

NARCIS (Netherlands)

Withagen, C.A.A.M.

1994-01-01

The use of fossil fuels causes environmental damage. This is modeled and the ‘optimal’ rate of depletion is derived. Also this trajectory is compared with the case where there occurs no environmental damage.

Circular arc fuel plate stability experiments and analyses for the advanced neutron source

International Nuclear Information System (INIS)

Swinson, W.F.; Battiste, R.L.; Yahr, G.T.

1995-08-01

The thin fuel plates planned for the Advanced Neutron Source are to be cooled by forcing heavy water at high velocity, 25 m/s, through thin cooling channels on each side of each plate. Because the potential for structural failure of the plates is a design concern, considerable effort has been expended in assessing this potential. As part of this effort, experimental flow tests and analyses to evaluate the structural response of circular arc plates have been conducted, and the results are given in this report

SCALE Validation Experience Using an Expanded Isotopic Assay Database for Spent Nuclear Fuel

International Nuclear Information System (INIS)

Gauld, Ian C.; Radulescu, Georgeta; Ilas, Germina

2009-01-01

The availability of measured isotopic assay data to validate computer code predictions of spent fuel compositions applied in burnup-credit criticality calculations is an essential component for bias and uncertainty determination in safety and licensing analyses. In recent years, as many countries move closer to implementing or expanding the use of burnup credit in criticality safety for licensing, there has been growing interest in acquiring additional high-quality assay data. The well-known open sources of assay data are viewed as potentially limiting for validating depletion calculations for burnup credit due to the relatively small number of isotopes measured (primarily actinides with relatively few fission products), sometimes large measurement uncertainties, incomplete documentation, and the limited burnup and enrichment range of the fuel samples. Oak Ridge National Laboratory (ORNL) recently initiated an extensive isotopic validation study that includes most of the public data archived in the Organization for Economic Cooperation and Development/Nuclear Energy Agency (OECD/NEA) electronic database, SFCOMPO, and new datasets obtained through participation in commercial experimental programs. To date, ORNL has analyzed approximately 120 different spent fuel samples from pressurized-water reactors that span a wide enrichment and burnup range and represent a broad class of assembly designs. The validation studies, completed using SCALE 5.1, are being used to support a technical basis for expanded implementation of burnup credit for spent fuel storage facilities, and other spent fuel analyses including radiation source term, dose assessment, decay heat, and waste repository safety analyses. This paper summarizes the isotopic assay data selected for this study, presents validation results obtained with SCALE 5.1, and discusses some of the challenges and experience associated with evaluating the results. Preliminary results obtained using SCALE 6 and ENDF

Process for removal of sulfur compounds from fuel gases

Science.gov (United States)

Moore, Raymond H.; Stegen, Gary E.

1978-01-01

Fuel gases such as those produced in the gasification of coal are stripped of sulfur compounds and particulate matter by contact with molten metal salt. The fuel gas and salt are intimately mixed by passage through a venturi or other constriction in which the fuel gas entrains the molten salt as dispersed droplets to a gas-liquid separator. The separated molten salt is divided into a major and a minor flow portion with the minor flow portion passing on to a regenerator in which it is contacted with steam and carbon dioxide as strip gas to remove sulfur compounds. The strip gas is further processed to recover sulfur. The depleted, minor flow portion of salt is passed again into contact with the fuel gas for further sulfur removal from the gas. The sulfur depleted, fuel gas then flows through a solid absorbent for removal of salt droplets. The minor flow portion of the molten salt is then recombined with the major flow portion for feed to the venturi.

MCOR - Monte Carlo depletion code for reference LWR calculations

Energy Technology Data Exchange (ETDEWEB)

Puente Espel, Federico, E-mail: fup104@psu.edu [Department of Mechanical and Nuclear Engineering, Pennsylvania State University (United States); Tippayakul, Chanatip, E-mail: cut110@psu.edu [Department of Mechanical and Nuclear Engineering, Pennsylvania State University (United States); Ivanov, Kostadin, E-mail: kni1@psu.edu [Department of Mechanical and Nuclear Engineering, Pennsylvania State University (United States); Misu, Stefan, E-mail: Stefan.Misu@areva.com [AREVA, AREVA NP GmbH, Erlangen (Germany)

2011-04-15

Research highlights: > Introduction of a reference Monte Carlo based depletion code with extended capabilities. > Verification and validation results for MCOR. > Utilization of MCOR for benchmarking deterministic lattice physics (spectral) codes. - Abstract: The MCOR (MCnp-kORigen) code system is a Monte Carlo based depletion system for reference fuel assembly and core calculations. The MCOR code is designed as an interfacing code that provides depletion capability to the LANL Monte Carlo code by coupling two codes: MCNP5 with the AREVA NP depletion code, KORIGEN. The physical quality of both codes is unchanged. The MCOR code system has been maintained and continuously enhanced since it was initially developed and validated. The verification of the coupling was made by evaluating the MCOR code against similar sophisticated code systems like MONTEBURNS, OCTOPUS and TRIPOLI-PEPIN. After its validation, the MCOR code has been further improved with important features. The MCOR code presents several valuable capabilities such as: (a) a predictor-corrector depletion algorithm, (b) utilization of KORIGEN as the depletion module, (c) individual depletion calculation of each burnup zone (no burnup zone grouping is required, which is particularly important for the modeling of gadolinium rings), and (d) on-line burnup cross-section generation by the Monte Carlo calculation for 88 isotopes and usage of the KORIGEN libraries for PWR and BWR typical spectra for the remaining isotopes. Besides the just mentioned capabilities, the MCOR code newest enhancements focus on the possibility of executing the MCNP5 calculation in sequential or parallel mode, a user-friendly automatic re-start capability, a modification of the burnup step size evaluation, and a post-processor and test-matrix, just to name the most important. The article describes the capabilities of the MCOR code system; from its design and development to its latest improvements and further ameliorations. Additionally

MCOR - Monte Carlo depletion code for reference LWR calculations

International Nuclear Information System (INIS)

Puente Espel, Federico; Tippayakul, Chanatip; Ivanov, Kostadin; Misu, Stefan

2011-01-01

Research highlights: → Introduction of a reference Monte Carlo based depletion code with extended capabilities. → Verification and validation results for MCOR. → Utilization of MCOR for benchmarking deterministic lattice physics (spectral) codes. - Abstract: The MCOR (MCnp-kORigen) code system is a Monte Carlo based depletion system for reference fuel assembly and core calculations. The MCOR code is designed as an interfacing code that provides depletion capability to the LANL Monte Carlo code by coupling two codes: MCNP5 with the AREVA NP depletion code, KORIGEN. The physical quality of both codes is unchanged. The MCOR code system has been maintained and continuously enhanced since it was initially developed and validated. The verification of the coupling was made by evaluating the MCOR code against similar sophisticated code systems like MONTEBURNS, OCTOPUS and TRIPOLI-PEPIN. After its validation, the MCOR code has been further improved with important features. The MCOR code presents several valuable capabilities such as: (a) a predictor-corrector depletion algorithm, (b) utilization of KORIGEN as the depletion module, (c) individual depletion calculation of each burnup zone (no burnup zone grouping is required, which is particularly important for the modeling of gadolinium rings), and (d) on-line burnup cross-section generation by the Monte Carlo calculation for 88 isotopes and usage of the KORIGEN libraries for PWR and BWR typical spectra for the remaining isotopes. Besides the just mentioned capabilities, the MCOR code newest enhancements focus on the possibility of executing the MCNP5 calculation in sequential or parallel mode, a user-friendly automatic re-start capability, a modification of the burnup step size evaluation, and a post-processor and test-matrix, just to name the most important. The article describes the capabilities of the MCOR code system; from its design and development to its latest improvements and further ameliorations

The investigation of HTGR fuel regeneration process

Energy Technology Data Exchange (ETDEWEB)

Lazarev, L N; Bertina, L E; Popik, V P; Isakov, V P; Alkhimov, N B; Pokhitonov, Yu A

1985-07-01

The aim of this report is the investigation of HTGR fuel regeneration. The operation in the technologic scheme of uranium extraction from fuel depleted elements is separation of fuel from graphite. Available methods of graphite matrix destruction are: mechanical destruction, chemical destruction, and burning. Mechanical destruction is done in combination with leaching or chlorination. Methods of chemical destruction of graphite matrix are not sufficiently studied. Most of the investigations nowadays sre devoted to removal of graphite by burning.

The investigation of HTGR fuel regeneration process

International Nuclear Information System (INIS)

Lazarev, L.N.; Bertina, L.E.; Popik, V.P.; Isakov, V.P.; Alkhimov, N.B.; Pokhitonov, Yu.A.

1985-01-01

The aim of this report is the investigation of HTGR fuel regeneration. The operation in the technologic scheme of uranium extraction from fuel depleted elements is separation of fuel from graphite. Available methods of graphite matrix destruction are: mechanical destruction, chemical destruction, and burning. Mechanical destruction is done in combination with leaching or chlorination. Methods of chemical destruction of graphite matrix are not sufficiently studied. Most of the investigations nowadays sre devoted to removal of graphite by burning

Tuneable diode laser gas analyser for methane measurements on a large scale solid oxide fuel cell

Science.gov (United States)

Lengden, Michael; Cunningham, Robert; Johnstone, Walter

2011-10-01

A new in-line, real time gas analyser is described that uses tuneable diode laser spectroscopy (TDLS) for the measurement of methane in solid oxide fuel cells. The sensor has been tested on an operating solid oxide fuel cell (SOFC) in order to prove the fast response and accuracy of the technology as compared to a gas chromatograph. The advantages of using a TDLS system for process control in a large-scale, distributed power SOFC unit are described. In future work, the addition of new laser sources and wavelength modulation will allow the simultaneous measurement of methane, water vapour, carbon-dioxide and carbon-monoxide concentrations.

Ego depletion in visual perception: Ego-depleted viewers experience less ambiguous figure reversal.

Science.gov (United States)

Wimmer, Marina C; Stirk, Steven; Hancock, Peter J B

2017-10-01

This study examined the effects of ego depletion on ambiguous figure perception. Adults (N = 315) received an ego depletion task and were subsequently tested on their inhibitory control abilities that were indexed by the Stroop task (Experiment 1) and their ability to perceive both interpretations of ambiguous figures that was indexed by reversal (Experiment 2). Ego depletion had a very small effect on reducing inhibitory control (Cohen's d = .15) (Experiment 1). Ego-depleted participants had a tendency to take longer to respond in Stroop trials. In Experiment 2, ego depletion had small to medium effects on the experience of reversal. Ego-depleted viewers tended to take longer to reverse ambiguous figures (duration to first reversal) when naïve of the ambiguity and experienced less reversal both when naïve and informed of the ambiguity. Together, findings suggest that ego depletion has small effects on inhibitory control and small to medium effects on bottom-up and top-down perceptual processes. The depletion of cognitive resources can reduce our visual perceptual experience.

A method for predicting fuel maintenance in once-through MSRs

International Nuclear Information System (INIS)

Ridley, Gavin; Chvala, Ondrej

2017-01-01

Highlights: •Novel method to control LEU-fueled MSR reactivity and chemistry is developed. •The method relies on adjusting material flows only. •Ten year full power depletion of a FLiBe core demonstrates the method. •Material flows and reactor physics data are presented. •Graphite lifetime and trifluoride actinide limits are satisfied. -- Abstract: Liquid fuel molten salt reactors allow reactivity control by material addition. This paper presents a method to adjust material flows in a molten salt reactor to keep the core critical, and to maintain desired reduction-oxidation potential in the core salt melt. The method is aimed at low-enriched uranium fueled thermal systems. It is developed as a Python library and uses Serpent2 Monte-Carlo transport and depletion code. A toy 300 MW(th) reactor with a FLiBe carrier salt is employed to demonstrate the performance of the method over 10 full power years. Results of the calculation are presented, including material flows, conversion ratio, effective delayed neutron fraction, and expected limits on trifluoride concentrations and graphite lifetime are investigated. This method lays a foundation for future studies including fuel cycle performance of molten salt reactors and dynamic behavior of the core during depletion.

Core design and fuel rod analyses of a super fast reactor with high power density

International Nuclear Information System (INIS)

Ju, Haitao; Cao, Liangzhi; Lu, Haoliang; Oka, Yoshiaki; Ikejiri, Satoshi; Ishiwatari, Yuki

2009-01-01

A Super Fast Reactor is a pressure-vessel type, fast spectrum SuperCritical Water Reactor (SCWR) that is presently researched in a Japanese project. One of the most important advantages of the Super Fast Reactor is the higher power density compared to the thermal spectrum SCWR, which reduces the capital cost. A preliminary core has an average power density of 158.8W/cc. In this paper, the principle of improving the average power density is studied and the core design is improved. After the sensitivity analyses on the fuel rod configurations, the fuel assembly configurations and the core configurations, an improved core with an average power density of 294.8W/cc is designed by 3-D neutronic/thermal-hydraulic coupled calculations. This power density is competitive with that of typical Liquid Metal Fast Breeder Reactors (LMFBR). In order to ensure the fuel rod integrity of this core design, the fuel rod behaviors on the normal operating condition are analyzed using FEMAXI-6 code. The power histories of each fuel rod are taken from the neutronics calculation results in the core design. The cladding surface temperature histories are taken from the thermal-hydraulic calculation results in the core design. Four types of the limiting fuel rods, with the Maximum Cladding Surface Temperature (MCST), Maximum Power Peak(MPP), Maximum Discharge Burnup(MDB) and Different Coolant Flow Pattern (DCFP), are chosen to cover all the fuel rods in the core. The available design range of the fuel rod design parameters, such as initial gas plenum pressure, gas plenum position, gas plenum length, grain size and gap size, are found out in order to satisfy the following design criteria: (1) Maximum fuel centerline temperature should be less than 1900degC. (2) Maximum cladding stress in circumstance direction should be less than 100MPa. (3) Pressure difference on the cladding should be less than 1/3 of buckling collapse pressure. (4) Cumulative damage faction (CDF) of the cladding should be

PENBURN - A 3-D Zone-Based Depletion/Burnup Solver

International Nuclear Information System (INIS)

Manalo, Kevin; Plower, Thomas; Rowe, Mireille; Mock, Travis; Sjoden, Glenn E.

2008-01-01

PENBURN (Parallel Environment Burnup) is a general depletion/burnup solver which, when provided with zone-based reaction rates, computes time-dependent isotope concentrations for a set of actinides and fission products. Burnup analysis in PENBURN is performed with a direct Bateman-solver chain solution technique. Specifically, in tandem with PENBURN is the use of PENTRAN, a parallel multi-group anisotropic Sn code for 3-D Cartesian geometries. In PENBURN, the linear chain method is actively used to solve individual isotope chains which are then fully attributed by the burnup code to yield integrated isotope concentrations for each nuclide specified. Included with the discussion of code features, a single PWR fuel pin calculation with the burnup code is performed and detailed with a benchmark comparison to PIE (Post-Irradiation Examination) data within the SFCOMPO (Spent Fuel Composition / NEA) database, and also with burnup codes in SCALE5.1. Conclusions within the paper detail, in PENBURN, the accuracy of major actinides, flux profile behavior as a function of burnup, and criticality calculations for the PWR fuel pin model. (authors)

TVSA-T fuel assembly for 'Temelin' NPP. Main results of design and safety analyses. Trends of development

International Nuclear Information System (INIS)

Samojlov, O.B.; Kajdalov, V.B.; Falkov, A.A.; Bolnov, V.A.; Morozkin, O.N.; Molchanov, V.L.; Ugryumov, A.V.

2010-01-01

TVSA is a fuel assembly with rigid skeleton formed by 6 angle pieces and SG is successfully operated at 17 VVER-1000 power units of Kalinin NPP, as well as at Ukrainian and Bulgarian NPPs. Based on a contract for fuel supply to the Temelin NPP, the TVSA-T fuel assembly was developed, building on proven solutions confirmed by operation of TVSA modifications during 4-6 years and by the results of post-irradiation examination. The TVSA-T design includes combined spacer grids (SG+MG) and by fuel column elongation by 150 mm. A set of analyses and experiments was performed to validate the design, including thermal hydraulic tests, validation of critical heat flux correlation for TVSA-T, integrated mechanical, vibration and lifetime tests. A licence to use the fuel has been granted by the Czech State Office for Nuclear Safety. The TVSA-T core is currently in operation at the Temelin-1 reactor unit. The presentation is concluded as follows: TVSA-T fuel assembly for Temelin has been validated. The TVSA-T design is based on approved technical decisions and meets the current requirements for lifetime, operational maneuverability and safety. The results of post-irradiation examination of TVSA-T operated at the Kalinin-1 unit for 4 years confirm the assembly operability, skeleton stiffness, geometric stability and normal fuel rod cladding condition. The properties of the TVSA fuel with MG allow the core power to be increased up to 3300 MW to match the envisaged future VVER (MIR-1200) design, providing allowable fuel rod power FΔh =1.63 (to implement effective fuel cycles). (P.A.)

Comparative evaluation of fuel temperature coefficient of standard and CANFLEX fuels in CANDU 6

International Nuclear Information System (INIS)

Kim, Woosong; Hartant, Donny; Kim, Yonghee

2012-01-01

The fuel temperature reactivity coefficient (FTC) of CANDU 6 has become a concerning issue. The FTC was found to be slightly positive for the operating condition of CANDU 6. Since CANDU 6 has unique fuel arrangement and very soft neutron spectrum, its Doppler reactivity feedback of U 238 is rather weak. The upscattering by oxygen in fuel and Pu 239 buildup with fuel depletion are responsible for the positive FTC value at high temperature. In this study, FTC of both standard CANDU and CANFLEX fuel lattice are re evaluated. A Monte Carlo code Serpent2 was chosen as the analysis tool because of its high calculational speed and it can account for the thermal motion of heavy nuclides in fuel by using the Doppler Broadening Rejection Correction (DBRC) method. It was reported that the fuel Doppler effect is noticeably enhanced by accounting the target thermal motion. Recently, it was found that the FTC of the CANDU 6 standard fuel is noticeably enhanced by the DBRC

Fuel tourism - a Scenario analysis; Tanktourismus - eine Szenario-Analyse

Energy Technology Data Exchange (ETDEWEB)

Michaelis, P. [Augsburg Univ. (Germany). Wirtschafts- und Sozialwissenschaftliche Fakultaet

2004-07-01

The present paper analyzes the incentives of domestic car drivers to get their fuel beyond the border line due to given price differences ('fuel tourism'). The paper distinguishes the cases of limited and complete rationality. Limited rationality means that the decision of car drivers is solely based on additional fuel costs and time effort; complete rationality, in contrast, means that all private costs are taken into account. The outcome shows that, regarding the case of limited rationality, even comparably small price differences induce a strong incentive for 'fuel tourism'. The key to a solution for this problem is to make car drivers more aware of the complete private costs of driving which they are already paying for today. (orig.)

Measurement of thermal diffusivity of depleted uranium metal microspheres

Science.gov (United States)

Humrickhouse-Helmreich, Carissa J.; Corbin, Rob; McDeavitt, Sean M.

2014-03-01

The high void space of nuclear fuels composed of homogeneous uranium metal microspheres may allow them to achieve ultra-high burnup by accommodating fuel swelling and reducing fuel/cladding interactions; however, the relatively low thermal conductivity of microsphere nuclear fuels may limit their application. To support the development of microsphere nuclear fuels, an apparatus was designed in a glovebox and used to measure the apparent thermal diffusivity of a packed bed of depleted uranium (DU) microspheres with argon fill in the void spaces. The developed Crucible Heater Test Assembly (CHTA) recorded radial temperature changes due to an initial heat pulse from a central thin-diameter cartridge heater. Using thermocouple positions and time-temperature data, the apparent thermal diffusivity was calculated. The thermal conductivity of the DU microspheres was calculated based on the thermal diffusivity from the CHTA, known material densities and specific heat capacities, and an assumed 70% packing density based on prior measurements. Results indicate that DU metal microspheres have very low thermal conductivity, relative to solid uranium metal, and rapidly form an oxidation layer even in a low oxygen environment. At 500 °C, the thermal conductivity of the DU metal microsphere bed was 0.431 ± 0.0560 W/m-K compared to the literature value of approximately 32 W/m-K for solid uranium metal.

Measurement of thermal diffusivity of depleted uranium metal microspheres

Energy Technology Data Exchange (ETDEWEB)

Humrickhouse-Helmreich, Carissa J., E-mail: carissahelmreich@tamu.edu [Texas A and M University, Department of Nuclear Engineering, 337 Zachry Engineering Center, 3133 TAMU, College Station, TX 77843 (United States); Corbin, Rob, E-mail: rcorbin@terrapower.com [TerraPower, LLC, 330 120th Ave NE, Suite 100, Bellevue, WA 98005 (United States); McDeavitt, Sean M., E-mail: mcdeavitt@tamu.edu [Texas A and M University, Department of Nuclear Engineering, 337 Zachry Engineering Center, 3133 TAMU, College Station, TX 77843 (United States)

2014-03-15

The high void space of nuclear fuels composed of homogeneous uranium metal microspheres may allow them to achieve ultra-high burnup by accommodating fuel swelling and reducing fuel/cladding interactions; however, the relatively low thermal conductivity of microsphere nuclear fuels may limit their application. To support the development of microsphere nuclear fuels, an apparatus was designed in a glovebox and used to measure the apparent thermal diffusivity of a packed bed of depleted uranium (DU) microspheres with argon fill in the void spaces. The developed Crucible Heater Test Assembly (CHTA) recorded radial temperature changes due to an initial heat pulse from a central thin-diameter cartridge heater. Using thermocouple positions and time–temperature data, the apparent thermal diffusivity was calculated. The thermal conductivity of the DU microspheres was calculated based on the thermal diffusivity from the CHTA, known material densities and specific heat capacities, and an assumed 70% packing density based on prior measurements. Results indicate that DU metal microspheres have very low thermal conductivity, relative to solid uranium metal, and rapidly form an oxidation layer even in a low oxygen environment. At 500 °C, the thermal conductivity of the DU metal microsphere bed was 0.431 ± 0.0560 W/m-K compared to the literature value of approximately 32 W/m-K for solid uranium metal.

Measurement of thermal diffusivity of depleted uranium metal microspheres

International Nuclear Information System (INIS)

Humrickhouse-Helmreich, Carissa J.; Corbin, Rob; McDeavitt, Sean M.

2014-01-01

The high void space of nuclear fuels composed of homogeneous uranium metal microspheres may allow them to achieve ultra-high burnup by accommodating fuel swelling and reducing fuel/cladding interactions; however, the relatively low thermal conductivity of microsphere nuclear fuels may limit their application. To support the development of microsphere nuclear fuels, an apparatus was designed in a glovebox and used to measure the apparent thermal diffusivity of a packed bed of depleted uranium (DU) microspheres with argon fill in the void spaces. The developed Crucible Heater Test Assembly (CHTA) recorded radial temperature changes due to an initial heat pulse from a central thin-diameter cartridge heater. Using thermocouple positions and time–temperature data, the apparent thermal diffusivity was calculated. The thermal conductivity of the DU microspheres was calculated based on the thermal diffusivity from the CHTA, known material densities and specific heat capacities, and an assumed 70% packing density based on prior measurements. Results indicate that DU metal microspheres have very low thermal conductivity, relative to solid uranium metal, and rapidly form an oxidation layer even in a low oxygen environment. At 500 °C, the thermal conductivity of the DU metal microsphere bed was 0.431 ± 0.0560 W/m-K compared to the literature value of approximately 32 W/m-K for solid uranium metal

Thermal-hydraulic analyses for in-pile SCWR fuel qualification test loops and SCWR material loop

Energy Technology Data Exchange (ETDEWEB)

Vojacek, A.; Mazzini, G.; Zmitkova, J.; Ruzickova, M. [Research Centre Rez (Czech Republic)

2014-07-01

One of the R&D directions of Research Centre Rez is dedicated to the supercritical water-cooled reactor concept (SCWR). Among the developed experimental facilities and infrastructure in the framework of the SUSEN project (SUStainable ENergy) is construction and experimental operation of the supercritical water loop SCWL focusing on material tests. At the first phase, this SCWL loop is assembled and operated out-of-pile in the dedicated loop facilities hall. At this out-of-pile operation various operational conditions are tested and verified. After that, in the second phase, the SCWL loop will be situated in-pile, in the core of the research reactor LVR-15, operated at CVR. Furthermore, it is planned to carry out a test of a small scale fuel assembly within the SuperCritical Water Reactor Fuel Qualification Test (SCWR-FQT) loop, which is now being designed. This paper presents the results of the thermal-hydraulic analyses of SCWL loop out-of-pile operation using the RELAP5/MOD3.3. The thermal-hydraulic modeling and the performed analyses are focused on the SCWL loop model validation through a comparison of the calculation results with the experimental results obtained at various operation conditions. Further, the present paper focuses on the transient analyses for start-up and shut-down of the FQT loop, particularly to explore the ability of system codes ATHLET 3.0A to simulate the transient between subcritical conditions and supercritical conditions. (author)

The pseudo-harmonics method applied to depletion calculation

International Nuclear Information System (INIS)

Silva, F.C. da; Amaral, J.A.C.; Thome, Z.D.

1989-01-01

In this paper, a new method for performing depletion calculations, based on the use of the Pseudo-Harmonics perturbation method, was developed. The fuel burnup was considered as a global perturbation and the multigroup difusion equations were rewriten in such a way as to treat the soluble boron concentration as the eigenvalue. By doing this, the critical boron concentration can be obtained by a perturbation method. A test of the new method was performed for a H 2 O-colled, D 2 O-moderated reactor. Comparison with direct calculation showed that this method is very accurate and efficient. (author) [pt

“When the going gets tough, who keeps going?” Depletion sensitivity moderates the ego-depletion effect

Science.gov (United States)

Salmon, Stefanie J.; Adriaanse, Marieke A.; De Vet, Emely; Fennis, Bob M.; De Ridder, Denise T. D.

2014-01-01

Self-control relies on a limited resource that can get depleted, a phenomenon that has been labeled ego-depletion. We argue that individuals may differ in their sensitivity to depleting tasks, and that consequently some people deplete their self-control resource at a faster rate than others. In three studies, we assessed individual differences in depletion sensitivity, and demonstrate that depletion sensitivity moderates ego-depletion effects. The Depletion Sensitivity Scale (DSS) was employed to assess depletion sensitivity. Study 1 employs the DSS to demonstrate that individual differences in sensitivity to ego-depletion exist. Study 2 shows moderate correlations of depletion sensitivity with related self-control concepts, indicating that these scales measure conceptually distinct constructs. Study 3 demonstrates that depletion sensitivity moderates the ego-depletion effect. Specifically, participants who are sensitive to depletion performed worse on a second self-control task, indicating a stronger ego-depletion effect, compared to participants less sensitive to depletion. PMID:25009523

Full and semi-analytic analyses of two-pump parametric amplification with pump depletion

DEFF Research Database (Denmark)

Steffensen, Henrik; Ott, Johan Raunkjær; Rottwitt, Karsten

2011-01-01

This paper solves the four coupled equations describing non-degenerate four-wave mixing, with the focus on amplifying a signal in a fiber optical parametric amplifier (FOPA). Based on the full analytic solution, a simple approximate solution describing the gain is developed. The advantage...... of this new approximation is that it includes the depletion of the pumps, which is lacking in the usual quasi-linearized approximation. With the proposed model it is thus simple to predict the gain of a FOPA, which we demonstrate with a highly nonlinear fiber to show that an undepleted FOPA can produce a flat...

Uranium, depleted uranium, biological effects

International Nuclear Information System (INIS)

2001-01-01

Physicists, chemists and biologists at the CEA are developing scientific programs on the properties and uses of ionizing radiation. Since the CEA was created in 1945, a great deal of research has been carried out on the properties of natural, enriched and depleted uranium in cooperation with university laboratories and CNRS. There is a great deal of available data about uranium; thousands of analyses have been published in international reviews over more than 40 years. This presentation on uranium is a very brief summary of all these studies. (author)

Energizing and depletion of neutrals by a collisional plasma

International Nuclear Information System (INIS)

Fruchtman, A

2008-01-01

Neutral depletion can significantly affect the steady state of low temperature plasmas. Recent theoretical analyses predicted previously unexpected effects of neutral depletion in both collisional and collisionless regimes. In this paper we address the effect of the energy deposited in the neutral gas by a collisional plasma. The fraction of power deposited in the neutrals is shown to be independent of the amount of power. The first case we address is of a thermalized neutral gas. It is shown that a low heat conductivity of the neutral gas is followed by a high neutral temperature that results in a high neutral depletion even if the plasma pressure is small. In the second case neutrals are accelerated through charge exchange with ions leading to what we call neutral pumping, which is equivalent to ion pumping in a collisionless plasma. Neutral depletion is found in the second case for both a closed system (no net mass flow) and an open system (a finite mass flow). A thruster that employs a collisional plasma and pumped neutrals is compared with the thruster analyzed before that employs collisionless plasma.

Preliminary LEU fuel cycle analyses for the Belgian BR2 reactor

International Nuclear Information System (INIS)

Deen, J.R.; Snelgrove, J.L.

1986-01-01

Fuel cycle calculations have been performed with reference HEU fuel and LEU fuel using Cd wires or boron as burnable absorbers. The 235 U content in the LEU element has increased 20% to 480g compared to the reference HEU element. The number of fuel plates has remained unchanged while the fuel meat thickness has increased to 0.76 mm from 0.51 mm. The LEU meat density is 5.1 Mg U/m 3 . The reference fuel cycle was a 31 element core operating at 56 MW with a 19.8 day cycle length and eight fresh elements loaded per cycle. Comparable fuel cycle characteristics can be achieved using the proposed LEU fuel element with either Cd wires or boron burnable absorbers. The neutron flux for E/sub n/ > 1 eV changes very little (<5%) in LEU relative to HEU cores. Thermal flux reductions are 5 to 10% in non-fueled positions, and 20 to 30% in fuel elements

Standard guide for pyrophoricity/combustibility testing in support of pyrophoricity analyses of metallic uranium spent nuclear fuel

CERN Document Server

American Society for Testing and Materials. Philadelphia

2007-01-01

1.1 This guide covers testing protocols for testing the pyrophoricity/combustibility characteristics of metallic uranium-based spent nuclear fuel (SNF). The testing will provide basic data for input into more detailed computer codes or analyses of thermal, chemical, and mechanical SNF responses. These analyses would support the engineered barrier system (EBS) design bases and safety assessment of extended interim storage facilities and final disposal in a geologic repository. The testing also could provide data related to licensing requirements for the design and operation of a monitored retrievable storage facility (MRS) or independent spent fuel storage installation (ISFSI). 1.2 This guide describes testing of metallic uranium and metallic uranium-based SNF in support of transportation (in accordance with the requirements of 10CFR71), interim storage (in accordance with the requirements of 10CFR72), and geologic repository disposal (in accordance with the requirements of 10CFR60/63). The testing described ...

Fuel Handbook[Wood and other renewable fuels

Energy Technology Data Exchange (ETDEWEB)

Stroemberg, Birgitta [TPS Termiska Processer AB, Nykoeping (SE)] (ed.)

2006-03-15

This handbook on renewable fuels is intended for power and heat producers in Sweden. This fuel handbook provides, from a plant owner's perspective, a method to evaluate different fuels on the market. The fuel handbook concerns renewable fuels (but does not include household waste) that are available on the Swedish market today or fuels that have potential to be available within the next ten years. The handbook covers 26 different fuels. Analysis data, special properties, operating experiences and literature references are outlined for each fuel. [Special properties, operating experiences and literature references are not included in this English version] The handbook also contains: A proposed methodology for introduction of new fuels. A recommendation of analyses and tests to perform in order to reduce the risk of problems is presented. [The recommendation of analyses and tests is not included in the English version] A summary of relevant laws and taxes for energy production, with references to relevant documentation. [Only laws and taxes regarding EU are included] Theory and background to evaluate a fuel with respect to combustion, ash and corrosion properties and methods that can be used for such evaluations. Summary of standards, databases and handbooks on biomass fuels and other solid fuels, and links to web sites where further information about the fuels can be found. The appendices includes: A methodology for trial firing of fuels. Calculations procedures for, amongst others, heating value, flue gas composition, key number and free fall velocity [Free fall velocity is not included in the English version]. In addition, conversion routines between different units for a number of different applications are provided. Fuel analyses are presented in the appendix. (The report is a translation of parts of the report VARMEFORSK--911 published in 2005)

Applications of liquid phase chromatographies for the analysis of streams arising at the back end of nuclear fuel cycle

International Nuclear Information System (INIS)

Deshingkar, D.S.; Yalmali, Vrunda S.; Wattal, P.K.

2000-06-01

India has opted for a closed fuel cycle comprising of reprocessing and recycling technology. The back end of such nuclear fuel cycle involves the reprocessing of spent nuclear fuels for recovery of plutonium and depleted uranium by Purex technology. Wastes arising from the reprocessing plant are classified as high, intermediate and low level wastes (HLW, ILW, LLW). HLW is mixture of over 50 elements present in different chemical forms. The accurate analyses of dissolver solution and HLW are the most challenging but essential tasks for reprocessing plant operations and also for further development of treatment methods. Inductively coupled plasma - atomic emission spectroscopy and atomic absorption spectroscopy techniques are suitable for analysis of metallic anions. Ion chromatography has proven capability to analyse number of cations or anions at ppm or even ppb level in single run. The report reviews the literature regarding the title subject. To assess the technical feasibility of ion chromatography for waste analysis, a simulated PHWR-HLW analogue was prepared. The PHWR-HLW analogue and ground water samples were analysed on DIONEX-DX 500 and Metrohm IC. Results obtained clearly demonstrated the usefulness of ion chromatography as vital analytical tool. HLW and other process or waste streams arising at the back end of nuclear fuel cycle can be analysed for alkali, alkaline earth, rare earth and transition metal cations and important anions. Use of fraction collector along with ion chromatography can enhance it's sensitivity to few Bq/ml for radioactive samples. (author)

Multiple recycle of REMIX fuel at VVER-1000 operation in closed fuel cycle

Energy Technology Data Exchange (ETDEWEB)

Alekseev, P. N.; Bobrov, E. A., E-mail: evgeniybobrov89@rambler.ru; Chibinyaev, A. V.; Teplov, P. S.; Dudnikov, A. A. [National Research Center Kurchatov Institute (Russian Federation)

2015-12-15

The basic features of loading the VVER-1000 core with a new variant of REMIX fuel (REgenerated MIXture of U–Pu oxides) are considered during its multiple recycle in a closed nuclear fuel cycle. The fuel composition is produced on the basis of the uranium–plutonium regenerate extracted at processing the spent nuclear fuel (SNF) from a VVER-1000, depleted uranium, and the fissionable material: {sup 235}U as a part of highly enriched uranium (HEU) from warheads superfluous for defense purposes or {sup 233}U accumulated in thorium blankets of fusion (electronuclear) neutron sources or fast reactors. Production of such a fuel assumes no use of natural uranium in addition. When converting a part of the VVER-1000 reactors to the closed fuel cycle based on the REMIX technology, the consumption of natural uranium decreases considerably, and there is no substantial degradation of the isotopic composition of plutonium or change in the reactor-safety characteristics at the passage from recycle to recycle.

Vitamin C Depletion and All-Cause Mortality in Renal Transplant Recipients.

Science.gov (United States)

Sotomayor, Camilo G; Eisenga, Michele F; Gomes Neto, Antonio W; Ozyilmaz, Akin; Gans, Rijk O B; Jong, Wilhelmina H A de; Zelle, Dorien M; Berger, Stefan P; Gaillard, Carlo A J M; Navis, Gerjan J; Bakker, Stephan J L

2017-06-02

Vitamin C may reduce inflammation and is inversely associated with mortality in the general population. We investigated the association of plasma vitamin C with all-cause mortality in renal transplant recipients (RTR); and whether this association would be mediated by inflammatory biomarkers. Vitamin C, high sensitive C-reactive protein (hs-CRP), soluble intercellular cell adhesion molecule 1 (sICAM-1), and soluble vascular cell adhesion molecule 1 (sVCAM-1) were measured in a cohort of 598 RTR. Cox regression analyses were used to analyze the association between vitamin C depletion (≤28 µmol/L; 22% of RTR) and mortality. Mediation analyses were performed according to Preacher and Hayes's procedure. At a median follow-up of 7.0 (6.2-7.5) years, 131 (21%) patients died. Vitamin C depletion was univariately associated with almost two-fold higher risk of mortality (Hazard ratio (HR) 1.95; 95% confidence interval (95%CI) 1.35-2.81, p C depletion is frequent and independently associated with almost two-fold higher risk of mortality in RTR. It may be hypothesized that the beneficial effect of vitamin C at least partly occurs through decreasing inflammation.

Comparison of matrix exponential methods for fuel burnup calculations

International Nuclear Information System (INIS)

Oh, Hyung Suk; Yang, Won Sik

1999-01-01

Series expansion methods to compute the exponential of a matrix have been compared by applying them to fuel depletion calculations. Specifically, Taylor, Pade, Chebyshev, and rational Chebyshev approximations have been investigated by approximating the exponentials of bum matrices by truncated series of each method with the scaling and squaring algorithm. The accuracy and efficiency of these methods have been tested by performing various numerical tests using one thermal reactor and two fast reactor depletion problems. The results indicate that all the four series methods are accurate enough to be used for fuel depletion calculations although the rational Chebyshev approximation is relatively less accurate. They also show that the rational approximations are more efficient than the polynomial approximations. Considering the computational accuracy and efficiency, the Pade approximation appears to be better than the other methods. Its accuracy is better than the rational Chebyshev approximation, while being comparable to the polynomial approximations. On the other hand, its efficiency is better than the polynomial approximations and is similar to the rational Chebyshev approximation. In particular, for fast reactor depletion calculations, it is faster than the polynomial approximations by a factor of ∼ 1.7. (author). 11 refs., 4 figs., 2 tabs

Sensitivity analyses of seismic behavior of spent fuel dry cask storage systems

International Nuclear Information System (INIS)

Luk, V.K.; Spencer, B.W.; Shaukat, S.K.; Lam, I.P.; Dameron, R.A.

2003-01-01

Sandia National Laboratories is conducting a research project to develop a comprehensive methodology for evaluating the seismic behavior of spent fuel dry cask storage systems (DCSS) for the Office of Nuclear Regulatory Research of the U.S. Nuclear Regulatory Commission (NRC). A typical Independent Spent Fuel Storage Installation (ISFSI) consists of arrays of free-standing storage casks resting on concrete pads. In the safety review process of these cask systems, their seismically induced horizontal displacements and angular rotations must be quantified to determine whether casks will overturn or neighboring casks will collide during a seismic event. The ABAQUS/Explicit code is used to analyze three-dimensional coupled finite element models consisting of three submodels, which are a cylindrical cask or a rectangular module, a flexible concrete pad, and an underlying soil foundation. The coupled model includes two sets of contact surfaces between the submodels with prescribed coefficients of friction. The seismic event is described by one vertical and two horizontal components of statistically independent seismic acceleration time histories. A deconvolution procedure is used to adjust the amplitudes and frequency contents of these three-component reference surface motions before applying them simultaneously at the soil foundation base. The research project focused on examining the dynamic and nonlinear seismic behavior of the coupled model of free-standing DCSS including soil-structure interaction effects. This paper presents a subset of analysis results for a series of parametric analyses. Input variables in the parametric analyses include: designs of the cask/module, time histories of the seismic accelerations, coefficients of friction at the cask/pad interface, and material properties of the soil foundation. In subsequent research, the analysis results will be compiled and presented in nomograms to highlight the sensitivity of seismic response of DCSS to

Credible accident analyses for TRIGA and TRIGA-fueled reactors

International Nuclear Information System (INIS)

Hawley, S.C.; Kathren, R.L.

1982-04-01

Credible accidents were developed and analyzed for TRIGA and TRIGA-fueled reactors. The only potential for offsite exposure appears to be from a fuel-handling accident that, based on highly conservative assumptions, would result in dose equivalents of less than or equal to 1 mrem to the total body from noble gases and less than or equal to 1.2 rem to the thyroid from radioiodines. Credible accidents from excess reactivity insertions, metal-water reactions, lost, misplaced, or inadvertent experiments, core rearrangements, and changes in fuel morphology and ZrH/sub x/ composition are also evaluated, and suggestions for further study provided

Physicochemical characterization of Capstone depleted uranium aerosols III: morphologic and chemical oxide analyses.

Science.gov (United States)

Krupka, Kenneth M; Parkhurst, Mary Ann; Gold, Kenneth; Arey, Bruce W; Jenson, Evan D; Guilmette, Raymond A

2009-03-01

The impact of depleted uranium (DU) penetrators against an armored target causes erosion and fragmentation of the penetrators, the extent of which is dependent on the thickness and material composition of the target. Vigorous oxidation of the DU particles and fragments creates an aerosol of DU oxide particles and DU particle agglomerations combined with target materials. Aerosols from the Capstone DU aerosol study, in which vehicles were perforated by DU penetrators, were evaluated for their oxidation states using x-ray diffraction (XRD), and particle morphologies were examined using scanning electron microscopy/energy dispersive spectroscopy (SEM/EDS). The oxidation state of a DU aerosol is important as it offers a clue to its solubility in lung fluids. The XRD analysis showed that the aerosols evaluated were a combination primarily of U3O8 (insoluble) and UO3 (relatively more soluble) phases, though intermediate phases resembling U4O9 and other oxides were prominent in some samples. Analysis of particle residues in the micrometer-size range by SEM/EDS provided microstructural information such as phase composition and distribution, fracture morphology, size distribution, and material homogeneity. Observations from SEM analysis show a wide variability in the shapes of the DU particles. Some of the larger particles were spherical, occasionally with dendritic or lobed surface structures. Others appear to have fractures that perhaps resulted from abrasion and comminution, or shear bands that developed from plastic deformation of the DU material. Amorphous conglomerates containing metals other than uranium were also common, especially with the smallest particle sizes. A few samples seemed to contain small bits of nearly pure uranium metal, which were verified by EDS to have a higher uranium content exceeding that expected for uranium oxides. Results of the XRD and SEM/EDS analyses were used in other studies described in this issue of Health Physics to interpret the

LEU fuel cycle analyses for the Belgian BR2 Research Reactor

International Nuclear Information System (INIS)

Deen, J.R.; Snelgrove, J.L.

1988-01-01

Equilibrium fuel cycle characteristics were calculated for reference HEU and two proposed LEU fuel cycles using an 11-group diffusion-theory neutron flux solution in hexagonal-Z geometry. The diffusion theory model was benchmarked with a detailed Monte Carlo core model. The two proposed LEU fuel designs increased the 235 U loading 20% and the fuel meat volume 51%. The first LEU design used 10 B as a burnable absorber. Either proposed LEU fuel element would provide equilibrium fuel cycle characteristics similar to those of the HEU fuel cycle. Irradiation rates of Co control followers and Ir disks in the center of the core were reduced 6 ± 1% in the LEU equilibrium core compared to reference HEU core. 11 refs., 4 figs., 5 tabs

Advanced Fuel Cycle Cost Basis

Energy Technology Data Exchange (ETDEWEB)

D. E. Shropshire; K. A. Williams; W. B. Boore; J. D. Smith; B. W. Dixon; M. Dunzik-Gougar; R. D. Adams; D. Gombert; E. Schneider

2009-12-01

This report, commissioned by the U.S. Department of Energy (DOE), provides a comprehensive set of cost data supporting a cost analysis for the relative economic comparison of options for use in the Advanced Fuel Cycle Initiative (AFCI) Program. The report describes the AFCI cost basis development process, reference information on AFCI cost modules, a procedure for estimating fuel cycle costs, economic evaluation guidelines, and a discussion on the integration of cost data into economic computer models. This report contains reference cost data for 25 cost modules—23 fuel cycle cost modules and 2 reactor modules. The cost modules were developed in the areas of natural uranium mining and milling, conversion, enrichment, depleted uranium disposition, fuel fabrication, interim spent fuel storage, reprocessing, waste conditioning, spent nuclear fuel (SNF) packaging, long-term monitored retrievable storage, near surface disposal of low-level waste (LLW), geologic repository and other disposal concepts, and transportation processes for nuclear fuel, LLW, SNF, transuranic, and high-level waste.

Advanced Fuel Cycle Cost Basis

Energy Technology Data Exchange (ETDEWEB)

D. E. Shropshire; K. A. Williams; W. B. Boore; J. D. Smith; B. W. Dixon; M. Dunzik-Gougar; R. D. Adams; D. Gombert

2007-04-01

This report, commissioned by the U.S. Department of Energy (DOE), provides a comprehensive set of cost data supporting a cost analysis for the relative economic comparison of options for use in the Advanced Fuel Cycle Initiative (AFCI) Program. The report describes the AFCI cost basis development process, reference information on AFCI cost modules, a procedure for estimating fuel cycle costs, economic evaluation guidelines, and a discussion on the integration of cost data into economic computer models. This report contains reference cost data for 26 cost modules—24 fuel cycle cost modules and 2 reactor modules. The cost modules were developed in the areas of natural uranium mining and milling, conversion, enrichment, depleted uranium disposition, fuel fabrication, interim spent fuel storage, reprocessing, waste conditioning, spent nuclear fuel (SNF) packaging, long-term monitored retrievable storage, near surface disposal of low-level waste (LLW), geologic repository and other disposal concepts, and transportation processes for nuclear fuel, LLW, SNF, and high-level waste.

Advanced Fuel Cycle Cost Basis

Energy Technology Data Exchange (ETDEWEB)

D. E. Shropshire; K. A. Williams; W. B. Boore; J. D. Smith; B. W. Dixon; M. Dunzik-Gougar; R. D. Adams; D. Gombert; E. Schneider

2008-03-01

This report, commissioned by the U.S. Department of Energy (DOE), provides a comprehensive set of cost data supporting a cost analysis for the relative economic comparison of options for use in the Advanced Fuel Cycle Initiative (AFCI) Program. The report describes the AFCI cost basis development process, reference information on AFCI cost modules, a procedure for estimating fuel cycle costs, economic evaluation guidelines, and a discussion on the integration of cost data into economic computer models. This report contains reference cost data for 25 cost modules—23 fuel cycle cost modules and 2 reactor modules. The cost modules were developed in the areas of natural uranium mining and milling, conversion, enrichment, depleted uranium disposition, fuel fabrication, interim spent fuel storage, reprocessing, waste conditioning, spent nuclear fuel (SNF) packaging, long-term monitored retrievable storage, near surface disposal of low-level waste (LLW), geologic repository and other disposal concepts, and transportation processes for nuclear fuel, LLW, SNF, transuranic, and high-level waste.

Distribution of trichloroethylene and selected aliphatic and aromatic hydrocarbons between ''weathered'' and ''unweathered'' fuel mixtures and groundwater: Equilibrium and kinetic considerations

International Nuclear Information System (INIS)

Doucette, W.J.; Dupont, R.R.

1995-01-01

The distribution of trichloroethylene and several aliphatic and aromatic fuel components between 46 weathered and 11 unweathered fuel mixtures and groundwater was investigated using a slow stirring method. The weathered fuel mixtures were obtained from several contaminated field sites. Both unlabeled and 14C-labeled test compounds were used in the distribution experiments. Analyses of the test compound concentrations over time was performed by gas chromatograph or liquid scintillation counting. The time required to reach equilibrium varied from about 24 to 72 hours. Generally, the greater the hydrophobicity of the test compounds the longer time that was required to reach equilibrium. It was also observed that the fuel/water distribution coefficients were generally larger for the weathered fuels than those measured for the unweathered fuels, in some cases by a factor of 100. The weathered fuel mixtures obtained from the field site were depleted of the more water soluble compounds over time and became significantly more enriched in long chain aliphatic hydrocarbons. The ability of several models to describe the observed distribution behavior was examined

Halo Star Lithium Depletion

International Nuclear Information System (INIS)

Pinsonneault, M. H.; Walker, T. P.; Steigman, G.; Narayanan, Vijay K.

1999-01-01

The depletion of lithium during the pre-main-sequence and main-sequence phases of stellar evolution plays a crucial role in the comparison of the predictions of big bang nucleosynthesis with the abundances observed in halo stars. Previous work has indicated a wide range of possible depletion factors, ranging from minimal in standard (nonrotating) stellar models to as much as an order of magnitude in models that include rotational mixing. Recent progress in the study of the angular momentum evolution of low-mass stars permits the construction of theoretical models capable of reproducing the angular momentum evolution of low-mass open cluster stars. The distribution of initial angular momenta can be inferred from stellar rotation data in young open clusters. In this paper we report on the application of these models to the study of lithium depletion in main-sequence halo stars. A range of initial angular momenta produces a range of lithium depletion factors on the main sequence. Using the distribution of initial conditions inferred from young open clusters leads to a well-defined halo lithium plateau with modest scatter and a small population of outliers. The mass-dependent angular momentum loss law inferred from open cluster studies produces a nearly flat plateau, unlike previous models that exhibited a downward curvature for hotter temperatures in the 7Li-Teff plane. The overall depletion factor for the plateau stars is sensitive primarily to the solar initial angular momentum used in the calibration for the mixing diffusion coefficients. Uncertainties remain in the treatment of the internal angular momentum transport in the models, and the potential impact of these uncertainties on our results is discussed. The 6Li/7Li depletion ratio is also examined. We find that the dispersion in the plateau and the 6Li/7Li depletion ratio scale with the absolute 7Li depletion in the plateau, and we use observational data to set bounds on the 7Li depletion in main-sequence halo

BWR spent fuel storage cask performance test. Volume 2. Pre- and post-test decay heat, heat transfer, and shielding analyses

International Nuclear Information System (INIS)

Wiles, L.E.; Lombardo, N.J.; Heeb, C.M.; Jenquin, U.P.; Michener, T.E.; Wheeler, C.L.; Creer, J.M.; McCann, R.A.

1986-06-01

This report describes the decay heat, heat transfer, and shielding analyses conducted in support of performance testing of a Ridhihalgh, Eggers and Associates REA 2033 boiling water reactor (BWR) spent fuel storage cask. The cask testing program was conducted for the US Department of Energy (DOE) Commercial Spent Fuel Management Program by the Pacific Northwest Laboratory (PNL) and by General Electric at the latters' Morris Operation (GE-MO) as reported in Volume I. The analyses effort consisted of performing pretest calculations to (1) select spent fuel for the test; (2) symmetrically load the spent fuel assemblies in the cask to ensure lateral symmetry of decay heat generation rates; (3) optimally locate temperature and dose rate instrumentation in the cask and spent fuel assemblies; and (4) evaluate the ORIGEN2 (decay heat), HYDRA and COBRA-SFS (heat transfer), and QAD and DOT (shielding) computer codes. The emphasis of this second volume is on the comparison of code predictions to experimental test data in support of the code evaluation process. Code evaluations were accomplished by comparing pretest (actually pre-look, since some predictions were not completed until testing was in progress) predictions with experimental cask testing data reported in Volume I. No attempt was made in this study to compare the two heat transfer codes because results of other evaluations have not been completed, and a comparison based on one data set may lead to erroneous conclusions

Comparative economics for DUCRETE spent fuel storage cask handling, transportation, and capital requirements

International Nuclear Information System (INIS)

Powell, F.P.

1995-04-01

This report summarizes economic differences between a DUCRETE spent nuclear fuel storage cask and a conventional concrete storage cask in the areas of handling, transportation, and capital requirements. The DUCRETE cask is under evaluation as a new technology that could substantially reduce the overall costs of spent fuel and depleted U disposal. DUCRETE incorporates depleted U in a Portland cement mixture and functions as the cask's primary radiation barrier. The cask system design includes insertion of the US DOE Multi-Purpose Canister inside the DUCRETE cask. The economic comparison is from the time a cask is loaded in a spent fuel pool until it is placed in the repository and includes the utility and overall US system perspectives

"When the going gets tough, who keeps going?" Depletion sensitivity moderates the ego-depletion effect

NARCIS (Netherlands)

Salmon, Stefanie J.; Adriaanse, Marieke A.; De Vet, Emely; Fennis, Bob M.; De Ridder, Denise T D

2014-01-01

Self-control relies on a limited resource that can get depleted, a phenomenon that has been labeled ego-depletion. We argue that individuals may differ in their sensitivity to depleting tasks, and that consequently some people deplete their self-control resource at a faster rate than others. In

"When the going gets tough, who keeps going?" : Depletion sensitivity moderates the ego-depletion effect

NARCIS (Netherlands)

Salmon, Stefanie J.; Adriaanse, Marieke A.; De Vet, Emely; Fennis, Bob M.; De Ridder, Denise T. D.

2014-01-01

Self-control relies on a limited resource that can get depleted, a phenomenon that has been labeled ego-depletion. We argue that individuals may differ in their sensitivity to depleting tasks, and that consequently some people deplete their self-control resource at a faster rate than others. In

The modality effect of ego depletion: Auditory task modality reduces ego depletion.

Science.gov (United States)

Li, Qiong; Wang, Zhenhong

2016-08-01

An initial act of self-control that impairs subsequent acts of self-control is called ego depletion. The ego depletion phenomenon has been observed consistently. The modality effect refers to the effect of the presentation modality on the processing of stimuli. The modality effect was also robustly found in a large body of research. However, no study to date has examined the modality effects of ego depletion. This issue was addressed in the current study. In Experiment 1, after all participants completed a handgrip task, one group's participants completed a visual attention regulation task and the other group's participants completed an auditory attention regulation task, and then all participants again completed a handgrip task. The ego depletion phenomenon was observed in both the visual and the auditory attention regulation task. Moreover, participants who completed the visual task performed worse on the handgrip task than participants who completed the auditory task, which indicated that there was high ego depletion in the visual task condition. In Experiment 2, participants completed an initial task that either did or did not deplete self-control resources, and then they completed a second visual or auditory attention control task. The results indicated that depleted participants performed better on the auditory attention control task than the visual attention control task. These findings suggest that altering task modality may reduce ego depletion. © 2016 Scandinavian Psychological Associations and John Wiley & Sons Ltd.

On the problem of in-core fuel management in power reactors

International Nuclear Information System (INIS)

Marinkovic, N.; Matausek, M.V.

1985-01-01

Within the scope of in-core fuel management including refuelling schedule and reactivity control it is indispensable to define nuclear fuel worth, optimal depletion of the spent fuel assemblies as well as isotopic composition of the spent fuel. This paper shows the computed values of the mentioned parameters in case of different reactor types, PWR, WWER, HWR and BWR of 1000 MWe as well as the intensity of radiation of the spent fuel 3 and 1 years after fission.(author)

Muscle Mass Depletion Associated with Poor Outcome of Sepsis in the Emergency Department.

Science.gov (United States)

Lee, YoonJe; Park, Hyun Kyung; Kim, Won Young; Kim, Myung Chun; Jung, Woong; Ko, Byuk Sung

2018-05-08

Muscle mass depletion has been suggested to predict morbidity and mortality in various diseases. However, it is not well known whether muscle mass depletion is associated with poor outcome in sepsis. We hypothesized that muscle mass depletion is associated with poor outcome in sepsis. Retrospective observational study was conducted in an emergency department during a 9-year period. Medical records of 627 patients with sepsis were reviewed. We divided the patients into 2 groups according to 28-day mortality and compared the presence of muscle mass depletion assessed by the cross-sectional area of the psoas muscle at the level of the third lumbar vertebra on abdomen CT scans. Univariate and multivariate logistic regression analyses were conducted to examine the association of scarcopenia on the outcome of sepsis. A total of 274 patients with sepsis were finally included in the study: 45 (16.4%) did not survive on 28 days and 77 patients (28.1%) were identified as having muscle mass depletion. The presence of muscle mass depletion was independently associated with 28-day mortality on multivariate logistic analysis (OR 2.79; 95% CI 1.35-5.74, p = 0.01). Muscle mass depletion evaluated by CT scan was associated with poor outcome of sepsis patients. Further studies on the appropriateness of specific treatment for muscle mass depletion with sepsis are needed. © 2018 S. Karger AG, Basel.

Energy Return on Investment from Recycling Nuclear Fuel

International Nuclear Information System (INIS)

2011-01-01

This report presents an evaluation of the Energy Return on Investment (EROI) from recycling an initial batch of 800 t/y of used nuclear fuel (UNF) through a Recycle Center under a number of different fuel cycle scenarios. The study assumed that apart from the original 800 t of UNF only depleted uranium was available as a feed. Therefore for each subsequent scenario only fuel that was derived from the previous fuel cycle scenario was considered. The scenarios represent a good cross section of the options available and the results contained in this paper and associated appendices will allow for other fuel cycle options to be considered.

A study on the sensitivity depletion laws for rhodium self-powered neutron detectors

International Nuclear Information System (INIS)

Kim, Gil Gon

1999-02-01

The rhodium self-powered neutron detectors (SPND) in a reactor core provide the operator with the on-line 3-dimensional nuclear power distribution. The signal produced by rhodium SPND is interpreted into the local neutron flux by using a sensitivity depletion law and the local neutron flux is interpreted into the local power by using a power conversion factor. This work on the sensitivity depletion laws for rhodium self-powered neutron detectors (SPND) is performed to improve the uncertainty of the sensitivity depletion law used in ABB-CE reactors employing a rhodium SPND and to develop a calculational tool for providing the sensitivity depletion laws to interpret the signal of the newly designed rhodium SPND into the local neutron flux. The calculational tools for a time dependent neutron flux distribution in the rhodium emitter during depletion and for a time dependent beta escape probability that a beta generated in the emitter is escaped into the collector were developed. Due to the cost, the exposure to the radiation, and the longer fuel cycle, there is a strong incentive that the loading density of an in-core instrumentation is reduced and the lifetime of the detector is lengthened. These objectives can be achieved by reducing the uncertainty which is amplified as it depletes. The calculational tools above provide the sensitivity depletion law and show the reduction of the uncertainty to about 1 % in interpreting the signal into the local neutron flux compared to the method employed by ABB-CE. The reduction in the uncertainty of 1 % in interpreting the signal into the local neutron flux is equivalent to the reduction in the uncertainty of 1 % or more in interpreting the signal into the local power and to the extension of the lifetime of rhodium SPND to about 10 % as reported by ABB-CE

The Toxicity of Depleted Uranium

Directory of Open Access Journals (Sweden)

Wayne Briner

2010-01-01

Full Text Available Depleted uranium (DU is an emerging environmental pollutant that is introduced into the environment primarily by military activity. While depleted uranium is less radioactive than natural uranium, it still retains all the chemical toxicity associated with the original element. In large doses the kidney is the target organ for the acute chemical toxicity of this metal, producing potentially lethal tubular necrosis. In contrast, chronic low dose exposure to depleted uranium may not produce a clear and defined set of symptoms. Chronic low-dose, or subacute, exposure to depleted uranium alters the appearance of milestones in developing organisms. Adult animals that were exposed to depleted uranium during development display persistent alterations in behavior, even after cessation of depleted uranium exposure. Adult animals exposed to depleted uranium demonstrate altered behaviors and a variety of alterations to brain chemistry. Despite its reduced level of radioactivity evidence continues to accumulate that depleted uranium, if ingested, may pose a radiologic hazard. The current state of knowledge concerning DU is discussed.

Transient and steady-state analyses of an electrically heated Topaz-II Thermionic Fuel Element

International Nuclear Information System (INIS)

El-Genk, M.S.; Xue, H.

1992-01-01

Transient and steady-state analyses of electrically heated, Thermionic Fuel Elements (TFEs) for Topaz-II space power system are performed. The calculated emitter and collector temperatures, load electric power and conversion efficiency are in good agreement with reported data. In this paper the effects or Cs pressure, thermal power input, and load resistance on the steady-state performance of the TFE are also investigated. In addition, the thermal response of the ZrH moderator during a startup transient and following a change in the thermal power input is examined

Criticality calculations of various spent fuel casks - possibilities for burn up credit implementation

International Nuclear Information System (INIS)

Apostolov, T; Manolova, M.; Prodanova, R.

2001-01-01

A methodology for criticality safety analysis of spent fuel casks with possibilities for burnup credit implementation is presented. This methodology includes the world well-known and applied program systems: NESSEL-NUKO for depletion and SCALE-4.4 for criticality calculations. The abilities of this methodology to analyze storage and transportation casks with different type of spent fuel are demonstrated on the base of various tests. The depletion calculations have been carried out for the power reactors (WWER-440 and WWER-1000) and the research reactor IRT-2000 (C-36) fuel assemblies. The criticality calculation models have been developed on the basis of real fuel casks, designed by the leading international companies (for WWER-440 and WWER-1000 spent fuel assemblies), as well as for real a WWER-440 storage cask, applied at the 'Kozloduy' NPP. The results obtained show that the criticality safety criterion K eff less than 0.95 is satisfied for both: fresh and spent fuel. Besides the implementation of burnup credit allows to account for the reduced reactivity of spent fuel and to evaluate the conservatism of the fresh fuel assumption. (author)

A vapor feed methanol microfluidic fuel cell with high fuel and energy efficiency

International Nuclear Information System (INIS)

Wang, Yifei; Leung, Dennis Y.C.; Xuan, Jin; Wang, Huizhi

2015-01-01

Highlights: • A microfluidic fuel cell with a vapor feed anode is investigated. • Its advantages include simpler design, direct usage of methanol and better performance. • The prototype achieves a peak power density of 55.4 mW cm −2 under room temperature. • The energy efficiency of 9.4% is much higher than its liquid feed counterpart. - Abstract: In this paper, a prototype of methanol microfluidic fuel cell with vapor feed anode configuration is proposed to improve the fuel and energy efficiency of the conventional liquid feed methanol microfluidic fuel cells. Peak power density of 55.4 mW cm −2 can be achieved with this prototype under room temperature, which is 30% higher than its conventional liquid feed counterpart. Moreover, an energy efficiency of 9.4% is achieved, which is 27.5 times higher than its liquid feed counterpart. This superiority on both cell performance and energy efficiency is directly benefitted from its vapor feed anode configuration, which alleviates the fuel crossover, eliminates the fuel depletion boundary layer, and avoids the bulk anolyte wastage. The tradeoff between cell performance and fuel utilization for conventional liquid feed microfluidic fuel cells is also evaded

Preliminary investigation of fuel cycle in fast reactors by the correlations method and sensitivity analyses of nuclear characteristics

International Nuclear Information System (INIS)

Amorim, E.S. do; Castro Lobo, P.D. de.

1980-11-01

A reduction of computing effort was achieved as a result of the application of space - independent continuous slowing down theory in the spectrum averaged cross sections and further expressing then in a quadratic corelation whith the temperature and the composition. The decoupling between variables that express some of the important nuclear characteristics allowed to introduce a sensitivity analyses treatment for the full prediction of the behavior, over the fuel cycle, of the LMFBR considered. As a potential application of the method here in developed is to predict the nuclear characteristics of another reactor, face some reference reactor of the family considered. Excellent agreement with exact calculation is observed only when perturbations occur in nuclear data and/or fuel isotopic characteristics, but fair results are obtained whith variations in system components other than the fuel. (Author) [pt

Nuclear fuel

International Nuclear Information System (INIS)

Azevedo, J.B.L. de.

1980-01-01

All stages of nuclear fuel cycle are analysed with respect to the present situation and future perspectives of supply and demand of services; the prices and the unitary cost estimation of these stages for the international fuel market are also mentioned. From the world resources and projections of uranium consumption, medium-and long term analyses are made of fuel availability for several strategies of use of different reactor types. Finally, the cost of nuclear fuel in the generation of electric energy is calculated to be used in the energetic planning of the electric sector. (M.A.) [pt

News technology utilization fossil fuel

Directory of Open Access Journals (Sweden)

Blišanová Monika

2004-09-01

Full Text Available Fossil fuel – “alternative energy“ is coal, petroleum, natural gas. Petroleum and natural gas are scarce resources, but they are delimited. Reserves petroleum will be depleted after 39 years and reserves natural gas after 60 years.World reserves coal are good for another 240 years. Coal is the most abundant fossil fuel. It is the least expensive energy source for generating electricity. Many environmental problems associated with use of coal:in coal production, mining creates environmental problems.On Slovakia representative coal only important internal fuel – power of source and coal is produced in 5 locality. Nowadays, oneself invest to new technology on utilization coal. Perspective solution onself shows UCG, IGCC.

Comparative Analysis of VERA Depletion Problems

International Nuclear Information System (INIS)

Park, Jinsu; Kim, Wonkyeong; Choi, Sooyoung; Lee, Hyunsuk; Lee, Deokjung

2016-01-01

Each code has its own solver for depletion, which can produce different depletion calculation results. In order to produce reference solutions for depletion calculation comparison, sensitivity studies should be preceded for each depletion solver. The sensitivity tests for burnup interval, number of depletion zones, and recoverable energy per fission (Q-value) were performed in this paper. For the comparison of depletion calculation results, usually the multiplication factors are compared as a function of burnup. In this study, new comparison methods have been introduced by using the number density of isotope or element, and a cumulative flux instead of burnup. In this paper, optimum depletion calculation options are determined through the sensitivity study of the burnup intervals and the number of depletion intrazones. Because the depletion using CRAM solver performs well for large burnup intervals, smaller number of burnup steps can be used to produce converged solutions. It was noted that the depletion intra-zone sensitivity is only pin-type dependent. The 1 and 10 depletion intra-zones for the normal UO2 pin and gadolinia rod, respectively, are required to obtain the reference solutions. When the optimized depletion calculation options are used, the differences of Q-values are found to be a main cause of the differences of solutions. In this paper, new comparison methods were introduced for consistent code-to-code comparisons even when different kappa libraries were used in the depletion calculations

CIEMAT analyses of transition fuel cycle scenarios

International Nuclear Information System (INIS)

Alvarez-Velarde, F.; Gonzalez-Romero, E.M.

2010-01-01

The efficient design of strategies for the long-term sustainability of nuclear energy or the phase-out of this technology is possible after the study of transition scenarios from the current fuel cycle to a future one with advanced technologies and concepts. CIEMAT has participated in numerous fuel cycle scenarios studies for more than a decade and, from some years ago, special attention has been put in the study of transition scenarios. In this paper, the main characteristics of each studied transition scenario are described. The main results and partial conclusions of each scenario are also analyzed. As general conclusions of transition studies, we highlight that the advantages of advanced technologies in transition scenarios can be obtained by countries or regions with sufficiently large nuclear parks, with a long-term implementation of the strategy. For small countries, these advantages are also accessible with an affordable cost, by means of the regional collaboration during several decades. (authors)

Analysis of Core Physics Experiments on Irradiated BWR MOX Fuel in REBUS Program

International Nuclear Information System (INIS)

Yamamoto, Toru; Ando, Yoshihira; Hayashi, Yamato

2008-01-01

As part of analyses of experimental data of a critical core containing a irradiated BWR MOX test bundle in the REBUS program, depletion calculations was performed for the BWR MOX fuel assemblies from that the MOX test rods were selected by using a general purpose neutronics code system SRAC. The core analyses were carried out using SRAC and a continuous energy Monte Carlo code MVP. The calculated k eff s were compared with those of the core containing a fresh MOX fuel bundle in the program. The SRAC-diffusion calculation underestimates k eff s of the both cores by 1.0 to 1.3 %dk and the k eff s of MVP are 1.001. The difference in k eff between the irradiated BWR MOX test bundle core and the fresh MOX one is 0.4 %dk in the SRAC-diffusion calculation and 0.0 %dk in the MVP calculation. The calculated fission rate distributions are in good agreement with the measurement in the SRAC-diffusion and MVP calculations. The calculated neutron flux distributions are also in good agreement with the measurement. The calculated burnup reactivity in the both calculations well reproduce the measurements. (authors)

Some implications of batch average burnup calculations on predicted spent fuel compositions

International Nuclear Information System (INIS)

Alexander, C.W.; Croff, A.G.

1984-01-01

The accuracy of using batch-averaged burnups to determine spent fuel characteristics (such as isotopic composition, activity, etc.) was examined for a typical pressurized-water reactor (PWR) fuel discharge batch by comparing characteristics computed by (a) performing a single depletion calculation using the average burnup of the spent fuel and (b) performing separate depletion calculations based on the relative amounts of spent fuel in each of twelve burnup ranges and summing the results. The computations were done using ORIGEN 2. Procedure (b) showed a significant shift toward a greater quantity of the heavier transuranics, which derive from multiple neutron captures, and a corresponding decrease in the amounts of lower transuranics. Those characteristics which derive primarily from fission products, such as total radioactivity and total thermal power, are essentially identical for the two procedures. Those characteristics that derive primarily from the heavier transuranics, such as spontaneous fission neutrons, are underestimated by procedure (a)

A higher order depletion perturbation theory with application to in-core fuel management optimization

International Nuclear Information System (INIS)

Kropaczek, D.J.; Turinsky, P.J.

1990-01-01

Perturbation techniques utilized in reactor analysis have recently been applied in the solution of the in-core nuclear fuel management optimization problem. The use of such methods is motivated by the need to evaluate many times over, the core physics characteristics of loading pattern solutions obtained through an optimization process, which is typically iterative. Perturbation theory provides an efficient alternative to the prohibitively expensive, repetitive solutions of the system few-group neutron diffusion equations required in solving the fuel placement problem. A primary concern in the use of such methods is the control of perturbation errors arising during the fuel shuffling process. First-order accurate models inevitably resort to undue restriction of fuel movement during the optimization process to control these errors. Higher order perturbation theory models have the potential to overcome such limitations, which may result in the identification of local versus global optima. An accurate, computationally efficient reactor physics model based on higher order perturbation theory and geared toward the needs of large-scale in-core fuel management optimization is presented in this paper

When will fossil fuel reserves be diminished?

International Nuclear Information System (INIS)

Shafiee, Shahriar; Topal, Erkan

2009-01-01

Crude oil, coal and gas are the main resources for world energy supply. The size of fossil fuel reserves and the dilemma that 'when non-renewable energy will be diminished' is a fundamental and doubtful question that needs to be answered. This paper presents a new formula for calculating when fossil fuel reserves are likely to be depleted and develops an econometrics model to demonstrate the relationship between fossil fuel reserves and some main variables. The new formula is modified from the Klass model and thus assumes a continuous compound rate and computes fossil fuel reserve depletion times for oil, coal and gas of approximately 35, 107 and 37 years, respectively. This means that coal reserves are available up to 2112, and will be the only fossil fuel remaining after 2042. In the Econometrics model, the main exogenous variables affecting oil, coal and gas reserve trends are their consumption and respective prices between 1980 and 2006. The models for oil and gas reserves unexpectedly show a positive and significant relationship with consumption, while presenting a negative and significant relationship with price. The econometrics model for coal reserves, however, expectedly illustrates a negative and significant relationship with consumption and a positive and significant relationship with price. Consequently, huge reserves of coal and low-level coal prices in comparison to oil and gas make coal one of the main energy substitutions for oil and gas in the future, under the assumption of coal as a clean energy source

Mathematical modeling of solid oxide fuel cells

Science.gov (United States)

Lu, Cheng-Yi; Maloney, Thomas M.

1988-01-01

Development of predictive techniques, with regard to cell behavior, under various operating conditions is needed to improve cell performance, increase energy density, reduce manufacturing cost, and to broaden utilization of various fuels. Such technology would be especially beneficial for the solid oxide fuel cells (SOFC) at it early demonstration stage. The development of computer models to calculate the temperature, CD, reactant distributions in the tubular and monolithic SOFCs. Results indicate that problems of nonuniform heat generation and fuel gas depletion in the tubular cell module, and of size limitions in the monolithic (MOD 0) design may be encountered during FC operation.

The secondary stress analyses in the fuel pin cladding due to the swelling gradient across the wall thickness

International Nuclear Information System (INIS)

Uwaba, Tomoyuki; Ukai, Shigeharu

2002-01-01

Irradiation deformation analyses of FBR fuel cladding were made by using the finite element method. In these analyses the history of the stress occurred in the cladding was evaluated paying attention to the secondary stress induced by the swelling difference across the wall thickness. It was revealed that the difference of the swelling incubation dose in the direction of the thickness and the irradiation creep deformation play an important role in the history of the secondary stress. The effect of the stress-enhanced swelling was also analyzed in this study

Sensitivity theory for reactor burnup analysis based on depletion perturbation theory

International Nuclear Information System (INIS)

Yang, Wonsik.

1989-01-01

The large computational effort involved in the design and analysis of advanced reactor configurations motivated the development of Depletion Perturbation Theory (DPT) for general fuel cycle analysis. The work here focused on two important advances in the current methods. First, the adjoint equations were developed for using the efficient linear flux approximation to decouple the neutron/nuclide field equations. And second, DPT was extended to the constrained equilibrium cycle which is important for the consistent comparison and evaluation of alternative reactor designs. Practical strategies were formulated for solving the resulting adjoint equations and a computer code was developed for practical applications. In all cases analyzed, the sensitivity coefficients generated by DPT were in excellent agreement with the results of exact calculations. The work here indicates that for a given core response, the sensitivity coefficients to all input parameters can be computed by DPT with a computational effort similar to a single forward depletion calculation

Development of methods for theoretical analysis of nuclear reactors (Phase II), I-V, Part IV, Fuel depletion; Razrada metoda teorijske analize nuklearnih reaktora (II faza), I-V, IV Deo, Promena izotopnog sastava goriva

Energy Technology Data Exchange (ETDEWEB)

Pop-Jordanov, J [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

1962-10-15

This report includes the analysis of plutonium isotopes from U{sup 238} depletion chain. Two theoretical approaches for solving the depletion of fuel are shown. One results in the system of differential equations that can be solved only by using electronic calculators and the second, Machinari-Goto method enables obtaining analytical equations for approximative values of particular nuclei. In addition, differential equations are given for different approximation levels in calculating Pu {sup 239}, as well as relations between the released energy and irradiation. Ova faza obuhvata analizu stvaranja izotopa plutonijuma u lancu U{sup 238}. Prikazana su dva teorijska pristupa resavanju problema 'konverzije goriva', jedan dovodi do sistema diferecijalnih jednacina za cije je resavanje neophodno koriscenje elektronskih racunskih masina, i drugi, Machinari-Goto metod koji omogucava da se dobiju analiticki izrazi vrednosti aproksimacije pojedinih jezgara. Osim toga date su diferencijalne jednacine raznih stepena aproksimacije u racunanju Pu {sup 239}, kao i veze izmedju oslobodjene energije i ozracivanja.

Hydrocarbons and fuels analyses with the supersonic gas chromatography mass spectrometry--the novel concept of isomer abundance analysis.

Science.gov (United States)

Fialkov, Alexander B; Gordin, Alexander; Amirav, Aviv

2008-06-27

Hydrocarbon analysis with standard GC-MS is confronted by the limited range of volatile compounds amenable for analysis and by the similarity of electron ionization mass spectra for many compounds which show weak or no molecular ions for heavy hydrocarbons. The use of GC-MS with supersonic molecular beams (Supersonic GC-MS) significantly extends the range of heavy hydrocarbons that can be analyzed, and provides trustworthy enhanced molecular ion to all hydrocarbons. In addition, unique isomer mass spectral features are obtained in the ionization of vibrationally cold hydrocarbons. The availability of molecular ions for all hydrocarbons results in the ability to obtain unique chromatographic isomer distribution patterns that can serve as a new method for fuel characterization and identification. Examples of the applicability and use of this novel isomer abundance analysis (IAA) method to diesel fuel, kerosene and oil analyses are shown. It is suggested that in similarity to the "three ions method" for identification purposes, three isomer abundance patterns can serve for fuel characterization. The applications of the Supersonic GC-MS for engine motor oil analysis and transformer oil analysis are also demonstrated and discussed, including the capability to achieve fast 1-2s sampling without separation for oil and fuel fingerprinting. The relatively fast analysis of biodiesel is described, demonstrating the provision of molecular ions to heavy triglycerides. Isomer abundance analysis with the Supersonic GC-MS could find broad range of applications including petrochemicals and fuel analysis, arson analysis, environmental oil/fuel spill analysis, fuel adulteration analysis and motor oil analysis.

Natural uranium equivalent fuel an innovative design for proven CANDU technology

Energy Technology Data Exchange (ETDEWEB)

Pineiro, F.; Ho, K.; Khaial, A.; Boubcher, M.; Cottrell, C.; Kuran, S., E-mail: fabricia.pineiro@candu.com [Candu Energy Inc., Mississauga, ON (Canada); Zhenhua, Z.; Zhiliang, M. [Third Qinshan Nuclear Power Company, Haiyan, Zhejiang (China)

2015-07-01

The high neutron economy, on-power refuelling capability and fuel bundle design simplicity in CANDU reactors allow for the efficient utilization of alternative fuels. Candu Energy Inc. (Candu), in collaboration with the Third Qinshan Nuclear Power Company (TQNPC), the China North Nuclear Fuel Corporation (CNNFC), and the Nuclear Power Institute of China (NPIC), has successfully developed an advanced fuel called Natural Uranium Equivalent (NUE). This innovative design consists of a mixture of recycled and depleted uranium, which can be implemented in existing CANDU stations thereby bringing waste products back into the energy stream, increasing fuel resources diversity and reducing fuel costs. (author)

Natural uranium equivalent fuel. An innovative design for proven CANDU technology

Energy Technology Data Exchange (ETDEWEB)

Pineiro, F.; Ho, K.; Khaial, A.; Boubcher, M.; Cottrell, C.; Kuran, S. [Candu Energy Inc., Mississauga, Ontario (Canada); Zhenhua, Z.; Zhiliang, M. [Third Qinshan Nuclear Power Co., Haiyan, Zhejiang (China)

2015-09-15

The high neutron economy, on-power refuelling capability and fuel bundle design simplicity in CANDU® reactors allow for the efficient utilization of alternative fuels. Candu Energy Inc. (Candu), in collaboration with the Third Qinshan Nuclear Power Company (TQNPC), the China North Nuclear Fuel Corporation (CNNFC), and the Nuclear Power Institute of China (NPIC), has successfully developed an advanced fuel called Natural Uranium Equivalent (NUE). This innovative design consists of a mixture of recycled and depleted uranium, which can be implemented in existing CANDU stations thereby bringing waste products back into the energy stream, increasing fuel resources diversity and reducing fuel costs. (author)

A model finite-element to analyse the mechanical behavior of a PWR fuel rod

International Nuclear Information System (INIS)

Galeao, A.C.N.R.; Tanajura, C.A.S.

1988-01-01

A model to analyse the mechanical behavior of a PWR fuel rod is presented. We drew our attention to the phenomenon of pellet-pellet and pellet-cladding contact by taking advantage of an elastic model which include the effects of thermal gradients, cladding internal and external pressures, swelling and initial relocation. The problem of contact gives rise ro a variational formulation which employs Lagrangian multipliers. An iterative scheme is constructed and the finite element method is applied to obtain the numerical solution. Some results and comments are presented to examine the performance of the model. (author) [pt

Depletion of fossil fuels and the impacts of global warming

International Nuclear Information System (INIS)

Hoel, M.; Kverndokk, S.

1996-01-01

This paper combines the theory of optimal extraction of exhaustible resources with the theory of greenhouse externalities, to analyze problems of global warming when the supply side is considered. The optimal carbon tax will initially rise but eventually fall when the externality is positively related to the stock of carbon in the atmosphere. It is shown that the tax will start falling before the stock of carbon in the atmosphere reaches its maximum. If there exists a non-polluting backstop technology, it will be optimal to extract and consume fossil fuels even when the price of fossil fuels is equal to the price of the backstop. The total extraction is the same as when the externality is ignored, but in the presence of the greenhouse effect, it will be optimal to slow the extraction and spread it over a longer period. If, on the other hand, the greenhouse externality depends on the rate of change in the atmospheric stock of carbon, the evolution of the optimal carbon tax is more complex. It can even be optimal to subsidize carbon emissions to avoid future rapid changes in the stock of carbon, and therefore future damages. 22 refs., 3 figs

Effect of fission yield libraries on the irradiated fuel composition in Monte Carlo depletion calculations

International Nuclear Information System (INIS)

Mitenkova, E.; Novikov, N.

2014-01-01

Improving the prediction of radiation parameters and reliability of fuel behaviour under different irradiation modes is particularly relevant for new fuel compositions, including recycled nuclear fuel. For fast reactors there is a strong dependence of nuclide accumulations on the nuclear data libraries. The effect of fission yield libraries on irradiated fuel is studied in MONTEBURNS-MCNP5-ORIGEN2 calculations of sodium fast reactors. Fission yield libraries are generated for sodium fast reactors with MOX fuel, using ENDF/B-VII.0, JEFF3.1, original library FY-Koldobsky, and GEFY 3.3 as sources. The transport libraries are generated from ENDF/B-VII.0 and JEFF-3.1. Analysis of irradiated MOX fuel using different fission yield libraries demonstrates the considerable spread in concentrations of fission products. The discrepancies in concentrations of inert gases being ∼25%, up to 5 times for stable and long-life nuclides, and up to 10 orders of magnitude for short-lived nuclides. (authors)

Light Nonaqueous-Phase Liquid Weathering at Various Fuel Release Sites

National Research Council Canada - National Science Library

Henry, Bruce

1999-01-01

...) contracted with Parsons ES to perform this fuels weathering study. Of particular interest for this study is the weathering or natural depletion of benzene, toluene, ethylbenzene, and xylenes (BTEX...

Ozone-depleting Substances (ODS)

Data.gov (United States)

U.S. Environmental Protection Agency — This site includes all of the ozone-depleting substances (ODS) recognized by the Montreal Protocol. The data include ozone depletion potentials (ODP), global warming...

Comparative analysis of LWR and FBR spent fuels for nuclear forensics evaluation

International Nuclear Information System (INIS)

Permana, Sidik; Suzuki, Mitsutoshi; Su'ud, Zaki

2012-01-01

Some interesting issues are attributed to nuclide compositions of spent fuels from thermal reactors as well as fast reactors such as a potential to reuse as recycled fuel, and a possible capability to be manage as a fuel for destructive devices. In addition, analysis on nuclear forensics which is related to spent fuel compositions becomes one of the interesting topics to evaluate the origin and the composition of spent fuels from the spent fuel foot-prints. Spent fuel compositions of different fuel types give some typical spent fuel foot prints and can be estimated the origin of source of those spent fuel compositions. Some technics or methods have been developing based on some science and technological capability including experimental and modeling or theoretical aspects of analyses. Some foot-print of nuclear forensics will identify the typical information of spent fuel compositions such as enrichment information, burnup or irradiation time, reactor types as well as the cooling time which is related to the age of spent fuels. This paper intends to evaluate the typical spent fuel compositions of light water (LWR) and fast breeder reactors (FBR) from the view point of some foot prints of nuclear forensics. An established depletion code of ORIGEN is adopted to analyze LWR spent fuel (SF) for several burnup constants and decay times. For analyzing some spent fuel compositions of FBR, some coupling codes such as SLAROM code, JOINT and CITATION codes including JFS-3-J-3.2R as nuclear data library have been adopted. Enriched U-235 fuel composition of oxide type is used for fresh fuel of LWR and a mixed oxide fuel (MOX) for FBR fresh fuel. Those MOX fuels of FBR come from the spent fuels of LWR. Some typical spent fuels from both LWR and FBR will be compared to distinguish some typical foot-prints of SF based on nuclear forensic analysis.

Comparative analysis of LWR and FBR spent fuels for nuclear forensics evaluation

Energy Technology Data Exchange (ETDEWEB)

Permana, Sidik; Suzuki, Mitsutoshi; Su' ud, Zaki [Department of Science and Technology for Nuclear Material Management (STNM), Japan Atomic Energy Agency (JAEA), 2-4 Shirane, Shirakata, Tokai Mura, Naka-gun, Ibaraki 319-1195 Nuclear Physics and Bio (Indonesia); Department of Science and Technology for Nuclear Material Management (STNM), Japan Atomic Energy Agency (JAEA), 2-4 Shirane, Shirakata, Tokai Mura, Naka-gun, Ibaraki 319-1195 (Japan); Nuclear Physics and Bio Physics Research Group, Department of Physics, Bandung Institute of Technology, Gedung Fisika, Jl. Ganesha 10, Bandung 40132 (Indonesia)

2012-06-06

Some interesting issues are attributed to nuclide compositions of spent fuels from thermal reactors as well as fast reactors such as a potential to reuse as recycled fuel, and a possible capability to be manage as a fuel for destructive devices. In addition, analysis on nuclear forensics which is related to spent fuel compositions becomes one of the interesting topics to evaluate the origin and the composition of spent fuels from the spent fuel foot-prints. Spent fuel compositions of different fuel types give some typical spent fuel foot prints and can be estimated the origin of source of those spent fuel compositions. Some technics or methods have been developing based on some science and technological capability including experimental and modeling or theoretical aspects of analyses. Some foot-print of nuclear forensics will identify the typical information of spent fuel compositions such as enrichment information, burnup or irradiation time, reactor types as well as the cooling time which is related to the age of spent fuels. This paper intends to evaluate the typical spent fuel compositions of light water (LWR) and fast breeder reactors (FBR) from the view point of some foot prints of nuclear forensics. An established depletion code of ORIGEN is adopted to analyze LWR spent fuel (SF) for several burnup constants and decay times. For analyzing some spent fuel compositions of FBR, some coupling codes such as SLAROM code, JOINT and CITATION codes including JFS-3-J-3.2R as nuclear data library have been adopted. Enriched U-235 fuel composition of oxide type is used for fresh fuel of LWR and a mixed oxide fuel (MOX) for FBR fresh fuel. Those MOX fuels of FBR come from the spent fuels of LWR. Some typical spent fuels from both LWR and FBR will be compared to distinguish some typical foot-prints of SF based on nuclear forensic analysis.

Fueling-Controlled the Growth of Massive Black Holes

Science.gov (United States)

Escala, A.

2009-05-01

We study the relation between nuclear massive black holes and their host spheroid gravitational potential. Using AMR numerical simulations, we analyze how gas is transported into the nuclear (central kpc) regions of galaxies. We study gas fueling onto the inner accretion disk (sub-pc scale) and star formation in a massive nuclear disk like those generally found in proto-spheroids (ULIRGs, SCUBA Galaxies). These sub-pc resolution simulations of gas fueling, which is mainly depleted by star formation, naturally satisfy the `M_BH-M_{virial}' relation, with a scatter considerably less than that observed. We find that a generalized version of the Kennicutt-Schmidt Law for starbursts is satisfied, in which the total gas depletion rate (dot M_gas=dot M_BH + M_SF scales as M_gas/t_orbital. See Escala (2007) for more details about this work.

PAPIRUS - a computer code for FBR fuel performance analysis

International Nuclear Information System (INIS)

Kobayashi, Y.; Tsuboi, Y.; Sogame, M.

1991-01-01

The FBR fuel performance analysis code PAPIRUS has been developed to design fuels for demonstration and future commercial reactors. A pellet structural model was developed to describe the generation, depletion and transport of vacancies and atomic elements in unified fashion. PAPIRUS results in comparison with the power - to - melt test data from HEDL showed validity of the code at the initial reactor startup. (author)

Heat and radiation analysis of NPP Krsko irradiated fuel

International Nuclear Information System (INIS)

Lalovic, M.

1986-01-01

Radioactive and heat potential for irradiated fuel in the region 2 with burnup of 13400 MWd/tHM, and in the region 4A with burnup of 9360 MWd/tHM for NPP KRSKO, was calculated. Computer code KORIGEN (Karlsruhe Oak Ridge Isotope Generation and Depletion Code) was used. The aspects of radiation (mainly gamma and neutrons) and of heat production was considered with respect to their impact on fuel handing and waste management. Isotopic concentrations for irradiated fuel was calculated and compared with Westinghouse data. (author)

Analysis of beryllium and depleted uranium: An overview of detection methods in aerosols and soils

International Nuclear Information System (INIS)

Camins, I.; Shinn, J.H.

1988-06-01

We conducted a survey of commercially available methods for analysis of beryllium and depleted uranium in aerosols and soils to find a reliable, cost-effective, and sufficiently precise method for researchers involved in environmental testing at the Yuma Proving Ground, Yuma, Arizona. Criteria used for evaluation include cost, method of analysis, specificity, sensitivity, reproducibility, applicability, and commercial availability. We found that atomic absorption spectrometry with graphite furnace meets these criteria for testing samples for beryllium. We found that this method can also be used to test samples for depleted uranium. However, atomic absorption with graphite furnace is not as sensitive a measurement method for depleted uranium as it is for beryllium, so we recommend that quality control of depleted uranium analysis be maintained by testing 10 of every 1000 samples by neutron activation analysis. We also evaluated 45 companies and institutions that provide analyses of beryllium and depleted uranium. 5 refs., 1 tab

Advanced fuel cycles and burnup increase of WWER-440 fuel

International Nuclear Information System (INIS)

Proselkov, V.; Saprykin, V.; Scheglov, A.

2003-01-01

Analyses of operational experience of 4.4% enriched fuel in the 5-year fuel cycle at Kola NPP Unit 3 and fuel assemblies with Uranium-Gadolinium fuel at Kola NPP Unit 4 are made. The operability of WWER-440 fuel under high burnup is studied. The obtained results indicate that the fuel rods of WWER-440 assemblies intended for operation within six years of the reviewed fuel cycle totally preserve their operability. Performed analyses have demonstrated the possibility of the fuel rod operability during the fuel cycle. 12 assemblies were loaded into the reactor unit of Kola 3 in 2001. The predicted burnup in six assemblies was 59.2 MWd/kgU. Calculated values of the burnup after operation for working fuel assemblies were ∼57 MWd/kgU, for fuel rods - up to ∼61 MWd/kgU. Data on the coolant activity, specific activity of the benchmark iodine radionuclides of the reactor primary circuit, control of the integrity of fuel rods of the assemblies that were operated for six years indicate that not a single assembly has reached the criterion for the early discharge

Spent Fuel Ratio Estimates from Numerical Models in ALE3D

Energy Technology Data Exchange (ETDEWEB)

Margraf, J. D. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Dunn, T. A. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

2016-08-02

Potential threat of intentional sabotage of spent nuclear fuel storage facilities is of significant importance to national security. Paramount is the study of focused energy attacks on these materials and the potential release of aerosolized hazardous particulates into the environment. Depleted uranium oxide (DUO₂) is often chosen as a surrogate material for testing due to the unreasonable cost and safety demands for conducting full-scale tests with real spent nuclear fuel. To account for differences in mechanical response resulting in changes to particle distribution it is necessary to scale the DUO₂ results to get a proper measure for spent fuel. This is accomplished with the spent fuel ratio (SFR), the ratio of respirable aerosol mass released due to identical damage conditions between a spent fuel and a surrogate material like depleted uranium oxide (DUO₂). A very limited number of full-scale experiments have been carried out to capture this data, and the oft-questioned validity of the results typically leads to overly-conservative risk estimates. In the present work, the ALE3D hydrocode is used to simulate DUO₂ and spent nuclear fuel pellets impacted by metal jets. The results demonstrate an alternative approach to estimate the respirable release fraction of fragmented nuclear fuel.

Analysis of changes in the fuel component of the cost of electricity in the transition to a closed fuel cycle in nuclear power system

International Nuclear Information System (INIS)

Gurin, Andrey V.; Alekseev, P.N.

2017-01-01

This paper presents a study of scenarios of transition to a closed fuel cycle in the system of nuclear power, built basing on resource availability requirements at the stage of full life-cycle reactors. Conventionally, there are three main scenarios for the development of nuclear energy: with VVER reactors operating in an open fuel cycle; with VVER reactors operating in a closed fuel cycle; and co-operating VVER and BN, operating in a closed fuel cycle. For the considered scenarios, a quantitative estimation of change in time of material balances were performed, including spent fuel balance, balance of plutonium, reprocessed and depleted uranium, radioactive waste, and the analysis of the fuel component of the cost of electricity.

Analysis of changes in the fuel component of the cost of electricity in the transition to a closed fuel cycle in nuclear power system

Energy Technology Data Exchange (ETDEWEB)

Gurin, Andrey V. [National Research Centre ' ' Kurchatov Institute' ' , Moscow (Russian Federation); Alekseev, P.N.

2017-09-15

This paper presents a study of scenarios of transition to a closed fuel cycle in the system of nuclear power, built basing on resource availability requirements at the stage of full life-cycle reactors. Conventionally, there are three main scenarios for the development of nuclear energy: with VVER reactors operating in an open fuel cycle; with VVER reactors operating in a closed fuel cycle; and co-operating VVER and BN, operating in a closed fuel cycle. For the considered scenarios, a quantitative estimation of change in time of material balances were performed, including spent fuel balance, balance of plutonium, reprocessed and depleted uranium, radioactive waste, and the analysis of the fuel component of the cost of electricity.

Safety evaluation for packaging (onsite) depleted uranium waste boxes

Energy Technology Data Exchange (ETDEWEB)

McCormick, W.A.

1997-08-27

This safety evaluation for packaging (SEP) allows the one-time shipment of ten metal boxes and one wooden box containing depleted uranium material from the Fast Flux Test Facility to the burial grounds in the 200 West Area for disposal. This SEP provides the analyses and operational controls necessary to demonstrate that the shipment will be safe for the onsite worker and the public.

Safety evaluation for packaging (onsite) depleted uranium waste boxes

International Nuclear Information System (INIS)

McCormick, W.A.

1997-01-01

This safety evaluation for packaging (SEP) allows the one-time shipment of ten metal boxes and one wooden box containing depleted uranium material from the Fast Flux Test Facility to the burial grounds in the 200 West Area for disposal. This SEP provides the analyses and operational controls necessary to demonstrate that the shipment will be safe for the onsite worker and the public

“When the going gets tough, who keeps going?” Depletion sensitivity moderates the ego-depletion effect

NARCIS (Netherlands)

Salmon, S.J.; Adriaanse, M.A.; Vet, de E.W.M.L.; Fennis, B.M.; Ridder, de D.T.D.

2014-01-01

Self-control relies on a limited resource that can get depleted, a phenomenon that has been labeled ego-depletion. We argue that individuals may differ in their sensitivity to depleting tasks, and that consequently some people deplete their self-control resource at a faster rate than others. In

Analyses of subchannel velocity distribution for HANARO fuel assembly

International Nuclear Information System (INIS)

Chae, Hee Taek; Han, Gee Yang; Park, Cheol; Lim, In Cheol

1998-10-01

MATRA-h which is a subchannel analysis computer code is used to evaluate the thermal margin of HANARO core. To estimate core thermal margin, accurate prediction of subchannel velocity is very important. The average subchannel velocities of 18 element fuel assembly were obtained from the results of velocity measurement test. To validate the adequacy of the hydraulic model code predictions were compared with the experimental results for the subchannel velocity distribution in 18 element fuel channel. The calculated subchannel velocity distributions in the central channels were larger than those of experiment. On the other hand the subchannel velocities in the outer channels were smaller. It is speculated that the prediction like as above would make CHF value lower because CHF phenomena had been occurred in the outer fuel element in the bundle CHF test of AECL. The prediction for axial pressure distribution coincided with the experimental results well. (author). 9 refs., 9 tabs., 14 figs

Decontamination of Cape Arza (Montenegro) from depleted Uranium

International Nuclear Information System (INIS)

Vukotich, P.; Kovachevich, M.; Vasich, V.; Ristich, N.

2002-01-01

On May 30, 1999, NATO A-10 aircrafts attacked Cape Arza, a very attractive touring area on peninsula Lustica, at the entrance of Boka Kotorska Bay, in Montenegro. They fired anti-armour rounds with penetrators made of depleted uranium. Such an armour-penetrating round has a length of 173 mm and a diameter of 30 mm. The bullet has an aluminium case (jacket) and inside it a conical DU penetrator. The length of the penetrator itself is 95 mm, and the diameter of its base is 16 mm. The penetrator weight is 292 g. According to the data reported by NATO (NATO, 2001), the total number of rounds fired against Cape Arza was 480. As to the data on combat mix of the A-10 aircraft gun, 300 (UNEP, 2001) or 400 (UNEP, 2001; FAS) of these rounds where with DU penetrators, and the rest with a classical charge. This means that Cape Arza was contaminated with 90 or 120 kg of DU, or with a radioactivity of (3.5 - 4.7) · 10 9 Bq. Depleted uranium is a waste product of the process of uranium enrichment in 2 35U isotope, for use in nuclear reactors or in nuclear weapons. The isotopic composition of depleted uranium is (Harley et al., 1999): (99.7 - 99.8) % of 2 38U , (0.2 - 0.3) % of 2 35U , 0.001 % of 2 34U , and only traces of 2 34T h, 2 34P a and 2 31T h. If traces of the isotopes 2 36U , 2 39P u and 2 40P u are also present, as it is the case with DU from Cape Arza (UNEP, 2002), the depleted uranium is obtained by reprocessing of spent nuclear reactor fuel. The activity concentration of depleted uranium is 39.42 · 10 6 Bq/kg. Most of it comes from 2 38U and its decay products 2 34T h and 2 34P a which are in radioactive equilibrium (12.27 · 10 6 Bq/kg per each of them), and the less part from 2 35U and 2 31T h (0.16 · 10 6 Bq/kg per each) (UNEP, 1999), while the activity concentration of 2 36U , 2 39P u and 2 40P u is below 100 Bq/kg (UNEP, 2001)

Thermodynamic analyses of solar thermal gasification of coal for hybrid solar-fossil power and fuel production

International Nuclear Information System (INIS)

Ng, Yi Cheng; Lipiński, Wojciech

2012-01-01

Thermodynamic analyses are performed for solar thermal steam and dry gasification of coal. The selected types of coal are anthracite, bituminous, lignite and peat. Two model conversion paths are considered for each combination of the gasifying agent and the coal type: production of the synthesis gas with its subsequent use in a combined cycle power plant to generate power, and production of the synthesis gas with its subsequent use to produce gasoline via the Fischer–Tropsch synthesis. Replacement of a coal-fired 35% efficient Rankine cycle power plant and a combustion-based integrated gasification combined cycle power plant by a solar-based integrated gasification combined cycle power plant leads to the reduction in specific carbon dioxide emissions by at least 47% and 27%, respectively. Replacement of a conventional gasoline production process via coal gasification and a subsequent Fischer–Tropsch synthesis with gasoline production via solar thermal coal gasification with a subsequent Fischer–Tropsch synthesis leads to the reduction in specific carbon dioxide emissions by at least 39%. -- Highlights: ► Thermodynamic analyses for steam and dry gasification of coal are presented. ► Hybrid solar-fossil paths to power and fuels are compared to those using only combustion. ► Hybrid power production can reduce specific CO 2 emissions by more than 27%. ► Hybrid fuel production can reduce specific CO 2 emissions by more than 39%.

Performance Analysis of Depleted Oil Reservoirs for Underground Gas Storage

Directory of Open Access Journals (Sweden)

Dr. C.I.C. Anyadiegwu

2014-02-01

Full Text Available The performance of underground gas storage in depleted oil reservoir was analysed with reservoir Y-19, a depleted oil reservoir in Southern region of the Niger Delta. Information on the geologic and production history of the reservoir were obtained from the available field data of the reservoir. The verification of inventory was done to establish the storage capacity of the reservoir. The plot of the well flowing pressure (Pwf against the flow rate (Q, gives the deliverability of the reservoir at various pressures. Results of the estimated properties signified that reservoir Y-19 is a good candidate due to its storage capacity and its flow rate (Q of 287.61 MMscf/d at a flowing pressure of 3900 psig

Characterization of spent EBR-II driver fuel

International Nuclear Information System (INIS)

McKnight, R. D.

1998-01-01

Operations and material control and accountancy requirements for the Fuel Conditioning Facility demand accurate prediction of the mass flow of spent EBR-II driver fuel into the facility. This requires validated calculational tools that can predict the burnup and isotopic distribution in irradiated Zr-alloy fueled driver assemblies. Detailed core-follow depletion calculations have been performed for an extensive series of EBR-II runs to produce a database of material inventories for the spent fuel to be processed. As this fuel is processed, comparison of calculated values with measured data obtained from samples of this fuel is producing a growing set of validation data. A more extensive set of samples and measurements from the initial processing of irradiated driver fuel has produced valuable estimates of the biases and uncertainties in both the measured and calculated values. Results of these comparisons are presented herein and indicate the calculated values adequately predict the mass flows

Chemical analyses and calculation of isotopic compositions of high-burnup UO{sub 2} fuels and MOX fuels

Energy Technology Data Exchange (ETDEWEB)

Matsumura, Tetsuo; Sasahara, Akihiro [Central Research Inst. of Electric Power Industry, Tokyo (Japan)

2001-08-01

Chemical analysis activities of isotopic compositions of high-burnup UO{sub 2} fuels and MOX fuels in CRIEPI and calculation evaluation are reviewed briefly. C/E values of ORIGEN2, in which original libraries and JENDL-3.2 libraries are used, and other codes with chemical analysis data are reviewed and evaluated. Isotopic compositions of main U and Pu in fuels can be evaluated within 10% relative errors by suitable libraries and codes. Void ratio is effective parameter for C/E values in BWR fuels. JENDL-3.2 library shows remarkable improvement compared with original libraries in isotopic composition evaluations of FP nuclides. (author)

ORIGEN-ARP 2.00, Isotope Generation and Depletion Code System-Matrix Exponential Method with GUI and Graphics Capability

International Nuclear Information System (INIS)

2002-01-01

1 - Description of program or function: ORIGEN-ARP was developed for the Nuclear Regulatory Commission and the Department of Energy to satisfy a need for an easy-to-use standardized method of isotope depletion/decay analysis for spent fuel, fissile material, and radioactive material. It can be used to solve for spent fuel characterization, isotopic inventory, radiation source terms, and decay heat. This release of ORIGEN-ARP is a standalone code package that contains an updated version of the SCALE-4.4a ORIGEN-S code. It contains a subset of the modules, data libraries, and miscellaneous utilities in SCALE-4.4a. This package is intended for users who do not need the entire SCALE package. ORIGEN-ARP 2.00 (2-12-2002) differs from the previous release ORIGEN-ARP 1.0 (July 2001) in the following ways: 1.The neutron source and energy spectrum routines were replaced with computational algorithms and data from the SOURCES-4B code (RSICC package CCC-661) to provide more accurate spontaneous fission and (alpha,n) neutron sources, and a delayed neutron source capability was added. 2.The printout of the fixed energy group structure photon tables was removed. Gamma sources and spectra are now printed for calculations using the Master Photon Library only. 2 - Methods: ORIGEN-ARP is an automated sequence to perform isotopic depletion / decay calculations using the ARP and ORIGEN-S codes of the SCALE system. The sequence includes the OrigenArp for Windows graphical user interface (GUI) that prepares input for ARP (Automated Rapid Processing) and ORIGEN-S. ARP automatically interpolates cross sections for the ORIGEN-S depletion/decay analysis using enrichment, burnup, and, optionally moderator density, from a set of libraries generated with the SCALE SAS2 depletion sequence. Library sets for four LWR fuel assembly designs (BWR 8 x 8, PWR 14 x 14, 15 x 15, 17 x 17) are included. The libraries span enrichments from 1.5 to 5 wt% U-235 and burnups of 0 to 60,000 MWD/MTU. Other

Ego Depletion Impairs Implicit Learning

Science.gov (United States)

Thompson, Kelsey R.; Sanchez, Daniel J.; Wesley, Abigail H.; Reber, Paul J.

2014-01-01

Implicit skill learning occurs incidentally and without conscious awareness of what is learned. However, the rate and effectiveness of learning may still be affected by decreased availability of central processing resources. Dual-task experiments have generally found impairments in implicit learning, however, these studies have also shown that certain characteristics of the secondary task (e.g., timing) can complicate the interpretation of these results. To avoid this problem, the current experiments used a novel method to impose resource constraints prior to engaging in skill learning. Ego depletion theory states that humans possess a limited store of cognitive resources that, when depleted, results in deficits in self-regulation and cognitive control. In a first experiment, we used a standard ego depletion manipulation prior to performance of the Serial Interception Sequence Learning (SISL) task. Depleted participants exhibited poorer test performance than did non-depleted controls, indicating that reducing available executive resources may adversely affect implicit sequence learning, expression of sequence knowledge, or both. In a second experiment, depletion was administered either prior to or after training. Participants who reported higher levels of depletion before or after training again showed less sequence-specific knowledge on the post-training assessment. However, the results did not allow for clear separation of ego depletion effects on learning versus subsequent sequence-specific performance. These results indicate that performance on an implicitly learned sequence can be impaired by a reduction in executive resources, in spite of learning taking place outside of awareness and without conscious intent. PMID:25275517

Ego depletion impairs implicit learning.

Directory of Open Access Journals (Sweden)

Kelsey R Thompson

Full Text Available Implicit skill learning occurs incidentally and without conscious awareness of what is learned. However, the rate and effectiveness of learning may still be affected by decreased availability of central processing resources. Dual-task experiments have generally found impairments in implicit learning, however, these studies have also shown that certain characteristics of the secondary task (e.g., timing can complicate the interpretation of these results. To avoid this problem, the current experiments used a novel method to impose resource constraints prior to engaging in skill learning. Ego depletion theory states that humans possess a limited store of cognitive resources that, when depleted, results in deficits in self-regulation and cognitive control. In a first experiment, we used a standard ego depletion manipulation prior to performance of the Serial Interception Sequence Learning (SISL task. Depleted participants exhibited poorer test performance than did non-depleted controls, indicating that reducing available executive resources may adversely affect implicit sequence learning, expression of sequence knowledge, or both. In a second experiment, depletion was administered either prior to or after training. Participants who reported higher levels of depletion before or after training again showed less sequence-specific knowledge on the post-training assessment. However, the results did not allow for clear separation of ego depletion effects on learning versus subsequent sequence-specific performance. These results indicate that performance on an implicitly learned sequence can be impaired by a reduction in executive resources, in spite of learning taking place outside of awareness and without conscious intent.

Ego depletion impairs implicit learning.

Science.gov (United States)

Thompson, Kelsey R; Sanchez, Daniel J; Wesley, Abigail H; Reber, Paul J

2014-01-01

Implicit skill learning occurs incidentally and without conscious awareness of what is learned. However, the rate and effectiveness of learning may still be affected by decreased availability of central processing resources. Dual-task experiments have generally found impairments in implicit learning, however, these studies have also shown that certain characteristics of the secondary task (e.g., timing) can complicate the interpretation of these results. To avoid this problem, the current experiments used a novel method to impose resource constraints prior to engaging in skill learning. Ego depletion theory states that humans possess a limited store of cognitive resources that, when depleted, results in deficits in self-regulation and cognitive control. In a first experiment, we used a standard ego depletion manipulation prior to performance of the Serial Interception Sequence Learning (SISL) task. Depleted participants exhibited poorer test performance than did non-depleted controls, indicating that reducing available executive resources may adversely affect implicit sequence learning, expression of sequence knowledge, or both. In a second experiment, depletion was administered either prior to or after training. Participants who reported higher levels of depletion before or after training again showed less sequence-specific knowledge on the post-training assessment. However, the results did not allow for clear separation of ego depletion effects on learning versus subsequent sequence-specific performance. These results indicate that performance on an implicitly learned sequence can be impaired by a reduction in executive resources, in spite of learning taking place outside of awareness and without conscious intent.

Exposure to nature counteracts aggression after depletion.

Science.gov (United States)

Wang, Yan; She, Yihan; Colarelli, Stephen M; Fang, Yuan; Meng, Hui; Chen, Qiuju; Zhang, Xin; Zhu, Hongwei

2018-01-01

Acts of self-control are more likely to fail after previous exertion of self-control, known as the ego depletion effect. Research has shown that depleted participants behave more aggressively than non-depleted participants, especially after being provoked. Although exposure to nature (e.g., a walk in the park) has been predicted to replenish resources common to executive functioning and self-control, the extent to which exposure to nature may counteract the depletion effect on aggression has yet to be determined. The present study investigated the effects of exposure to nature on aggression following depletion. Aggression was measured by the intensity of noise blasts participants delivered to an ostensible opponent in a competition reaction-time task. As predicted, an interaction occurred between depletion and environmental manipulations for provoked aggression. Specifically, depleted participants behaved more aggressively in response to provocation than non-depleted participants in the urban condition. However, provoked aggression did not differ between depleted and non-depleted participants in the natural condition. Moreover, within the depletion condition, participants in the natural condition had lower levels of provoked aggression than participants in the urban condition. This study suggests that a brief period of nature exposure may restore self-control and help depleted people regain control over aggressive urges. © 2017 Wiley Periodicals, Inc.

Description and use of NUFCOS-2 for fuel cycle analyses

International Nuclear Information System (INIS)

Westerberg, R.; Vira, J.

1980-12-01

NUFCOS-2 is a continuation to the series of computerized nuclear fuel cycle models which recently have been developed at the Nuclear Engineering Laboratory of the Technical Research Centre of Finland. While the main purpose of the previous NUFCOS model was the multiobjective optimization of the light water reactor fuel cycle, the present version takes a broader view of the global development of the different fuel cycles, especially with regard to the use of plutonium as a fuel in converters or breeders. In this respect an essential feature of NUFCOS-2 is the consideration of the coupling between the consumption and the prices of natural uranium. This report describes the NUFCOS-2 model with main emphasis on the computer application. First the underlying flow equations for the nuclear material are explained in some detail and the structure of the calculation system is represented. The rest of the report, including the appendices, then describes the practical use of the computer model. An example of the input has been provided. (author)

Pu-rich MOX agglomerate-by-agglomerate model for fuel pellet burnup analysis

International Nuclear Information System (INIS)

Chang, G.S.

2004-01-01

In support of potential licensing of the mixed oxide (MOX) fuel made from weapons-grade (WG) plutonium and depleted uranium for use in United States reactors, an experiment containing WG-MOX fuel is being irradiated in the Advanced Test Reactor (ATR) at the Idaho National Engineering and Environmental Laboratory (INEEL). The WG-MOX comprises five percent PuO 2 and 95% depleted UO 2 . Based on the Post Irradiation Examination (PIE) observation, the volume fraction (VF) of MOX agglomerates in the fuel pellet is about 16.67%, and PuO 2 concentration of 30.0 = (5 / 16.67 x 100) wt% in the agglomerate. A pressurized water reactor (PWR) unit WG-MOX lattice with Agglomerate-by-Agglomerate Fuel (AbAF) modeling has been developed. The effect of the irregular agglomerate distribution can be addressed through the use of the Monte Carlo AbAF model. The AbAF-calculated cumulative ratio of Agglomerate burnup to U-MAtrix burnup (AG/MA) is 9.17 at the beginning of life, and decreases to 2.88 at 50 GWd/t. The MCNP-AbAF-calculated results can be used to adjust the parameters in the MOX fuel fission gas release modeling. (author)

Elements of nuclear reactor fueling theory

International Nuclear Information System (INIS)

Egan, M.R.

1984-01-01

Starting with a review of the simple batch size effect, a more general theory of nuclear fueling is derived to describe the behavior and physical requirements of operating cycle sequences and fueling strategies having practical use in the management of nuclear fuel. The generalized theory, based on linear reactivity modeling, is analytical and represents the effects of multiple-stream, multiple-depletion-batch fueling configurations in systems employing arbitrary, non-integer batch size strategies, and containing fuel with variable energy generation rates. Reactor operating cycles and cycle sequences are represented with realistic structure that includes the effects of variable cycle energy production, cycle lengths, end-of-cycle operating extensions and maneuvering allowances. Results of the analytical theory are first applied to the special case of degenerate equilibrium cycle sequences, yielding several fundamental principles related to the selection of refueling strategy, and which govern fueling decisions normally made by the fuel manager. It is also demonstrated in this application that the simple batch size effect is not valid for non-integer fueling strategies, even in the simplest sequence configurations, and that it systematically underestimates the fueling requirements of degenerate sequences in general

Advanced fuels for gas turbines: Fuel system corrosion, hot path deposit formation and emissions

International Nuclear Information System (INIS)

Seljak, Tine; Širok, Brane; Katrašnik, Tomaž

2016-01-01

Highlights: • Technical feasibility analysis of alternative fuels requires a holistic approach. • Fuel, combustion, corrosion and component functionality are strongly related. • Used approach defines design constraints for microturbines using alternative fuels. - Abstract: To further expand the knowledge base on the use of innovative fuels in the micro gas turbines, this paper provides insight into interrelation between specific fuel properties and their impact on combustion and emission formation phenomena in micro gas turbines for stationary power generation as well as their impact on material corrosion and deposit formation. The objective of this study is to identify potential issues that can be related to specific fuel properties and to propose counter measures for achieving stable, durable, efficient and low emission operation of the micro gas turbine while utilizing advanced/innovative fuels. This is done by coupling combustion and emission formation analyses to analyses of material degradation and degradation of component functionality while interpreting them through fuel-specific properties. To ensure sufficiently broad range of fuel properties to demonstrate the applicability of the method, two different fuels with significantly different properties are analysed, i.e. tire pyrolysis oil and liquefied wood. It is shown that extent of required micro gas turbine adaptations strongly correlates with deviations of the fuel properties from those of the baseline fuel. Through the study, these adaptations are supported by in-depth analyses of impacts of fuel properties on different components, parameters and subsystems and their quantification. This holistic approach is further used to propose methodologies and innovative approaches for constraining a design space of micro gas turbine to successfully utilize wide spectra of alternative/innovative fuels.

Criticality safety analyses in SKODA JS a.s

International Nuclear Information System (INIS)

Mikolas, P.; Svarny, J.

1999-01-01

This paper describes criticality safety analyses of spent fuel systems for storage and transport of spent fuel performed in SKODA JS s.r.o.. Analyses were performed for different systems both at NPP site including originally designed spent fuel pool with a large pitch between assemblies without any special absorbing material, high density spent fuel pool with an additional absorption by boron steel, depository rack for fresh fuel assemblies with a very large pitch between fuel assemblies, a container for transport of fresh fuel into the reactor pool and a cask for transport and storage of spent fuel and container for final storage depository. required subcriticality has been proven taking into account all possible unfavourable conditions, uncertainties etc. In two cases, burnup credit methodology is expected to be used. (Authors)

The Impact of Operating Parameters and Correlated Parameters for Extended BWR Burnup Credit

Energy Technology Data Exchange (ETDEWEB)

Ade, Brian J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Marshall, William B. J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Ilas, Germina [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Betzler, Benjamin R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Bowman, Stephen M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2017-06-01

Applicants for certificates of compliance for spent nuclear fuel (SNF) transportation and dry storage systems perform analyses to demonstrate that these systems are adequately subcritical per the requirements of Title 10 of the Code of Federal Regulations (10 CFR) Parts 71 and 72. For pressurized water reactor (PWR) SNF, these analyses may credit the reduction in assembly reactivity caused by depletion of fissile nuclides and buildup of neutron-absorbing nuclides during power operation. This credit for reactivity reduction during depletion is commonly referred to as burnup credit (BUC). US Nuclear Regulatory Commission (NRC) staff review BUC analyses according to the guidance in the Division of Spent Fuel Storage and Transportation Interim Staff Guidance (ISG) 8, Revision 3, Burnup Credit in the Criticality Safety Analyses of PWR Spent Fuel in Transportation and Storage Casks.

Distinction between natural and depleted uranium using instrumental neutron activation analysis

International Nuclear Information System (INIS)

Haddad, Kh.

2008-01-01

A convenient method to discriminate between natural and depleted uranium samples was developed in this work. Traces of natural and depleted uranium were irradiated separately and the ratios of 95 Zr/ 103 Ru, 239 Np/ 95 Zr, 239 Np/ 103 Ru were measured. The results show that these ratios can be used as indicators of the uranium isotopic composition of the sample. These ratios are independent of the secular equilibrium of the 238 U with its daughters in the sample and indicate the isotopic composition for trace amounts. Date and truffle samples has been analysed also using this method. Results show that the uranium content in this product was less than the detection limit.(author)

Using depleted uranium to shield vitrified high-level waste packages

International Nuclear Information System (INIS)

Yoshimura, H.R.; Gildea, P.D.; Bernard, E.A.

1995-01-01

The underlying report for this paper evaluates options for using depleted uranium as shielding materials for transport systems for disposal of vitrified high-level waste (VHLW). In addition, economic analyses are presented to compare costs associated with these options to costs, associated with existing and proposed storage, transport, and diposal capabilities. A more detailed evaluation is provided elsewhere. (Yoshimura et al. 1995.)

Early results of studies on the levels of depleted uranium excreted by Balkan residents

International Nuclear Information System (INIS)

Priest, N.D.; Thirlwell, M.

2002-01-01

Urine samples collected from residents of Bosnia and Herzegovina and Kosovo were analysed to determine their natural and depleted uranium content using MC-ICP-MS. All may have been exposed to depleted uranium released as a consequence of the deployment of armour-piercing rounds by the US Air Force. A 236 U tracer was employed to determine chemical recovery. Early results suggest that the levels of natural and depleted uranium excretion by the subjects, which ranged in age from 1 to 71 years, ranged from 2.8 - 58.2 ng d -1 and 1.3 - 46.3 ng d -1 , respectively. The results suggest accumulated body burdens of depleted uranium ranging from close to zero to 46 μg. All the body burdens predicted are lower than published values for the uranium content of the body (90μg) and health effects are not predicted. Further studies are underway to check the provenance of the results. (author)

Depletion of key protein components of the RISC pathway impairs pre-ribosomal RNA processing.

Science.gov (United States)

Liang, Xue-Hai; Crooke, Stanley T

2011-06-01

Little is known about whether components of the RNA-induced silencing complex (RISC) mediate the biogenesis of RNAs other than miRNA. Here, we show that depletion of key proteins of the RISC pathway by antisense oligonucleotides significantly impairs pre-rRNA processing in human cells. In cells depleted of Drosha or Dicer, different precursors to 5.8S rRNA strongly accumulated, without affecting normal endonucleolytic cleavages. Moderate yet distinct processing defects were also observed in Ago2-depleted cells. Physical links between pre-rRNA and these proteins were identified by co-immunoprecipitation analyses. Interestingly, simultaneous depletion of Dicer and Drosha led to a different processing defect, causing slower production of 28S rRNA and its precursor. Both Dicer and Ago2 were detected in the nuclear fraction, and reduction of Dicer altered the structure of the nucleolus, where pre-rRNA processing occurs. Together, these results suggest that Drosha and Dicer are implicated in rRNA biogenesis.

SSYST, a code-system for analysing transient LWR fuel rod behaviour under off-normal conditions

International Nuclear Information System (INIS)

Borgwaldt, H.; Gulden, W.

1983-01-01

SSYST is a code-system for analysing transient fuel rod behaviour under off-normal conditions, developed conjointly by the Institut fuer Kernenergetik und Energiesysteme (IKE), Stuttgart, and Kernforschungszentrum Karlsruhe (KfK) under contract of Projek Nukleare Sicherheit (PNS) at KfK. The main differences between SSYST and similar codes are (1) an open-ended modular code organisation, and (2) a preference for simple models, wherever possible. While the first feature makes SSYST a very flexible tool, easily adapted to changing requirements, the second feature leads to short execution times. The analysis of transient rod behaviour under LOCA boundary conditions takes 2 min cpu-time (IBM-3033), so that extensive parametric studies become possible. This paper gives an outline of the overall code organisation and a general overview of the physical models implemented. Besides explaining the routine application of SSYST in the analysis of loss-of-coolant accidents, examples are given of special applications which have led to a satisfactory understanding of the decisive influence of deviations from rotational symmetry on the fuel rod perimeter. (author)

Magnetotomography—a new method for analysing fuel cell performance and quality

Science.gov (United States)

Hauer, Karl-Heinz; Potthast, Roland; Wüster, Thorsten; Stolten, Detlef

Magnetotomography is a new method for the measurement and analysis of the current density distribution of fuel cells. The method is based on the measurement of the magnetic flux surrounding the fuel cell stack caused by the current inside the stack. As it is non-invasive, magnetotomography overcomes the shortcomings of traditional methods for the determination of current density in fuel cells [J. Stumper, S.A. Campell, D.P. Wilkinson, M.C. Johnson, M. Davis, In situ methods for the determination of current distributions in PEM fuel cells, Electrochem. Acta 43 (1998) 3773; S.J.C. Cleghorn, C.R. Derouin, M.S. Wilson, S. Gottesfeld, A printed circuit board approach to measuring current distribution in a fuel cell, J. Appl. Electrochem. 28 (1998) 663; Ch. Wieser, A. Helmbold, E. Gülzow, A new technique for two-dimensional current distribution measurements in electro-chemical cells, J. Appl. Electrochem. 30 (2000) 803; Grinzinger, Methoden zur Ortsaufgelösten Strommessung in Polymer Elektrolyt Brennstoffzellen, Diploma thesis, TU-München, 2003; Y.-G. Yoon, W.-Y. Lee, T.-H. Yang, G.-G. Park, C.-S. Kim, Current distribution in a single cell of PEMFC, J. Power Sources 118 (2003) 193-199; M.M. Mench, C.Y. Wang, An in situ method for determination of current distribution in PEM fuel cells applied to a direct methanol fuel cell, J. Electrochem. Soc. 150 (2003) A79-A85; S. Schönbauer, T. Kaz, H. Sander, E. Gülzow, Segmented bipolar plate for the determination of current distribution in polymer electrolyte fuel cells, in: Proceedings of the Second European PEMFC Forum, vol. 1, Lucerne/Switzerland, 2003, pp. 231-237; G. Bender, S.W. Mahlon, T.A. Zawodzinski, Further refinements in the segmented cell approach to diagnosing performance in polymer electrolyte fuel cells, J. Power Sources 123 (2003) 163-171]. After several years of research a complete prototype system is now available for research on single cells and stacks. This paper describes the basic system (fundamentals

The interaction between Otto fuel II and aqueous hydroxylammonium perchlorate (HAP). Pt. 3: depletion of components within the reacting liquids

Energy Technology Data Exchange (ETDEWEB)

Bellerby, John M.; Blackman, Christopher S. [Department of Environmental and Ordnance Systems, Cranfield University, Defence College of Management and Technology, Shrivenham, Swindon SN6 8LA (United Kingdom)

2007-06-15

Gas chromatography (GC) with a Flame Ionisation Detector (FID) has been used to determine changes in the concentrations of the components of Otto Fuel II (OF) in contact with an 82% aqueous solution of hydroxylammonium perchlorate (HAP) in sealed vials at 31.7 C during the period leading up to auto-ignition of the two liquids. The concentration of hydroxylamine in HAP was monitored over the same period using a titration method. It was found that 2-nitrodiphenylamine (2NDPA), the stabiliser in the OF, is completely consumed after about 65-70 h and that the concentration of hydroxylamine begins to fall at this point. 1,2-Propanediol dinitrate (propylene glycol dinitrate, PGDN), the energetic component in the OF, is not depleted significantly until after about 90 h. The evolution of nitrous oxide (N{sub 2}O) between 65 and 90 h is attributed to the reaction of the hydroxylammonium ion with nitrous acids produced by PGDN decomposition at the liquid-liquid interface. Carbon dioxide (CO{sub 2}) is evolved after 90 h and is attributed to PGDN decomposition. HAP and PGDN are each thought to contribute to N{sub 2}O evolution after 90 h. (Abstract Copyright [2007], Wiley Periodicals, Inc.)

Toxicity of Depleted Uranium Dust Particles: Results of a New Model

International Nuclear Information System (INIS)

Zucchetti, M.

2013-01-01

Depleted uranium (DU) is mostly composed of U-238, a naturally radioactive isotope. Concerning chemical toxicity, uranium, being a heavy metal, is known to have toxic effects on specific organs in the body, the kidneys in particular. Its effects are similar to those of other heavy metals, such as lead and cadmium. Scientific evidence resulting both from in vitro and in vivo analyses shows that current models of the mechanisms of toxicity of uranium dust are not fully satisfactory. They should be refined in order to obtain more effective responses and predictions regarding health effects. In particular, radiotoxicity potential of Depleted Uranium dust originated by military use of this material for ammunition must be re-evaluated taking into account the bystander effect, the dose enhancing effect and other minor phenomena. Uranium dust has both chemical and radiological toxicity: the synergistic aspect of the two effects has to be accounted for, in order to arrive to a complete description of the phenomenon. The combination of the two different toxicities (chemical and radiological) of depleted uranium is attempted here for the first time, approaching the long-term effects of Depleted Uranium, and in particular the carcinogenetic effects. A case study (Balkan war, 1999) is discussed. (Author)

Neutron multiplication and shielding problems in pressurized water reactor spent fuel shipping casks

International Nuclear Information System (INIS)

Devillers, C.; Blum, P.

1977-01-01

To evaluate the degree of accuracy of computational methods used in the shield design of spent fuel shipping casks, comparisons have been made between biological dose-rate calculations and measurements at the surface of a cask carrying three pressurized water reactor fuel assemblies. Neutron dose-rate measurements made with the fuel-carrying region successively wet and dry are also used to derive an experimental value of the k/sub eff/ of the wet fuel assemblies. Results obtained by this method are shown to be consistent with criticality calculations, taking into account fuel depletion

Potential For Stratospheric Ozone Depletion During Carboniferous

Science.gov (United States)

Bill, M.; Goldstein, A. H.

Methyl bromide (CH3Br) constitutes the largest source of bromine atoms to the strato- sphere whereas methyl chloride (CH3Cl) is the most abundant halocarbon in the tro- posphere. Both gases play an important role in stratospheric ozone depletion. For in- stance, Br coupled reactions are responsible for 30 to 50 % of total ozone loss in the polar vortex. Currently, the largest natural sources of CH3Br and CH3Cl appear to be biological production in the oceans, inorganic production during biomass burning and plant production in salt marsh ecosystems. Variations of paleofluxes of CH3Br and CH3Cl can be estimated by analyses of oceanic paleoproductivity, stratigraphic analyses of frequency and distribution of fossil charcoal indicating the occurrence of wildfires, and/or by paleoreconstruction indicating the extent of salt marshes. Dur- ing the lower Carboniferous time (Tournaisian-Visean), the southern margin of the Laurasian continent was characterized by charcoal deposits. Estimation on frequency of charcoal layers indicates that wildfires occur in a range of 3-35 years (Falcon-Lang 2000). This suggests that biomass burning could be an important source of CH3Br and CH3Cl during Tournaisian-Viesan time. During Tounaisian and until Merame- cian carbon and oxygen isotope records have short term oscillations (Bruckschen et al. 1999, Mii et al. 1999). Chesterian time (mid- Carboniferous) is marked by an in- crease in delta18O values ( ~ 2 permil) and an increase of glacial deposit frequency suggesting lower temperatures. The occurrence of glacial deposits over the paleopole suggests polar conditions and the associated special features of polar mete- orology such as strong circumpolar wind in the stratosphere (polar vortex) and polar stratospheric clouds. Thus, conditions leading to polar statospheric ozone depletion can be found. Simultaneously an increase in delta13C values is documented. We interpret the positive shift in delta13C as a result of higher bioproductivity

Comparison of depletion results for a boiling water reactor fuel element with CASMO and SCALE 6.1 (TRITON/NEWT)

International Nuclear Information System (INIS)

Mesado, C.; Morera, D.; Miro, R.; Barrachina, T.; Verdu, G.; Soler, Amparo; Melara, Jose

2013-01-01

In this work, the results of depletion calculations with CASMO and SCALE 6.1 (TRITON) are compared. To achieve it, a region of a Boiling Water Reactor (BWR) fuel element is modeled, using both codes. To take into account different operating conditions, the simulations are repeated with different void fraction, ranging from null void fraction to a void fraction closed to one. Special care was used to keep in mind that the homogenization of the materials and the two group approach was the same in both codes. Additionally, KENO-VI and MCDANCOFF modules are used. The k-effective calculated by KENO-VI is used to ensure that the starting point was correct and MCDANCOFF module is used to calculate the Dancoff factors in order to improve the model accuracy. To validate the whole process, a comparison of k eff , and cross-sections collapsed and homogenized is shown. The results show a very good agreement, with an average error around the 1.75%. Furthermore, an automatic process for translating CASMO data to SCALE input decks was developed. The reason for the translation is the fact that SCALE's TRITON module is a new code very powerful and continuously being developed. Thus, great advantage can be taken from the current and future SCALE features. This is hoped to produce more realistic models, and hence, increase the accuracy of neutronic libraries. (author)

Kinetics of depletion interactions

NARCIS (Netherlands)

Vliegenthart, G.A.; Schoot, van der P.P.A.M.

2003-01-01

Depletion interactions between colloidal particles dispersed in a fluid medium are effective interactions induced by the presence of other types of colloid. They are not instantaneous but built up in time. We show by means of Brownian dynamics simulations that the static (mean-field) depletion force

Numerical Investigation into CO Emission, O Depletion, and Thermal Decomposition in a Reacting Slab

Directory of Open Access Journals (Sweden)

O. D. Makinde

2011-01-01

Full Text Available The emission of carbon dioxide (CO2 is closely associated with oxygen (O2 depletion, and thermal decomposition in a reacting stockpile of combustible materials like fossil fuels (e.g., coal, oil, and natural gas. Moreover, it is understood that proper assessment of the emission levels provides a crucial reference point for other assessment tools like climate change indicators and mitigation strategies. In this paper, a nonlinear mathematical model for estimating the CO2 emission, O2 depletion, and thermal stability of a reacting slab is presented and tackled numerically using a semi-implicit finite-difference scheme. It is assumed that the slab surface is subjected to a symmetrical convective heat and mass exchange with the ambient. Both numerical and graphical results are presented and discussed quantitatively with respect to various parameters embedded in the problem.

ORIGEN2: a revised and updated version of the Oak Ridge isotope generation and depletion code

International Nuclear Information System (INIS)

Croff, A.G.

1980-07-01

ORIGEN2 is a versatile point depletion and decay computer code for use in simulating nuclear fuel cycles and calculating the nuclide compositions of materials contained therein. This code represents a revision and update of the original ORIGEN computer code which has been distributed world-wide beginning in the early 1970s. The purpose of this report is to give a summary description of a revised and updated version of the original ORIGEN computer code, which has been designated ORIGEN2. A detailed description of the computer code ORIGEN2 is presented. The methods used by ORIGEN2 to solve the nuclear depletion and decay equations are included. Input information necessary to use ORIGEN2 that has not been documented in supporting reports is documented

Tylosin depletion from edible pig tissues.

Science.gov (United States)

Prats, C; El Korchi, G; Francesch, R; Arboix, M; Pérez, B

2002-12-01

The depletion of tylosin from edible pig tissues was studied following 5 days of intramuscular (i.m.) administration of 10 mg/kg of tylosin to 16 crossbreed pigs. Animals were slaughtered at intervals after treatment and samples of muscle, kidney, liver, skin+fat, and injection site were collected and analysed by high-performance liquid chromatography (HPLC). Seven days after the completion of treatment, the concentration of tylosin in kidney, skin+fat, and at the injection site was higher than the European Union maximal residue limit (MRL) of 100 microg/kg. Tylosin residues in all tissues were below the quantification limit (50 microg/kg) at 10 and 14 days post-treatment.

Metallography of pitted aluminum-clad, depleted uranium fuel

International Nuclear Information System (INIS)

Nelson, D.Z.; Howell, J.P.

1994-01-01

The storage of aluminum-clad fuel and target materials in the L-Disassembly Basin at the Savannah River Site for more than 5 years has resulted in extensive pitting corrosion of these materials. In many cases the pitting corrosion of the aluminum clad has penetrated in the uranium metal core, resulting in the release of plutonium, uranium, cesium-137, and other fission product activity to the basin water. In an effort to characterize the extent of corrosion of the Mark 31A target slugs, two unirradiated slug assemblies were removed from basin storage and sent to the Savannah River Technology Center for evaluation. This paper presents the results of the metallography and photographic documentation of this evaluation. The metallography confirmed that pitting depths varied, with the deepest pit found to be about 0.12 inches (3.05 nun). Less than 2% of the aluminum cladding was found to be breached resulting in less than 5% of the uranium surface area being affected by corrosion. The overall integrity of the target slug remained intact

A novel concept of QUADRISO particles Part III: applications to the plutonium-thorium fuel cycle

International Nuclear Information System (INIS)

Talamo, A.

2009-01-01

In the present study, a plutonium-thorium fuel cycle is investigated including the 233 U production and utilization. A prismatic thermal High Temperature Gas Reactor (HTGR) and the novel concept of quadruple isotropic (QUADRISO) coated particles, designed at the Argonne National Laboratory, have been used for the study. In absorbing QUADRISO particles, a burnable poison layer surrounds the central fuel kernel to flatten the reactivity curve as a function of time. At the beginning of life, the fuel in the QUADRISO particles is hidden from neutrons, since they get absorbed in the burnable poison before they reach the fuel kernel. Only when the burnable poison depletes, neutrons start streaming into the fuel kernel inducing fission reactions and compensating the fuel depletion of ordinary TRISO particles. In fertile QUADRISO particles, the absorber layer is replaced by natural thorium with the purpose of flattening the excess of reactivity by the thorium resonances and producing 233 U. The above configuration has been compared with a configuration where fissile (neptunium-plutonium oxide from Light Water Reactors irradiated fuel) and fertile (natural thorium oxide) fuels are homogeneously mixed in the kernel of ordinary TRISO particles. For the 233 U utilization, the core has been equipped with europium oxide absorbing QUADRISO particles.

Fully Depleted Charge-Coupled Devices

International Nuclear Information System (INIS)

Holland, Stephen E.

2006-01-01

We have developed fully depleted, back-illuminated CCDs that build upon earlier research and development efforts directed towards technology development of silicon-strip detectors used in high-energy-physics experiments. The CCDs are fabricated on the same type of high-resistivity, float-zone-refined silicon that is used for strip detectors. The use of high-resistivity substrates allows for thick depletion regions, on the order of 200-300 um, with corresponding high detection efficiency for near-infrared and soft x-ray photons. We compare the fully depleted CCD to the p-i-n diode upon which it is based, and describe the use of fully depleted CCDs in astronomical and x-ray imaging applications

Deuterium - depleted water. Achievements and perspectives

International Nuclear Information System (INIS)

Titescu, Gh.; Stefanescu, I.; Saros-Rogobete, I.

2001-01-01

Deuterium - depleted water represents water that has an isotopic content lower than 145 ppm D/(D+H) which is the natural isotopic content of water. The research conducted at ICSI Ramnicu Valcea, regarding deuterium - depleted water were completed by the following patents: - technique and installation for deuterium - depleted water production; - distilled water with low deuterium content; - technique and installation for the production of distilled water with low deuterium content; - mineralized water with low deuterium content and technique to produce it. The gold and silver medals won at international salons for inventions confirmed the novelty of these inventions. Knowing that deuterium content of water has a big influence on living organisms, beginning with 1996, the ICSI Ramnicu Valcea, deuterium - depleted water producer, co-operated with Romanian specialized institutes for biological effects' evaluation of deuterium - depleted water. The role of natural deuterium in living organisms was examined by using deuterium - depleted water instead of natural water. These investigations led to the following conclusions: 1. deuterium - depleted water caused a tendency towards the increase of the basal tone, accompanied by the intensification of the vasoconstrictor effects of phenylefrine, noradrenaline and angiotensin; the increase of the basal tone and vascular reactivity produced by the deuterium - depleted water persists after the removal of the vascular endothelium; -2. animals treated with deuterium - depleted water showed an increase of the resistance both to sublethal and to lethal gamma radiation doses, suggesting a radioprotective action by the stimulation of non-specific immune defence mechanism; 3, deuterium - depleted water stimulates immune defence reactions, represented by the opsonic, bactericidal and phagocyte capacity of the immune system, together with increase in the numbers of polymorphonuclear neutrophils; 4. investigations regarding artificial

Simulations and observations of plasma depletion, ion composition, and airglow emissions in two auroral ionospheric depletion experiments

International Nuclear Information System (INIS)

Yau, A.W.; Whalen, B.A.; Harris, F.R.; Gattinger, R.L.; Pongratz, M.B.; Bernhardt, P.A.

1985-01-01

In an ionospheric depletion experiment where chemically reactive vapors such as H 2 O and CO 2 are injected into the O + dominant F region to accelerate the plasma recombination rate and to reduce the plasma density, the ion composition in the depleted region is modified, and photometric emissions are produced. We compare in situ ion composition, density, and photometric measurements from two ionospheric depletion experiments with predictions from chemical modeling. The two injections, Waterhole I and III, were part of an auroral perturbation experiment and occurred in different ambient conditions. In both injections a core region of greater than fivefold plasma depletion was observed over roughly-equal5-km diameter within seconds of the injection, surrounded by an outer region of less drastic and slower depletion. In Waterhole I the plasma density was depleted tenfold over a 30-km diamter region after 2 min. The ambient O + density was drastically reduced, and the molecular O + 2 abundance was enhanced fivehold in the depletion region. OH airglow emission associated with the depletion was observed with a peak emission intensity of roughly-equal1 kR. In Waterhole III the ambient density was a decade lower, and the plasma depletion was less drastic, being twofold over 30 km after 2 min. The airglow emissions were also much less intense and below measurement sensitivity (30 R for the OH 306.4-nm emission; 50 R for the 630.0-nm emission)

Effective fiber hypertrophy in satellite cell-depleted skeletal muscle

Science.gov (United States)

McCarthy, John J.; Mula, Jyothi; Miyazaki, Mitsunori; Erfani, Rod; Garrison, Kelcye; Farooqui, Amreen B.; Srikuea, Ratchakrit; Lawson, Benjamin A.; Grimes, Barry; Keller, Charles; Van Zant, Gary; Campbell, Kenneth S.; Esser, Karyn A.; Dupont-Versteegden, Esther E.; Peterson, Charlotte A.

2011-01-01

An important unresolved question in skeletal muscle plasticity is whether satellite cells are necessary for muscle fiber hypertrophy. To address this issue, a novel mouse strain (Pax7-DTA) was created which enabled the conditional ablation of >90% of satellite cells in mature skeletal muscle following tamoxifen administration. To test the hypothesis that satellite cells are necessary for skeletal muscle hypertrophy, the plantaris muscle of adult Pax7-DTA mice was subjected to mechanical overload by surgical removal of the synergist muscle. Following two weeks of overload, satellite cell-depleted muscle showed the same increases in muscle mass (approximately twofold) and fiber cross-sectional area with hypertrophy as observed in the vehicle-treated group. The typical increase in myonuclei with hypertrophy was absent in satellite cell-depleted fibers, resulting in expansion of the myonuclear domain. Consistent with lack of nuclear addition to enlarged fibers, long-term BrdU labeling showed a significant reduction in the number of BrdU-positive myonuclei in satellite cell-depleted muscle compared with vehicle-treated muscle. Single fiber functional analyses showed no difference in specific force, Ca2+ sensitivity, rate of cross-bridge cycling and cooperativity between hypertrophied fibers from vehicle and tamoxifen-treated groups. Although a small component of the hypertrophic response, both fiber hyperplasia and regeneration were significantly blunted following satellite cell depletion, indicating a distinct requirement for satellite cells during these processes. These results provide convincing evidence that skeletal muscle fibers are capable of mounting a robust hypertrophic response to mechanical overload that is not dependent on satellite cells. PMID:21828094

Safety analysis of MOX fuels by fuel performance code

Energy Technology Data Exchange (ETDEWEB)

Suzuki, Motoe [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

2002-12-01

Performance of plutonium rick mixed oxide fuels specified for the Reduced-Moderation Water Reactor (RMWR) has been analysed by modified fuel performance code. Thermodynamic properties of these fuels up to 120 GWd/t burnup have not been measured and estimated using existing uranium fuel models. Fission product release, pressure rise inside fuel rods and mechanical loads of fuel cans due to internal pressure have been preliminarily assessed based on assumed axial power distribution history, which show the integrity of fuel performance. Detailed evaluation of fuel-cladding interactions due to thermal expansion or swelling of fuel pellets due to high burnup will be required for safety analysis of mixed oxide fuels. Thermal conductivity and swelling of plutonium rich mixed oxide fuels shall be taken into consideration. (T. Tanaka)

First steps towards modelling high burnup effect in UO{sub 2} fuel

Energy Technology Data Exchange (ETDEWEB)

O` Carroll, C; Lassmann, K; Laar, J Van De; Walker, C T [CEC Joint Research Centre, Karlsruhe (Germany)

1997-08-01

High burnup initiates a process that can lead to major microstructural changes near the edge of the fuel: formation of subgrains, the loss of matrix fission gas and an increase in porosity. A consequence of this, is a decrease of thermal conductivity near the edge of the fuel which may be major implications for the performance of LWR fuels at higher burnup. The mechanism for the changes in grain structure, the apparent depletion of Xe and increase in porosity is associated with the high fission density at the fuel periphery. This is in turn due to the preferential capture of epithermal neutrons in the resonances of {sup 238}U. The new model TUBRNP predicts the radial burnup profile as a function of time together with the radial profile of plutonium. The model has been validated with data from LWR UO{sub 2} fuels with enrichments in the range 2 to 8.25% and burnups between 21 to 75 Gwd/t. It has been reported that at high burnup EPMA measures a sharp decrease in the concentration of Xe near the fuel surface. This loss of Xe is interpreted as a signal that the gas has been swept out of the original grains into pores: this ``missing`` Xe has been measured by XRF. It has been noted experimentally that the restructuring (Xe depletion and changes in grain structure) have an onset threshold local burnup in the region of 70 to 80 GWd/t: a specific value was taken for use in the model. For a given fuel TUBRNP predicts the local burnup profile, and the depth corresponding to the threshold value is taken to be the thickness of the Xe depleted region. The theoretical predictions have been compared with experimental data. The results are presented and should be seen as a first step in the development of a more detailed model of this phenomenon. (author). 22 refs, 9 figs, 2 tabs.

ANL calculational methodologies for determining spent nuclear fuel source term

International Nuclear Information System (INIS)

McKnight, R. D.

2000-01-01

Over the last decade Argonne National Laboratory has developed reactor depletion methods and models to determine radionuclide inventories of irradiated EBR-II fuels. Predicted masses based on these calculational methodologies have been validated using available data from destructive measurements--first from measurements of lead EBR-II experimental test assemblies and later using data obtained from processing irradiated EBR-II fuel assemblies in the Fuel Conditioning Facility. Details of these generic methodologies are described herein. Validation results demonstrate these methods meet the FCF operations and material control and accountancy requirements

Fusion as a source of synthetic fuels

International Nuclear Information System (INIS)

Powell, J.R.; Fillo, J.A.; Steinberg, M.

1981-01-01

In the near-term, coal derived synthetic fuels will be used; but in the long-term, resource depletion and environmental effects will mandate synthetic fuels from inexhaustible sources - fission, fusion, and solar. Of the three sources, fusion appears uniquely suited for the efficient production of hydrogen-based fuels, due to its ability to directly generate very high process temperatures (up to approx. 2000 0 C) for water splitting reactions. Fusion-based water splitting reactions include high temperature electrolysis (HTE) of steam, thermochemical cycles, hybrid electrochemical/thermochemical, and direct thermal decomposition. HTE appears to be the simplest and most efficient process with efficiencies of 50 to 70% (fusion to hydrogen chemical energy), depending on process conditions

Radiation physics and shielding codes and analyses applied to design-assist and safety analyses of CANDUR and ACRTM reactors

International Nuclear Information System (INIS)

Aydogdu, K.; Boss, C. R.

2006-01-01

This paper discusses the radiation physics and shielding codes and analyses applied in the design of CANDU and ACR reactors. The focus is on the types of analyses undertaken rather than the inputs supplied to the engineering disciplines. Nevertheless, the discussion does show how these analyses contribute to the engineering design. Analyses in radiation physics and shielding can be categorized as either design-assist or safety and licensing (accident) analyses. Many of the analyses undertaken are designated 'design-assist' where the analyses are used to generate recommendations that directly influence plant design. These recommendations are directed at mitigating or reducing the radiation hazard of the nuclear power plant with engineered systems and components. Thus the analyses serve a primary safety function by ensuring the plant can be operated with acceptable radiation hazards to the workers and public. In addition to this role of design assist, radiation physics and shielding codes are also deployed in safety and licensing assessments of the consequences of radioactive releases of gaseous and liquid effluents during normal operation and gaseous effluents following accidents. In the latter category, the final consequences of accident sequences, expressed in terms of radiation dose to members of the public, and inputs to accident analysis, e.g., decay heat in fuel following a loss-of-coolant accident, are also calculated. Another role of the analyses is to demonstrate that the design of the plant satisfies the principle of ALARA (as low as reasonably achievable) radiation doses. This principle is applied throughout the design process to minimize worker and public doses. The principle of ALARA is an inherent part of all design-assist recommendations and safety and licensing assessments. The main focus of an ALARA exercise at the design stage is to minimize the radiation hazards at the source. This exploits material selection and impurity specifications and relies

Too Depleted to Try? Testing the Process Model of Ego Depletion in the Context of Unhealthy Snack Consumption.

Science.gov (United States)

Haynes, Ashleigh; Kemps, Eva; Moffitt, Robyn

2016-11-01

The process model proposes that the ego depletion effect is due to (a) an increase in motivation toward indulgence, and (b) a decrease in motivation to control behaviour following an initial act of self-control. In contrast, the reflective-impulsive model predicts that ego depletion results in behaviour that is more consistent with desires, and less consistent with motivations, rather than influencing the strength of desires and motivations. The current study sought to test these alternative accounts of the relationships between ego depletion, motivation, desire, and self-control. One hundred and fifty-six undergraduate women were randomised to complete a depleting e-crossing task or a non-depleting task, followed by a lab-based measure of snack intake, and self-report measures of motivation and desire strength. In partial support of the process model, ego depletion was related to higher intake, but only indirectly via the influence of lowered motivation. Motivation was more strongly predictive of intake for those in the non-depletion condition, providing partial support for the reflective-impulsive model. Ego depletion did not affect desire, nor did depletion moderate the effect of desire on intake, indicating that desire may be an appropriate target for reducing unhealthy behaviour across situations where self-control resources vary. © 2016 The International Association of Applied Psychology.

The Toxicity of Depleted Uranium

OpenAIRE

Briner, Wayne

2010-01-01

Depleted uranium (DU) is an emerging environmental pollutant that is introduced into the environment primarily by military activity. While depleted uranium is less radioactive than natural uranium, it still retains all the chemical toxicity associated with the original element. In large doses the kidney is the target organ for the acute chemical toxicity of this metal, producing potentially lethal tubular necrosis. In contrast, chronic low dose exposure to depleted uranium may not produce a c...

Preliminary nuclear design for test MOX Fuel rods

Energy Technology Data Exchange (ETDEWEB)

Joo, Hyung Kook; Kim, Taek Kyum; Jeong, Hyung Guk; Noh, Jae Man; Cho, Jin Young; Kim, Young Il; Kim, Young Jin; Sohn, Dong Seong

1997-10-01

As a part of activity for future fuel development project, test MOX fuel rods are going to be loaded and irradiated in Halden reactor core as a KAERI`s joint international program with Paul Scherrer Institute (PSI). PSI will fabricate test MOX rods with attrition mill device which was developed by KAERI. The test fuel assembly rig contains three MOX rods and three inert matrix rods. One of three MOX rods will be fabricated by BNFL, the other two MOX fuel rods will be manufacturing jointly by KAERI and PSI. Three inert matrix fuel rods will be fabricated with Zr-Y-Er-Pu oxide. Neutronic evaluation was preliminarily performed for test fuel assembly suggested by PSI. The power distribution of test fuel rod in test fuel assembly was analyzed for various fuel rods position in assembly and the depletion characteristic curve for test fuel was also determined. The fuel rods position in test fuel assembly does not effect the rod power distribution, and the proposal for test fuel rods suggested by PSI is proved to be feasible. (author). 2 refs., 13 tabs., 16 figs.

LIFE Materials: Overview of Fuels and Structural Materials Issues Volume 1

Energy Technology Data Exchange (ETDEWEB)

Farmer, J

2008-09-08

The National Ignition Facility (NIF) project, a laser-based Inertial Confinement Fusion (ICF) experiment designed to achieve thermonuclear fusion ignition and burn in the laboratory, is under construction at the Lawrence Livermore National Laboratory (LLNL) and will be completed in April of 2009. Experiments designed to accomplish the NIF's goal will commence in late FY2010 utilizing laser energies of 1 to 1.3 MJ. Fusion yields of the order of 10 to 20 MJ are expected soon thereafter. Laser initiated fusion-fission (LIFE) engines have now been designed to produce nuclear power from natural or depleted uranium without isotopic enrichment, and from spent nuclear fuel from light water reactors without chemical separation into weapons-attractive actinide streams. A point-source of high-energy neutrons produced by laser-generated, thermonuclear fusion within a target is used to achieve ultra-deep burn-up of the fertile or fissile fuel in a sub-critical fission blanket. Fertile fuels including depleted uranium (DU), natural uranium (NatU), spent nuclear fuel (SNF), and thorium (Th) can be used. Fissile fuels such as low-enrichment uranium (LEU), excess weapons plutonium (WG-Pu), and excess highly-enriched uranium (HEU) may be used as well. Based upon preliminary analyses, it is believed that LIFE could help meet worldwide electricity needs in a safe and sustainable manner, while drastically shrinking the nation's and world's stockpile of spent nuclear fuel and excess weapons materials. LIFE takes advantage of the significant advances in laser-based inertial confinement fusion that are taking place at the NIF at LLNL where it is expected that thermonuclear ignition will be achieved in the 2010-2011 timeframe. Starting from as little as 300 to 500 MW of fusion power, a single LIFE engine will be able to generate 2000 to 3000 MWt in steady state for periods of years to decades, depending on the nuclear fuel and engine configuration. Because the fission

Core design options for high conversion BWRs operating in Th–233U fuel cycle

International Nuclear Information System (INIS)

Shaposhnik, Y.; Shwageraus, E.; Elias, E.

2013-01-01

Highlights: • BWR core operating in a closed self-sustainable Th– 233 U fuel cycle. • Seed blanket optimization that includes assembly size array and axial dimensions. • Fully coupled MC with fuel depletion and thermo-hydraulic feedback modules. • Thermal-hydraulic analysis includes MCPR observation. -- Abstract: Several options of fuel assembly design are investigated for a BWR core operating in a closed self-sustainable Th– 233 U fuel cycle. The designs rely on an axially heterogeneous fuel assembly structure consisting of a single axial fissile zone “sandwiched” between two fertile blanket zones, in order to improve fertile to fissile conversion ratio. The main objective of the study was to identify the most promising assembly design parameters, dimensions of fissile and fertile zones, for achieving net breeding of 233 U. The design challenge, in this respect, is that the fuel breeding potential is at odds with axial power peaking and the core minimum critical power ratio (CPR), hence limiting the maximum achievable core power rating. Calculations were performed with the BGCore system, which consists of the MCNP code coupled with fuel depletion and thermo-hydraulic feedback modules. A single 3-dimensional fuel assembly having reflective radial boundaries was modeled applying simplified restrictions on the maximum centerline fuel temperature and the CPR. It was found that axially heterogeneous fuel assembly design with a single fissile zone can potentially achieve net breeding, while matching conventional BWR core power rating under certain restrictions to the core loading pattern design

Experience in producing LEU fuel elements for the RSG-GAS

International Nuclear Information System (INIS)

Suripto, A.; Soentono, S.

1991-01-01

To achieve a self-reliance in the operation of the 30 MW Multipurpose Research Reactor at Serpong (the RSG-GAS), a fuel element production facility has been constructed nearby. The main task of the facility is to produce MTR type fuel and control elements containing U 3 O 8 -Al dispersion LEU fuel for the RSG-GAS. The hot commissioning activity has started in early 1988 after completion of the cold commissioning using depleted uranium in 1987, marking the beginning of the real production activity. This paper briefly describes the main features of the fuel production facility, the production experience gained so far, and its current production activity. (orig.)

Deuterium-depleted water. Romanian achievements and perspective

International Nuclear Information System (INIS)

Stefanescu, Ion; Saros-Rogobete, Irina; Titescu, Gheorghe

2001-01-01

Deuterium-depleted water has an isotopic content smaller than 145 ppm D/(D+H) which is the natural isotopic content of water. Beginning with 1996 ICSI Rm. Valcea, deuterium-depleted water producer, co-operated with Romanian specialized institutes for biological effect's evaluation of deuterium-depleted water. These investigations lead to the following conclusions: - Deuterium-depleted water caused a tendency towards the increase of the basal tonus, accompanied by the intensification of the vasoconstrictor effects of phenylefrine, noradrenaline and angiotensin; the increase of the basal tonus and vascular reactivity produced by the deuterium-depleted water persist after the removal of the vascular endothelium; - Animals treated with deuterium-depleted water showed an increase of the resistance both to sublethal and to lethal gamma radiation doses, suggesting a radioprotective action; - Deuterium-depleted water stimulates immune defence reactions and increases the numbers of polymorphonuclear neutrophils; - Investigations regarding artificial reproduction of fish with deuterium-depleted water fecundated solutions confirmed favourable influence in embryo growth stage and resistance in subsequent growth stages; - It was studied germination, growth and quantitative character's variability in plants; one can remark the favourable influence of deuterium-depleted water on biological process in plants in various ontogenetic stages; - The deuterium depletion in seawater produces the diminution of the water spectral energy related to an increased metabolism of Tetraselmis Suecica. (authors)

On problems related to the deployment of depleted uranium weapons in the Balkans

International Nuclear Information System (INIS)

Mietelski, J. W.; Waligorski, M.P.R.; Zunic, Z.S.

2002-01-01

The likely long-term environmental and health effects of the deployment of weapons containing depleted uranium (DU) in the Balkans are discussed. To determine whether depleted uranium or spent reactor fuel was used in the weapons, knowledge is required of the 235 U to 238 U activity (or concentration) ratio in the measured samples. To this end, and to distinguish between uranium originating from natural and man-made sources, we discuss some of the methodology and metrology issues involved in performing alpha- and gamma- spectrometry of uranium in environmental and human samples. We present results of nuclear spectrometry performed on DU core deposits from the aluminium jacket of a PGU-14 bullet found in South Serbia. We draw attention to aspects involving ionising radiation, which are likely to be of importance when formulating a prognosis of the possible environmental and health impact of the deployment of DU weapons, indicating the importance of the inhalation pathway in children. (author)

The nuclear fuel cycle, an overview

International Nuclear Information System (INIS)

Ballery, J.L.; Cazalet, J.; Hagemann, R.

1995-01-01

Because uranium is widely distributed on the face of the Earth, nuclear energy has a very large potential as an energy source in view of future depletion of fossil fuel reserves. Also future energy requirements will be very sizeable as populations of developing countries are often growing and make the energy question one of the major challenges for the coming decades. Today, nuclear contributes some 340 GWe to the energy requirements of the world. Present and future nuclear programs require an adequate fuel cycle industry, from mining, refining, conversion, enrichment, fuel fabrication, fuel reprocessing and the storage of the resulting wastes. The commercial fuel cycle activities amount to an annual business in the 7-8 billions of US Dollars in the hands of a large number of industrial operators. This paper gives details about companies and countries involved in each step of the fuel cycle and about the national strategies and options chosen regarding the back end of the fuel cycle (waste storage and reprocessing). These options are illustrated by considering the policy adopted in three countries (France, United Kingdom, Japan) versed in reprocessing. (J.S.). 13 figs., 2 tabs

Analyses of expected rod performance during the dry storage of spent fuel

International Nuclear Information System (INIS)

Einziger, R.E.

1982-08-01

Within the next ten years, a number of utilities will be forced to increase their interim spent-fuel-storage capability or face the loss of full-core reserve. Dry storage is being considered to fill this need. This paper analyzes the fuel-rod-performance data supporting dry storage and discusses areas where there are still outstanding questions. Three storage temperature ranges (T 0 C, 250 0 C 0 C and T > 400 0 C), two atmospheres (inert, unlimited air) and two initial fuel-rod conditions (intact, breached) are considered. It is concluded that a fuel-performance data base exists that indicates that storage below 250 0 C can be accomplished with long-term fuel pellet and cladding stability. At higher temperatures, analytic studies and laboratory experiments are needed especially to extrapolate and interpret the result of demonstration tests. 2 figures, 2 tables

Nuclear safety analyses and core design calculations to convert the Texas A & M University Nuclear Science Center reactor to low enrichment uranium fuel. Final report

Energy Technology Data Exchange (ETDEWEB)

Parish, T.A.

1995-03-02

This project involved performing the nuclear design and safety analyses needed to modify the license issued by the Nuclear Regulatory Commission to allow operation of the Texas A& M University Nuclear Science Center Reactor (NSCR) with a core containing low enrichment uranium (LEU) fuel. The specific type of LEU fuel to be considered was the TRIGA 20-20 fuel produced by General Atomic. Computer codes for the neutronic analyses were provided by Argonne National Laboratory (ANL) and the assistance of William Woodruff of ANL in helping the NSCR staff to learn the proper use of the codes is gratefully acknowledged. The codes applied in the LEU analyses were WIMSd4/m, DIF3D, NCTRIGA and PARET. These codes allowed full three dimensional, temperature and burnup dependent calculations modelling the NSCR core to be performed for the first time. In addition, temperature coefficients of reactivity and pulsing calculations were carried out in-house, whereas in the past this modelling had been performed at General Atomic. In order to benchmark the newly acquired codes, modelling of the current NSCR core with highly enriched uranium fuel was also carried out. Calculated results were compared to both earlier licensing calculations and experimental data and the new methods were found to achieve excellent agreement with both. Therefore, even if an LEU core is never loaded at the NSCR, this project has resulted in a significant improvement in the nuclear safety analysis capabilities established and maintained at the NSCR.

Steady State and Transient Fuel Rod Performance Analyses by Pad and Transuranus Codes

International Nuclear Information System (INIS)

Slyeptsov, O.; Slyeptsov, S.; Kulish, G.; Ostapov, A.; Chernov, I.

2013-01-01

The report performed under IAEA research contract No.15370/L2 describes the analysis results of WWER and PWR fuel rod performance at steady state operation and transients by means of PAD and TRANSURANUS codes. The code TRANSURANUS v1m1j09 developed by Institute for of Transuranium Elements (ITU) was used based on the Licensing Agreement N31302. The code PAD 4.0 developed by Westinghouse Electric Company was utilized in the frame of the Ukraine Nuclear Fuel Qualification Project for safety substantiation for the use of Westinghouse fuel assemblies in the mixed core of WWER-1000 reactor. The experimental data for the Russian fuel rod behavior obtained during the steady-state operation in the WWER-440 core of reactor Kola-3 and during the power transients in the core of MIR research reactor were taken from the IFPE database of the OECD/NEA and utilized for assessing the codes themselves during simulation of such properties as fuel burnup, fuel centerline temperature (FCT), fuel swelling, cladding strain, fission gas release (FGR) and rod internal pressure (RIP) in the rod burnup range of (41 - 60) GWD/MTU. The experimental data of fuel behavior at steady-state operation during seven reactor cycles presented by AREVA for the standard PWR fuel rod design were used to examine the code FGR model in the fuel burnup range of (37 - 81) GWD/MTU. (author)

Presence of stratospheric humidity in the ozone column depletion on the west coast of South America

International Nuclear Information System (INIS)

Da Silva, M. Luis; Gutierrez, O. Luis; Morales, S. Luis; Universidad de Chile, Santiago; Torres, C. Arnaldo

2006-01-01

The ozone column depletion over the western coast of South America has been previously explained, based on the existence of winds in the area of the depletion, which cause compression and thinning of the ozone layer. However, the presence of humidity and methane transported by these winds to the stratosphere where the ozone depletion is present gives evidence that these compounds also participate in the depletion of the ozone layer. These two compounds, humidity and methane, are analysed during the ozone depletion of January, 1998. It is observed that when humidity presents fluctuations, ozone has fluctuations too. A maximum of humidity corresponds to a minimum of ozone, but there is a shift in altitude between them. This shift is observed in the stratosphere and upper troposphere and corresponds to approximately 500 m. It is important to point out that during this event El Nino was present and the sources of methane are the Amazon forest and the Pacific Ocean. The data for this study was obtained from NASA and HALOE

Fission gas release and pellet microstructure change of high burnup BWR fuel

International Nuclear Information System (INIS)

Itagaki, N.; Ohira, K.; Tsuda, K.; Fischer, G.; Ota, T.

1998-01-01

UO 2 fuel, with and without Gadolinium, irradiated for three, five, and six irradiation cycles up to about 60 GWd/t pellet burnup in a commercial BWR were studied. The fission gas release and the rim effect were investigated by the puncture test and gas analysis method, OM (optical microscope), SEM (scanning electron microscope), and EPMA (electron probe microanalyzer). The fission gas release rate of the fuel rods irradiated up to six cycles was below a few percent; there was no tendency for the fission gas release to increase abruptly with burnup. On the other hand, microstructure changes were revealed by OM and SEM examination at the rim position with burnup increase. Fission gas was found depleted at both the rim position and the pellet center region using EPMA. There was no correlation between the fission gas release measured by the puncture test and the fission gas depletion at the rim position using EPMA. However, the depletion of fission gas in the center region had good correlation with the fission gas release rate determined by the puncture test. In addition, because the burnup is very large at the rim position of high burnup fuel and also due to the fission rate of the produced Pu, the Xe/Kr ratio at the rim position of high burnup fuel is close to the value of the fission yield of Pu. The Xe/Kr ratio determined by the gas analysis after the puncture test was equivalent to the fuel average but not to the pellet rim position. From the results, it was concluded that fission gas at the rim position was released from the UO 2 matrix in high burnup, however, most of this released fission gas was held in the porous structure and not released from the pellet to the free volume. (author)

Development and preliminary analyses of material balance evaluation model in nuclear fuel cycle

International Nuclear Information System (INIS)

Matsumura, Tetsuo

1994-01-01

Material balance evaluation model in nuclear fuel cycle has been developed using ORIGEN-2 code as basic engine. This model has feature of: It can treat more than 1000 nuclides including minor actinides and fission products. It has flexibility of modeling and graph output using a engineering work station. I made preliminary calculation of LWR fuel high burnup effect (reloading fuel average burnup of 60 GWd/t) on nuclear fuel cycle. The preliminary calculation shows LWR fuel high burnup has much effect on Japanese Pu balance problem. (author)

Supplemental Thermal-Hydraulic Transient Analyses of BR2 in Support of Conversion to LEU Fuel

Energy Technology Data Exchange (ETDEWEB)

Licht, J. [Argonne National Lab. (ANL), Argonne, IL (United States); Dionne, B. [Argonne National Lab. (ANL), Argonne, IL (United States); Sikik, E. [Belgian Nuclear Research Center (SCK-CEN), Mol (Belgium); Van den Branden, G. [Belgian Nuclear Research Center (SCK-CEN), Mol (Belgium); Koonen, E. [Belgian Nuclear Research Center (SCK-CEN), Mol (Belgium)

2016-01-01

Belgian Reactor 2 (BR2) is a research and test reactor located in Mol, Belgium and is primarily used for radioisotope production and materials testing. The Materials Management and Minimization (M3) Reactor Conversion Program of the National Nuclear Security Administration (NNSA) is supporting the conversion of the BR2 reactor from Highly Enriched Uranium (HEU) fuel to Low Enriched Uranium (LEU) fuel. The RELAP5/Mod 3.3 code has been used to perform transient thermal-hydraulic safety analyses of the BR2 reactor to support reactor conversion. A RELAP5 model of BR2 has been validated against select transient BR2 reactor experiments performed in 1963 by showing agreement with measured cladding temperatures. Following the validation, the RELAP5 model was then updated to represent the current use of the reactor; taking into account core configuration, neutronic parameters, trip settings, component changes, etc. Simulations of the 1963 experiments were repeated with this updated model to re-evaluate the boiling risks associated with the currently allowed maximum heat flux limit of 470 W/cm² and temporary heat flux limit of 600 W/cm². This document provides analysis of additional transient simulations that are required as part of a modern BR2 safety analysis report (SAR). The additional simulations included in this report are effect of pool temperature, reduced steady-state flow rate, in-pool loss of coolant accidents, and loss of external cooling. The simulations described in this document have been performed for both an HEU- and LEU-fueled core.

Deterministic methods for multi-control fuel loading optimization

Science.gov (United States)

Rahman, Fariz B. Abdul

We have developed a multi-control fuel loading optimization code for pressurized water reactors based on deterministic methods. The objective is to flatten the fuel burnup profile, which maximizes overall energy production. The optimal control problem is formulated using the method of Lagrange multipliers and the direct adjoining approach for treatment of the inequality power peaking constraint. The optimality conditions are derived for a multi-dimensional multi-group optimal control problem via calculus of variations. Due to the Hamiltonian having a linear control, our optimal control problem is solved using the gradient method to minimize the Hamiltonian and a Newton step formulation to obtain the optimal control. We are able to satisfy the power peaking constraint during depletion with the control at beginning of cycle (BOC) by building the proper burnup path forward in time and utilizing the adjoint burnup to propagate the information back to the BOC. Our test results show that we are able to achieve our objective and satisfy the power peaking constraint during depletion using either the fissile enrichment or burnable poison as the control. Our fuel loading designs show an increase of 7.8 equivalent full power days (EFPDs) in cycle length compared with 517.4 EFPDs for the AP600 first cycle.

Down Under – Aspects of Microbial Fuel Cell’s sewer implementation

OpenAIRE

Andrich, Jonas Maximilian Sven

2017-01-01

Increasing energy demand and simultaneous depletion of raw materials requires us to use existing resources more wisely. Microbial Fuel Cells (MFCs) recover energy from waste water while clearing it. The sewage system with its million-kilometer-length is a highly interesting field for its application. The present work is therefore dedicated to aspects of Microbial Fuel Cells’ sewer implementation. Firstly, the wastewater infrastructure was evaluated with respect to suitable implementation s...

SCALE 5.1 Predictions of PWR Spent Nuclear Fuel Isotopic Compositions

Energy Technology Data Exchange (ETDEWEB)

Radulescu, Georgeta [ORNL; Gauld, Ian C [ORNL; Ilas, Germina [ORNL

2010-03-01

The purpose of this calculation report is to document the comparison to measurement of the isotopic concentrations for pressurized water reactor (PWR) spent nuclear fuel determined with the Standardized Computer Analysis for Licensing Evaluation (SCALE) 5.1 (Ref. ) epletion calculation method. Specifically, the depletion computer code and the cross-section library being evaluated are the twodimensional (2-D) transport and depletion module, TRITON/NEWT,2, 3 and the 44GROUPNDF5 (Ref. 4) cross-section library, respectively, in the SCALE .1 code system.

Attempt to produce silicide fuel elements in Indonesia

International Nuclear Information System (INIS)

Soentono, S.; Suripto, A.

1991-01-01

After the successful experiment to produce U 3 Si 2 powder and U 3 Si 2 -Al fuel plates using depleted U and Si of semiconductor quality, silicide fuel was synthesized using x -Al available at the Fuel Element Production Installation (FEPI) at Serpong, Indonesia. Two full-size U 3 Si 2 -Al fuel elements, having similar specifications to the ones of U 3 O 8 -Al for the RSG-GAS (formerly known as MPR-30), have been produced at the FEPI. All quality controls required have been imposed to the feeds, intermediate, as well as final products throughout the production processes of the two fuel elements. The current results show that these fuel elements are qualified from fabrication point of view, therefore it is expected that they will be permitted to be tested in the RSG-GAS, sometime by the end of 1989, for normal (∝50%) and above normal burn-up. (orig.)

Potential utilization of biomass in production of electricity, heat and transportation fuels including energy combines - Regional analyses and examples; Potentiell avsaettning av biomassa foer produktion av el, vaerme och drivmedel inklusive energikombinat - Regionala analyser och raekneexempel

Energy Technology Data Exchange (ETDEWEB)

Ericsson, Karin; Boerjesson, Paal

2008-01-15

The objective of this study is to analyse how the use of biomass may increase in the next 10-20 years in production of heat, electricity and transportation fuels in Sweden. In these analyses, the biomass is assumed to be used in a resource and cost efficient way. This means for example that the demand for heat determines the potential use of biomass in co-generation of heat and electricity and in energy combines, and that the markets for by-products determine the use of biomass in production of certain transportation fuels. The economic conditions are not analysed in this study. In the heat and electricity production sector, we make regional analyses of the potential use of biomass in production of small-scale heat, district heat, process heat in the forest industry and electricity produced in co-generation with heat in the district heating systems and forest industry. These analyses show that the use of biomass in heat and electricity production could increase from 87 TWh (the use in 2004/2005, excluding small-scale heat production with firewood) to between 113 TWh and 134 TWh, depending on the future expansion of the district heating systems. Geographically, the Stockholm province accounts for a large part of the potential increase owing to the great opportunities for increasing the use of biomass in production of district heat and CHP in this region. In the sector of transportation fuels we applied a partly different approach since we consider the market for biomass-based transportation fuels to be 'unconstrained' within the next 10-20 years. Factors that constrain the production of these fuels are instead the availability of biomass feedstock and the local conditions required for achieving effective production systems. Among the first generation biofuels this report focuses on RME and ethanol from cereals. We estimate that the domestic production of RME and ethanol could amount to up to 1.4 TWh/y and 0.7-3.8 TWh/y, respectively, where the higher figure

Depleted uranium hexafluoride management program : data compilation for the K-25 site

International Nuclear Information System (INIS)

Hartmann, H. M.

2001-01-01

This report is a compilation of data and analyses for the K-25 site on the Oak Ridge Reservation, Oak Ridge, Tennessee. The data were collected and the analyses were done in support of the U.S. Department of Energy (DOE) 1999 Programmatic Environmental Impact Statement for Alternative Strategies for the Long-Term Management and Use of Depleted Uranium Hexafluoride (DOE/EIS-0269). The report describes the affected environment at the K-25 site and summarizes the potential environmental impacts that could result from continued cylinder storage and preparation of cylinders for shipment at the site. It is probable that the cylinders at the K-25 site will be shipped to another site for conversion. Because conversion and long-term storage of the entire inventory at the K-25 site are highly unlikely, these data are not presented in this report. DOE's preferred alternative is to begin converting the depleted uranium hexafluoride inventory as soon as possible to either uranium oxide, uranium metal, or a combination of both, while allowing for use of as much of this inventory as possible

Fast reactor fuel pin behavior analyses in a LOF type transient event

International Nuclear Information System (INIS)

Mizuno, Tomoyasu; Koyama, Shin-ichi; Kaito, Takeji; Uwaba, Tomoyuki; Tanaka, Kenya

2013-06-01

In order to evaluate integrity limiting parameters of fuel pins during fast reactor core transient events, such as fuel center line temperature and cladding maximum temperature, fuel pin behavior calculations were made using the fast reactor fuel pin performance code CEDAR. The temperature histories of fuel pins during a loss of flow (LOF) type transient events was calculated based on Ross and Stoute type gap conductance model and constant gap conductance model, which is used in a core transient calculation code like HIPRAC. The calculated maximum temperatures of cladding and adjacent coolant channel were lower in the case with Ross and Stoute type model than in the case of constant gap conductance model due to the dynamic change of gap conductance of former case. It is indicated that core transient calculations with constant gap conductance give conservative cladding and coolant temperatures than that with Ross and Stoute type gap conductance model which is thought to be realistic. (author)

Methods and techniques of nuclear in-core fuel management

International Nuclear Information System (INIS)

Jong, A.J. de.

1992-04-01

Review of methods of nuclear in-core fuel management (the minimal critical mass problem, minimal power peaking) and calculational techniques: reactorphysical calculations (point reactivity models, continuous refueling, empirical methods, depletion perturbation theory, nodal computer programs); optimization techniques (stochastic search, linear programming, heuristic parameter optimization). (orig./HP)

No Evidence of the Ego-Depletion Effect across Task Characteristics and Individual Differences: A Pre-Registered Study.

Science.gov (United States)

Lurquin, John H; Michaelson, Laura E; Barker, Jane E; Gustavson, Daniel E; von Bastian, Claudia C; Carruth, Nicholas P; Miyake, Akira

2016-01-01

Ego-depletion, a psychological phenomenon in which participants are less able to engage in self-control after prior exertion of self-control, has become widely popular in the scientific community as well as in the media. However, considerable debate exists among researchers as to the nature of the ego-depletion effect, and growing evidence suggests the effect may not be as strong or robust as the extant literature suggests. We examined the robustness of the ego-depletion effect and aimed to maximize the likelihood of detecting the effect by using one of the most widely used depletion tasks (video-viewing attention control task) and by considering task characteristics and individual differences that potentially moderate the effect. We also sought to make our research plan transparent by pre-registering our hypotheses, procedure, and planned analyses prior to data collection. Contrary to the ego-depletion hypothesis, participants in the depletion condition did not perform worse than control participants on the subsequent self-control task, even after considering moderator variables. These findings add to a growing body of evidence suggesting ego-depletion is not a reliable phenomenon, though more research is needed that uses large sample sizes, considers moderator variables, and pre-registers prior to data collection.

Fuel assembly reconstitution

International Nuclear Information System (INIS)

Morgado, Mario M.; Oliveira, Monica G.N.; Ferreira Junior, Decio B.M.; Santos, Barbara O. dos; Santos, Jorge E. dos

2009-01-01

Fuel failures have been happened in Nuclear Power Plants worldwide, without lost of integrity and safety, mainly for the public, environment and power plants workers. The most common causes of these events are corrosion (CRUD), fretting and pellet cladding interaction. These failures are identified by increasing the activity of fission products, verified by chemical analyses of reactor coolant. Through these analyses, during the fourth operation cycle of Angra 2 Nuclear Power Plant, was possible to observe fuel failure indication. This indication was confirmed in the end of the cycle during the unloading of reactor core through leakage tests of fuel assembly, using the equipment called 'In Mast Sipping' and 'Box Sipping'. After confirmed, the fuel assembly reconstitution was scheduled, and happened in April, 2007, where was identified the cause and the fuel rod failure, which was substitute by dummy rods (zircaloy). The cause was fretting by 'debris'. The actions to avoid and prevent fuel assemblies failures are important. The goals of this work are to describe the methodology of fuel assembly reconstitution using the FARE (Fuel Assembly Reconstitution Equipment) system, to describe the results of this task in economic and security factors of the company and show how the fuel assembly failures are identified during operation and during the outage. (author)

Depletion of elements in shock-driven gas

International Nuclear Information System (INIS)

Gondhalekar, P.M.

1985-01-01

The depletion of elements in shocked gas in supernova remnants and in interstellar bubbles is examined. It is shown that elements are depleted in varying degrees in gas filaments shocked to velocities up to 200 km s -1 and that large differences in depletions are observed in gas filaments shocked to similar velocities. In the shocked gas the depletion of an element appears to be correlated with the electron density (or the neutral gas density) in the filaments. This correlation, if confirmed, is similar to the correlation between depletion and mean density of gas in the clouds in interstellar space. (author)

Tools for LWR spent fuel characterization: Assembly classes and fuel designs

International Nuclear Information System (INIS)

Moore, R.S.; Notz, K.J.

1991-01-01

The Characteristics Data Base (CDB) is sponsored by the DOE's Office of Civilian Radioactive Waste Management (OCRWM). The CDB provides a single, comprehensive source of data pertaining to radioactive wastes that will or may require geologic disposal, including detailed data describing the physical, quantitative, and radiological characteristics of light-water reactor (LWR) spent fuel. In developing the CDB, tools for the classification of fuel assembly types have been developed. The assembly class scheme is particularly useful for size- and handling-based describes these tools and presents results of their applications in the areas of fuel assembly type identification, characterization of projected discharges, cask accommodation analyses, and defective fuel analyses. Suggestions for additional applications are also made. 7 refs., 1 fig., 2 tabs

Towards proliferation-resistant thorium fuels

International Nuclear Information System (INIS)

Alhaj, M. Yousif; Mohamed, Nader M.A.; Badawi, Alya; Abou-Gabal, Hanaa H.

2017-01-01

Thorium-plutonium mixture is proposed as alternative nuclear reactor fuel to incinerate the increasing stockpile plutonium. However, this fuel will produce an amount of uranium with about 90% 233U at applicable discharge burnups (60GWD/MTU). This research focuses on proposing an optimum non proliferative thorium fuel, by adding a small amount of 238U to reduce the attractiveness of the resultant uranium. Three types of additive which contain 238U were used: 4.98% enriched, natural and depleted uranium. We found that introducing uranium to the fresh thorium-plutonium fuel reduces its performance even if the uranium was enriched up to 5%. While uranium admixtures reduce the quality of the reprocessed uranium, it also increases the quality of the plutonium. However, this increase is very low compared to the reduced quality of uranium. We also found that using uranium as admixture for thorium-plutonium mixed fuel increases the critical mass of the extracted uranium by a factor of two when using only 1% admixture of uranium. The higher the percentage of uranium admixture the higher the critical mass of the reprocessed one.

Depleted uranium - influence on the health and environment; Ochudobneny uran - vplyva na zdravie a zivotne prostredie

Energy Technology Data Exchange (ETDEWEB)

Rosskopfova, O [Katedra jadrovej chemie, Prirodovedecka fakulta, Univerzita Komenskeho, Bratislava (Slovakia)

2002-07-01

The uranium as radioactive element occurs in low concentrations in all components of environment. In the sample of natural uranium the isotope U-235 has the highest share (99.27 weight per cent). Chemical toxicity of uranium is comparable with toxicity of the elements like As and Pb. Depleted uranium is adjoining product in the production of enriched uranium, which is required in the production of the nuclear fuel and in the production of material used in nuclear arms. It mainly includes isotope U-238, and the content of isotopes U-235 and U-234 is sharply lowered. According to NRC depleted uranium is defined like uranium, in which percentile share of isotope U-235 is less than 0.711 weight percent. The activity of depleted uranium from viewpoint of external irradiation does not represent higher risk. Much higher risk for man represent the neurotoxic effects of uranium, which can get into human body by inhaling of dispersed particles, of contaminated dust and aerosols from atmosphere or by consumption of contaminated foodstuffs and water. Basic dangerous of irradiation by depleted uranium are mainly aerosols, which increase the probability of occurrence of lung cancer. The next dangerous is the damage of another organs like kidneys, liver and bones, where these aerosols are transported by blood like oxides from the lungs. In the environmental parts because of presence of natural uranium the depleted uranium is difficultly identifiably by standard detective methods. Thus it is necessary to use suitable radiochemical separative methods in combination with suitable detective method. (author)

Analyses with the FSTATE code: fuel performance in destructive in-pile experiments

International Nuclear Information System (INIS)

Bauer, T.H.; Meek, C.C.

1982-01-01

Thermal-mechanical analysis of a fuel pin is an essential part of the evaluation of fuel behavior during hypothetical accident transients. The FSTATE code has been developed to provide this required computational ability in situations lacking azimuthal symmetry about the fuel-pin axis by performing 2-dimensional thermal, mechanical, and fission gas release and redistribution computations for a wide range of possible transient conditions. In this paper recent code developments are described and application is made to in-pile experiments undertaken to study fast-reactor fuel under accident conditions. Three accident simulations, including a fast and slow ramp-rate overpower as well as a loss-of-cooling accident sequence, are used as representative examples, and the interpretation of STATE computations relative to experimental observations is made

German nuclear fuel exports and imports 1991

International Nuclear Information System (INIS)

Anon.

1993-01-01

The statistics compiled by the German Federal Office for Trade and Industry (Bundesamt fuer Wirtschaft) for the Federal Ministry for the Environment, Conservation of Nature, and Reactor Safety of imports and exports of nuclear fuels and source materials in 1991 show a major drop by 33.8% in imports and a pronounced rise by 191.5% in exports, compared to the levels in the previous year. Source material for the purposes of these statistics refers only to uranium concentrate. Quantitatively, the biggest import items are source materials, depleted uranium, and uranium enriched up to 3%. Exports of unirradiated material quantitatively comprise mainly depleted uranium, source material, and uranium enriched up to 10%. (orig.) [de

Fuel behavior in advanced water reactors

International Nuclear Information System (INIS)

Bolme, A.B.

1996-01-01

Fuel rod behavior of advanced pressurized water reactors under steady state conditions has been investigated in this study. System-80+ and Westinghouse Vantage-5 fuels have been considered as advanced pressurized water reactor fuels to be analyzed. The purpose of this study is to analyze the sensitivity of ditferent models and the effect of selected design parameters on the overall fuel behavior. FRAPCON-II computer code has been used for the analyses. Different modelling options of FRAPCON-II have also been considered in these analyses. Analyses have been performed in two main parts. In the first part, effects of operating conditions on fuel behavior have been investigated. First, fuel rod response under normal operating conditions has been analyzed. Then, fuel rod response to different fuel ratings has been calculated. In the second part, in order to estimate the effect of design parameters on fuel behavior, parametric analyses have been performed. In this part, the effects of initial gap thickness, as fabricated fuel density, and initial fill gas pressure on fuel behavior have been analyzed. The computations showed that both of the fuel rods used in this study operate within the safety limits. However, FRAPCON-II modelling options have been resulted in different behavior due to their modelling characteristics. Hence, with the absence of experimental data, it is difficult to make assesment for the best fuel parameters. It is also difficult to estimate error associated with the results. To improve the performance of the code, it is necessary to develop better experimental correlations for material properties in order to analyze the eftect ot considerably different design parameters rather than nominal rod parameters

Are relative depletions altered inside diffuse clouds?

International Nuclear Information System (INIS)

Joseph, C.L.

1988-01-01

The data of Jenkins, Savage, and Spitzer (1986) were used to analyze interstellar abundances and depletions of Fe, P, Mg, and Mn toward 37 stars, spanning nearly 1.0 (dex) in mean line-of-sight depletion. It was found that the depletions of these elements are linearly correlated and do not show evidence of differences in the rates of depletion or sputtering from one element to another. For a given level of overall depletion, the sightline-to-sightline rms variance in the depletion for each of these elements was less than 0.16 (dex), which is significantly smaller than is the element-to-element variance. The results suggest that, for most diffuse lines of sight, the relative abundances of these elements are set early in the lifetime of the grains and are not altered significantly thereafter. 53 references

Fuel assembly for FBR type reactor and reactor core thereof

International Nuclear Information System (INIS)

Kobayashi, Kaoru.

1998-01-01

The present invention provides a fuel assembly to be loaded to a reactor core of a large sized FBR type reactor, in which a coolant density coefficient can be reduced without causing power peaking in the peripheral region of neutron moderators loaded in the reactor core. Namely, the fuel assembly for the FBR type reactor comprises a plurality of fission product-loaded fuel rods and a plurality of fertile material-loaded fuel rods and one or more rods loading neutron moderators. In this case, the plurality of fertile material-loaded fuel rods are disposed to the peripheral region of the neutron moderator-loaded rods. The plurality of fission product-loaded fuel rods are disposed surrounding the peripheral region of the plurality of fertile material-loaded fuel rods. The neutron moderator comprises zirconium hydride, yttrium hydride and calcium hydride. The fission products are mixed oxide fuels. The fertile material comprises depleted uranium or natural uranium. (I.S.)

Integrated fuel-cycle models for fast breeder reactors

International Nuclear Information System (INIS)

Ott, K.O.; Maudlin, P.J.

1981-01-01

Breeder-reactor fuel-cycle analysis can be divided into four different areas or categories. The first category concerns questions about the spatial variation of the fuel composition for single loading intervals. Questions of the variations in the fuel composition over several cycles represent a second category. Third, there is a need for a determination of the breeding capability of the reactor. The fourth category concerns the investigation of breeding and long-term fuel logistics. Two fuel-cycle models used to answer questions in the third and fourth area are presented. The space- and time-dependent actinide balance, coupled with criticality and fuel-management constraints, is the basis for both the Discontinuous Integrated Fuel-Cycle Model and the Continuous Integrated Fuel-Cycle Model. The results of the continuous model are compared with results obtained from detailed two-dimensional space and multigroup depletion calculations. The continuous model yields nearly the same results as the detailed calculation, and this is with a comparatively insignificant fraction of the computational effort needed for the detailed calculation. Thus, the integrated model presented is an accurate tool for answering questions concerning reactor breeding capability and long-term fuel logistics. (author)

Feasibility assessment of burnup credit in the criticality analysis of shipping casks with boiling water reactor spent fuel

International Nuclear Information System (INIS)

Broadhead, B.L.

1991-08-01

Considerable interest in the allowance of reactivity credit for the exposure history of power reactor fuel currently exists. This ''burnup credit'' issue has the potential to greatly reduce risk and cost when applied to the design and certification of spent fuel casks used for transportation and storage. Recently, analyses have demonstrated the technical feasibility and estimated the risk and economic incentives for allowing burnup credit in pressurized water reactor (PWR) spent fuel shipping cask applications. This report summarizes the extension of the previous PWR technical feasibility assessment to boiling water reactor (BWR) fuel. This feasibility analysis aims to apply simple methods that adequately characterize the time-dependent isotopic compositions of typical BWR fuel. An initial analysis objective was to identify a simple and reliable method for characterizing BWR spent fuel. Two different aspects of fuel characterization were considered:l first, the generation of burn- up dependent material interaction probabilities; second, the prediction of material inventories over time (depletion). After characterizing the spent fuel at various stages of exposure and decay, three dimensional (3-D) models for an infinite array of assemblies and, in several cases, infinite arrays of assemblies in a typical shipping cask basket were analyzed. Results for assemblies without a basket provide reactivity control requirements as a function of burnup and decay, while results including the basket allow assessment of typical basket configurations to provide sufficient reactivity control for spent BWR fuel. Resulting basket worths and reactivity trends over time are then evaluated to determine whether burnup credit is needed and feasible in BWR applications

Fusion yield rate recovery by escaping hot-spot fast ions in the neighboring fuel layer

Science.gov (United States)

Tang, Xian-Zhu; McDevitt, C. J.; Guo, Zehua; Berk, H. L.

2014-02-01

Free-streaming loss by fast ions can deplete the tail population in the hot spot of an inertial confinement fusion (ICF) target. Escaping fast ions in the neighboring fuel layer of a cryogenic target can produce a surplus of fast ions locally. In contrast to the Knudsen layer effect that reduces hot-spot fusion reactivity due to tail ion depletion, the inverse Knudsen layer effect increases fusion reactivity in the neighboring fuel layer. In the case of a burning ICF target in the presence of significant hydrodynamic mix which aggravates the Knudsen layer effect, the yield recovery largely compensates for the yield reduction. For mix-dominated sub-ignition targets, the yield reduction is the dominant process.

Development of four-year fuel cycle based on the advanced fuel assembly with uranium-gadolinium fuel and its implementation to the operating WWER-440 units

International Nuclear Information System (INIS)

Lunin, G.; Novikov, A.; Pavlov, V.; Pavlovichev, P.; Filimonov, P.

2000-01-01

Over the past few years in Russia the investigations aimed at the increase of the reliability, safety and efficiency of operation of the WWER-1000 reactors as well as of its competitiveness in the world market were carried out. In the frame of these investigations the four-year fuel cycle, based on advanced fuel assemblies with zirconium alloy spacer grids and guide tubes and with fuel pellet having a reduced diameter of the central hole (1,5 mm), has been developed. For the compensation of a part of excess reactivity, Gd 2 O 3 integrated burnable absorbers are used. CPS absorbing rods contain a combine absorber (B 4 C + Dy 2 O 3 *TiO 2 ). A part of depleted fuel is located on the core periphery. The algorithms controlling the reactor power and power distribution have been updated. For checking of the solutions adopted and for verification of code package developed at the RRC 'Kurchatov Institute' the wide-scale experimental operation of advanced FA and its individual components is carried out. (Authors)