Review of Axial Burnup Distribution Considerations for Burnup Credit Calculations
Wagner, J.C.; DeHart, M.D.
2000-03-01
This report attempts to summarize and consolidate the existing knowledge on axial burnup distribution issues that are important to burnup credit criticality safety calculations. Recently released Nuclear Regulatory Commission (NRC) staff guidance permits limited burnup credit, and thus, has prompted resolution of the axial burnup distribution issue. The reactivity difference between the neutron multiplication factor (keff) calculated with explicit representation of the axial burnup distribution and keff calculated assuming a uniform axial burnup is referred to as the ``end effect.'' This end effect is shown to be dependent on many factors, including the axial-burnup profile, total accumulated burnup, cooling time, initial enrichment, assembly design, and the isotopics considered (i.e., actinide-only or actinides plus fission products). Axial modeling studies, efforts related to the development of axial-profile databases, and the determination of bounding axial profiles are also discussed. Finally, areas that could benefit from further efforts are identified.
Detailed Burnup Calculations for Testing Nuclear Data
NASA Astrophysics Data System (ADS)
Leszczynski, F.
2005-05-01
A general method (MCQ) has been developed by introducing a microscopic burnup scheme that uses the Monte Carlo calculated fluxes and microscopic reaction rates of a complex system and a depletion code for burnup calculations as a basis for solving nuclide material balance equations for each spatial region in which the system is divided. Continuous energy-dependent cross-section libraries and full 3D geometry of the system can be input for the calculations. The resulting predictions for the system at successive burnup time steps are thus based on a calculation route where both geometry and cross sections are accurately represented, without geometry simplifications and with continuous energy data, providing an independent approach for benchmarking other methods and nuclear data of actinides, fission products, and other burnable absorbers. The main advantage of this method over the classical deterministic methods currently used is that the MCQ System is a direct 3D method without the limitations and errors introduced on the homogenization of geometry and condensation of energy of deterministic methods. The Monte Carlo and burnup codes adopted until now are the widely used MCNP and ORIGEN codes, but other codes can be used also. For using this method, there is need of a well-known set of nuclear data for isotopes involved in burnup chains, including burnable poisons, fission products, and actinides. For fixing the data to be included in this set, a study of the present status of nuclear data is performed, as part of the development of the MCQ method. This study begins with a review of the available cross-section data of isotopes involved in burnup chains for power and research nuclear reactors. The main data needs for burnup calculations are neutron cross sections, decay constants, branching ratios, fission energy, and yields. The present work includes results of selected experimental benchmarks and conclusions about the sensitivity of different sets of cross-section data for burnup calculations, using some of the main available evaluated nuclear data files (ENDF-B-VI-Rel.8, JEFF-3.0, JENDL-3.3), on an isotope-by-isotope basis as much as possible. The selected experimental burnup benchmarks are reference cases for LWR and HWR reactors, with analysis of isotopic composition as a function of burnup. For LWR (H2O-moderated uranium oxide lattices) four benchmarks are included: ATM-104 NEA Burnup credit criticality benchmark; Yankee-Rowe Core V; H.B.Robinson Unit 2 and Turkey Point Unit 3. For HWR (D2O-moderated uranium oxide cluster lattices), three benchmarks were selected: NPD-19-rod Fuel Clusters; Pickering-28-rod Fuel Clusters; and Bruce-37-rod Fuel Clusters. The isotopes with experimental concentration data included in these benchmarks are: Se-79, Sr90, Tc99, Ru106, Sn126, Sb125,1129, Cs133-137, Nd143, 145, Sm149-150, 152, Eul53-155, U234-235, 238, Np237, Pu238-242, Am241-243, and Cm242-248. Results and analysis of differences between calculated and measured absolute and/or relative concentrations of these isotopes for the seven benchmarks are included in this work.
Gauntt, Randall O.; Powers, Dana Auburn; Ashbaugh, Scott G.; Leonard, Mark Thomas; Longmire, Pamela
2010-04-01
In this study, risk-significant pressurized-water reactor severe accident sequences are examined using MELCOR 1.8.5 to explore the range of fission product releases to the reactor containment building. Advances in the understanding of fission product release and transport behavior and severe accident progression are used to render best estimate analyses of selected accident sequences. Particular emphasis is placed on estimating the effects of high fuel burnup in contrast with low burnup on fission product releases to the containment. Supporting this emphasis, recent data available on fission product release from high-burnup (HBU) fuel from the French VERCOR project are used in this study. The results of these analyses are treated as samples from a population of accident sequences in order to employ approximate order statistics characterization of the results. These trends and tendencies are then compared to the NUREG-1465 alternative source term prescription used today for regulatory applications. In general, greater differences are observed between the state-of-the-art calculations for either HBU or low-burnup (LBU) fuel and the NUREG-1465 containment release fractions than exist between HBU and LBU release fractions. Current analyses suggest that retention of fission products within the vessel and the reactor coolant system (RCS) are greater than contemplated in the NUREG-1465 prescription, and that, overall, release fractions to the containment are therefore lower across the board in the present analyses than suggested in NUREG-1465. The decreased volatility of Cs2MoO4 compared to CsI or CsOH increases the predicted RCS retention of cesium, and as a result, cesium and iodine do not follow identical behaviors with respect to distribution among vessel, RCS, and containment. With respect to the regulatory alternative source term, greater differences are observed between the NUREG-1465 prescription and both HBU and LBU predictions than exist between HBU and LBU analyses. Additionally, current analyses suggest that the NUREG-1465 release fractions are conservative by about a factor of 2 in terms of release fractions and that release durations for in-vessel and late in-vessel release periods are in fact longer than the NUREG-1465 durations. It is currently planned that a subsequent report will further characterize these results using more refined statistical methods, permitting a more precise reformulation of the NUREG-1465 alternative source term for both LBU and HBU fuels, with the most important finding being that the NUREG-1465 formula appears to embody significant conservatism compared to current best-estimate analyses.
High-burnup core design using minor actinide-containing metal fuel
Ohta, Hirokazu; Ogata, Takanari; Obara, T.
2013-07-01
A neutronic design study of metal fuel fast reactor (FR) cores is conducted on the basis of an innovative fuel design concept to achieve an extremely high burnup and realize an efficient fuel cycle system. Since it is expected that the burnup reactivity swing will become extremely large in an unprecedented high burnup core, minor actinides (MAs) from light water reactors (LWRs) are added to fresh fuel to improve the core internal conversion. Core neutronic analysis revealed that high burnups of about 200 MWd/kg for a small-scale core and about 300 MWd/kg for a large-scale core can be attained while suppressing the burnup reactivity swing to almost the same level as that of conventional cores with normal burnup. An actinide burnup analysis has shown that the MA consumption ratio is improved to about 60% and that the accumulated MAs originating from LWRs can be efficiently consumed by the high-burnup metal fuel FR. (authors)
Calculation of burnup of a black neutron absorber
Yudkevich, M. S., E-mail: umark@adis.vver.kiae.ru [Russian Research Centre Kurchatov Institute (Russian Federation)
2011-12-15
The procedure of calculation of burnup of fuel and strong neutron absorber in a nuclear reactor is described. The method proposed here makes it possible to avoid difficulties associated with heterogeneous blocking of the absorption cross section. The effectiveness of the method is demonstrated by an example.
MONTE-CARLO BURNUP CALCULATION UNCERTAINTY QUANTIFICATION AND PROPAGATION DETERMINATION
Sternat, M.; Nichols, T.
2011-06-09
Reactor burnup or depletion codes are used thoroughly in the fields of nuclear forensics and nuclear safeguards. Two common codes include MONTEBURNS and MCNPX/CINDER. These are Monte-Carlo depletion routines utilizing MCNP for neutron transport calculations and either ORIGEN or CINDER for burnup calculations. Uncertainties exist in the MCNP steps, but this information is not passed to the depletion calculations or saved. To quantify this transport uncertainty and determine how it propagates between burnup steps, a statistical analysis of multiple repeated depletion runs is performed. The reactor model chosen is the Oak Ridge Research Reactor (ORR) in a single assembly, infinite lattice configuration. This model was burned for a 150 day cycle broken down into three steps. The output isotopics as well as effective multiplication factor (k-effective) were tabulated and histograms were created at each burnup step using the Scott Method to determine the bin width. The distributions for each code are a statistical benchmark and comparisons made. It was expected that the gram quantities and k-effective histograms would produce normally distributed results since they were produced from a Monte-Carlo routine, but some of the results appear to not. Statistical analyses are performed using the {chi}{sup 2} test against a normal distribution for the k-effective results and several isotopes including {sup 134}Cs, {sup 137}Cs, {sup 235}U, {sup 238}U, {sup 237}Np, {sup 238}Pu, {sup 239}Pu, and {sup 240}Pu.
MONTE-CARLO BURNUP CALCULATION UNCERTAINTY QUANTIFICATION AND PROPAGATION DETERMINATION
Nichols, T.; Sternat, M.; Charlton, W.
2011-05-08
MONTEBURNS is a Monte-Carlo depletion routine utilizing MCNP and ORIGEN 2.2. Uncertainties exist in the MCNP transport calculation, but this information is not passed to the depletion calculation in ORIGEN or saved. To quantify this transport uncertainty and determine how it propagates between burnup steps, a statistical analysis of a multiple repeated depletion runs is performed. The reactor model chosen is the Oak Ridge Research Reactor (ORR) in a single assembly, infinite lattice configuration. This model was burned for a 25.5 day cycle broken down into three steps. The output isotopics as well as effective multiplication factor (k-effective) were tabulated and histograms were created at each burnup step using the Scott Method to determine the bin width. It was expected that the gram quantities and k-effective histograms would produce normally distributed results since they were produced from a Monte-Carlo routine, but some of results do not. The standard deviation at each burnup step was consistent between fission product isotopes as expected, while the uranium isotopes created some unique results. The variation in the quantity of uranium was small enough that, from the reaction rate MCNP tally, round off error occurred producing a set of repeated results with slight variation. Statistical analyses were performed using the {chi}{sup 2} test against a normal distribution for several isotopes and the k-effective results. While the isotopes failed to reject the null hypothesis of being normally distributed, the {chi}{sup 2} statistic grew through the steps in the k-effective test. The null hypothesis was rejected in the later steps. These results suggest, for a high accuracy solution, MCNP cell material quantities less than 100 grams and greater kcode parameters are needed to minimize uncertainty propagation and minimize round off effects.
Burnup concept for a long-life fast reactor core using MCNPX.
Holschuh, Thomas Vernon,; Lewis, Tom Goslee,; Parma, Edward J.,
2013-02-01
This report describes a reactor design with a burnup concept for a long-life fast reactor core that was evaluated using Monte Carlo N-Particle eXtended (MCNPX). The current trend in advanced reactor design is the concept of a small modular reactor (SMR). However, very few of the SMR designs attempt to substantially increase the lifetime of a reactor core, especially without zone loading, fuel reshuffling, or other artificial mechanisms in the core that %E2%80%9Cflatten%E2%80%9D the power profile, including non-uniform cooling, non-uniform moderation, or strategic poison placement. Historically, the limitations of computing capabilities have prevented acceptable margins in the temporal component of the spatial excess reactivity in a reactor design, due primarily to the error in burnup calculations. This research was performed as an initial scoping analysis into the concept of a long-life fast reactor. It can be shown that a long-life fast reactor concept can be modeled using MCNPX to predict burnup and neutronics behavior. The inherent characteristic of this conceptual design is to minimize the change in reactivity over the lifetime of the reactor. This allows the reactor to operate substantially longer at full power than traditional Light Water Reactors (LWRs) or other SMR designs. For the purpose of this study, a single core design was investigated: a relatively small reactor core, yielding a medium amount of power (~200 to 400 MWth). The results of this scoping analysis were successful in providing a preliminary reactor design involving metal U-235/U-238 fuel with HT-9 fuel cladding and sodium coolant at a 20% volume fraction.
Accident source terms for boiling water reactors with high burnup cores.
Gauntt, Randall O.; Powers, Dana Auburn; Leonard, Mark Thomas
2007-11-01
The primary objective of this report is to provide the technical basis for development of recommendations for updates to the NUREG-1465 Source Term for BWRs that will extend its applicability to accidents involving high burnup (HBU) cores. However, a secondary objective is to re-examine the fundamental characteristics of the prescription for fission product release to containment described by NUREG-1465. This secondary objective is motivated by an interest to understand the extent to which research into the release and behaviors of radionuclides under accident conditions has altered best-estimate calculations of the integral response of BWRs to severe core damage sequences and the resulting radiological source terms to containment. This report, therefore, documents specific results of fission product source term analyses that will form the basis for the HBU supplement to NUREG-1465. However, commentary is also provided on observed differences between the composite results of the source term calculations performed here and those reflected NUREG-1465 itself.
Using Laguerre polynomials to compute the matrix exponential in burnup calculations
She, D.; Zhu, A.; Wang, K.
2012-07-01
An essential part of burnup analysis is to solve the burnup equations. The burnup equations can be regarded as a first-order linear system and solved by means of matrix exponential methods. Because of its large spectrum, it is difficult to compute the exponential of the burnup matrix. Conventional methods of computing the matrix exponential, such as the truncated Taylor expansion and the Pade approximation, are not applicable to burnup calculations. Recently the Chebyshev Rational Approximation Method (CRAM) has been applied to solve burnup matrix exponential and shown to be robust and accurate. However, the main defect of CRAM is that its coefficients are not easy to obtain. In this paper, an orthogonal polynomial expansion method, called Laguerre Polynomial Approximation Method (LPAM), is proposed to compute the matrix exponential in burnup calculations. The polynomial sequence of LPAM can be easily computed in any order and thus LPAM is quite convenient to be utilized into burnup codes. Two typical test cases with the decay and cross-section data taken from the standard ORIGEN 2.1 libraries are calculated for validation, against the reference results provided by CRAM of 14 order. Numerical results show that, LPAM is sufficiently accurate for burnup calculations. The influences of the parameters on the convergence of LPAM are also discussed. (authors)
Methodology for embedded transport core calculation
NASA Astrophysics Data System (ADS)
Ivanov, Boyan D.
The progress in the Nuclear Engineering field leads to developing new generations of Nuclear Power Plants (NPP) with complex rector core designs, such as cores loaded partially with mixed-oxide (MOX) fuel, high burn-up loadings, and cores with advanced designs of fuel assemblies and control rods. Such heterogeneous cores introduce challenges for the diffusion theory that has been used for several decades for calculations of the current Pressurized Water Rector (PWR) cores. To address the difficulties the diffusion approximation encounters new core calculation methodologies need to be developed by improving accuracy, while preserving efficiency of the current reactor core calculations. In this thesis, an advanced core calculation methodology is introduced, based on embedded transport calculations. Two different approaches are investigated. The first approach is based on embedded finite element (FEM), simplified P3 approximation (SP3), fuel assembly (FA) homogenization calculation within the framework of the diffusion core calculation with NEM code (Nodal Expansion Method). The second approach involves embedded FA lattice physics eigenvalue calculation based on collision probability method (CPM) again within the framework of the NEM diffusion core calculation. The second approach is superior to the first because most of the uncertainties introduced by the off-line cross-section generation are eliminated.
Boyarinov, V. F.; Davidenko, V. D.; Polismakov, A. A.; Tsibulsky, V. F. [Russian Research Center Kurchatov Inst., Nuclear Reactor Inst., 123182, Moscow (Russian Federation)
2006-07-01
Development of the SUHAM-U code for burnup calculations of VVER-1000 fuel assemblies with uranium and MOX fuel is described. Developed SUHAM-U code has capacity to calculate burnup in each fuel or poison zone of each cell of VVER-1000 fuel assembly. In so doing Surface Harmonics Method is used for calculation of the detail neutron spectra in fuel assembly at separated burnup values. Verification of SUHAM-U code by burnup calculations of VVER-1000 fuel assemblies with uranium and MOX fuel has been carried out. Comparisons were carried out with calculations by UNK and RECOL codes. UNK code uses the first collisions probabilities method for solution of the neutron transport equation and RECOL code uses Monte-Carlo method with point-wise continues energy presentation of cross-sections. The main conclusion of all comparisons is the SUHAM-U code calculates the fuel burnup of VVER-1000 fuel assemblies with uranium and MOX fuel with enough high accuracy. Time expenditures are adduced. (authors)
Karpushkin, T. Yu.
2012-12-15
A technique to calculate the burnup of materials of cells and fuel assemblies using the matrices of first-flight neutron collision probabilities rebuilt at a given burnup step is presented. A method to rebuild and correct first collision probability matrices using average chords prior to the first neutron collision, which are calculated with the help of geometric modules of constructed stochastic neutron trajectories, is described. Results of calculation of the infinite multiplication factor for elementary cells with a modified material composition compared to the reference one as well as calculation of material burnup in the cells and fuel assemblies of a VVER-1000 are presented.
OECD/NEA burnup credit calculational criticality benchmark Phase I-B results
DeHart, M.D.; Parks, C.V.; Brady, M.C.
1996-06-01
In most countries, criticality analysis of LWR fuel stored in racks and casks has assumed that the fuel is fresh with the maximum allowable initial enrichment. This assumption has led to the design of widely spaced and/or highly poisoned storage and transport arrays. If credit is assumed for fuel burnup, initial enrichment limitations can be raised in existing systems, and more compact and economical arrays can be designed. Such reliance on the reduced reactivity of spent fuel for criticality control is referred to as burnup credit. The Burnup Credit Working Group, formed under the auspices of the Nuclear Energy Agency of the Organization for Economic Cooperation and Development, has established a set of well-defined calculational benchmarks designed to study significant aspects of burnup credit computational methods. These benchmarks are intended to provide a means for the intercomparison of computer codes, methods, and data applied in spent fuel analysis. The benchmarks have been divided into multiple phases, each phase focusing on a particular feature of burnup credit analysis. This report summarizes the results and findings of the Phase I-B benchmark, which was proposed to provide a comparison of the ability of different code systems and data libraries to perform depletion analysis for the prediction of spent fuel isotopic concentrations. Results included here represent 21 different sets of calculations submitted by 16 different organizations worldwide and are based on a limited set of nuclides determined to have the most important effect on the neutron multiplication factor of light-water-reactor spent fuel. A comparison of all sets of results demonstrates that most methods agree to within 10% in the ability to estimate the spent fuel concentrations of most actinides. All methods agree within 11% about the average for all fission products studied. Most deviations are less than 10%, and many are less than 5%. The exceptions are Sm 149, Sm 151, and Gd 155.
Performance upgrades to the MCNP6 burnup capability for large scale depletion calculations
Fensin, M. L.; Galloway, J. D.; James, M. R.
2015-04-11
The first MCNP based inline Monte Carlo depletion capability was officially released from the Radiation Safety Information and Computational Center as MCNPX 2.6.0. With the merger of MCNPX and MCNP5, MCNP6 combined the capability of both simulation tools, as well as providing new advanced technology, in a single radiation transport code. The new MCNP6 depletion capability was first showcased at the International Congress for Advancements in Nuclear Power Plants (ICAPP) meeting in 2012. At that conference the new capabilities addressed included the combined distributive and shared memory parallel architecture for the burnup capability, improved memory management, physics enhancements, and newmore »predictability as compared to the H.B Robinson Benchmark. At Los Alamos National Laboratory, a special purpose cluster named “tebow,” was constructed such to maximize available RAM per CPU, as well as leveraging swap space with solid state hard drives, to allow larger scale depletion calculations (allowing for significantly more burnable regions than previously examined). As the MCNP6 burnup capability was scaled to larger numbers of burnable regions, a noticeable slowdown was realized.This paper details two specific computational performance strategies for improving calculation speedup: (1) retrieving cross sections during transport; and (2) tallying mechanisms specific to burnup in MCNP. To combat this slowdown new performance upgrades were developed and integrated into MCNP6 1.2.« less
Using ORIGEN and MCNP to calculate reactor criticals and burnup effects
Bowen, D.; Busch, R.D. [Univ. of New Mexico, Albuquerque, NM (United States)
1997-12-01
The purpose of this modeling effort was to verify the applicability of using ORIGEN-S and MCNP to the analysis of spent fuel of various enrichments and burnups. By comparing the results of criticality studies using MCNP and ORIGEN-S with the measured k{sub eff} of 1.0, the suitability of the coupled ORIGEN-S/ MCNP package was determined. This study presents the results of the benchmark modeling of five pressurized water reactor (PWR) critical configurations. For these analyses, a combination of ORIGEN-S and MCNP was used to analyze the fuel depletion and criticality of five power reactor core configuration.
CANDLE: The New Burnup Strategy
Sekimoto, Hiroshi; Ryu, Kouichi; Yoshimura, Yoshikane
2001-11-15
The new burnup strategy CANDLE (Constant Axial shape of Neutron flux, nuclide densities and power shape During Life of Energy production) is proposed. With this burnup strategy, distributions of fuel nuclide densities, neutron flux, and power density move with the same constant speed and without any change in their shapes. The excess reactivity is constant during the burnup. Therefore, any control mechanisms for the burnup are not required. Calculation procedures are presented to find these shapes and the speed of the burning region with the neutron multiplication factor of a reactor employing this burnup strategy.To demonstrate the CANDLE burnup strategy, it is applied to a fast reactor with excellent neutron economy. Only the initially built reactor requires some fissile material such as plutonium or enriched uranium for the nuclear ignition region of its core, but only natural uranium or depleted uranium is required for the other region. Succeeding reactors require only natural or depleted uranium since the burning region of the previous reactor can be utilized for the ignition region. The life of a reactor can be made longer by elongating the core height. The drift speed of the burning region for the presented fast reactor design is {approx}4 cm/yr, which is a preferable value for designing a long-life reactor. The burnup of spent fuel is {approx}40%. It is equivalent to 40% utilization of natural uranium without reprocessing and enrichment.
Advanced Fuel Cycle Initiative - Projected Linear Heat Generation Rate and Burnup Calculations
Richard G. Ambrosek; Gray S. Chang; Debbie J. Utterbeck
2005-02-01
This report provides documentation of the physics analysis performed to determine the linear heat generation rate (LHGR) and burnup calculations for the Advanced Fuel Cycle Initiative (AFCI) tests, AFC-1D, AFC-1H, and AFC-1G. The AFC-1D and AFC-1H tests consists of low-fertile metallic fuel compositions and the AFC-1G test consists of non-fertile and low-fertile nitride compositions. These tests will be irradiated in the East Flux Trap (EFT) positions E1, E2, and E3, respectively, during Advanced Test Reactor (ATR) Cycle 135B.
Burnup calculations and chemical analysis of irradiated fuel samples studied in LWR-PROTEUS phase II
Grimm, P.; Guenther-Leopold, I. [Paul Scherrer Inst., CH-5232 Villigen PSI (Switzerland); Berger, H. D. [AREVA NP GmbH, FEEP, Bunsenstrasse 43, D-91058 Erlangen (Germany)
2006-07-01
The isotopic compositions of 5 UO{sub 2} samples irradiated in a Swiss PWR power plant, which were investigated in the LWR-PROTEUS Phase II programme, were calculated using the CASMO-4 and BOXER assembly codes. The burnups of the samples range from 50 to 90 MWd/kg. The results for a large number of actinide and fission product nuclides were compared to those of chemical analyses performed using a combination of chromatographic separation and mass spectrometry. A good agreement of calculated and measured concentrations is found for many of the nuclides investigated with both codes. The concentrations of the Pu isotopes are mostly predicted within {+-}10%, the two codes giving quite different results, except for {sup 242}Pu. Relatively significant deviations are found for some isotopes of Cs and Sm, and large discrepancies are observed for Eu and Gd. The overall quality of the predictions by the two codes is comparable, and the deviations from the experimental data do not generally increase with burnup. (authors)
Mohammad Javed Khan; Aslam; Nasir Ahmad
2005-01-01
Standard reactor simulation codes WIMS-D\\/4 and CITATION were employed to analyze the proliferation resistance potential and burnup characteristics of a novel natural uranium fueled nuclear research reactor [Annals of Nuclear Energy 31(12), 1331–1356]. It was found that the proposed core, which provides twice the flux per unit core power compared to similar natural uranium fueled, light water cooled, heavy water
Accuracy considerations for Chebyshev rational approximation method (CRAM) in Burnup calculations
Pusa, M.
2013-07-01
The burnup equations can in principle be solved by computing the exponential of the burnup matrix. However, due to the difficult numerical characteristics of burnup matrices, the problem is extremely stiff and the matrix exponential solution has previously been considered infeasible for an entire burnup system containing over a thousand nuclides. It was recently discovered by the author that the eigenvalues of burnup matrices are generally located near the negative real axis, which prompted introducing the Chebyshev rational approximation method (CRAM) for solving the burnup equations. CRAM can be characterized as the best rational approximation on the negative real axis and it has been shown to be capable of simultaneously solving an entire burnup system both accurately and efficiently. In this paper, the accuracy of CRAM is further studied in the context of burnup equations. The approximation error is analyzed based on the eigenvalue decomposition of the burnup matrix. It is deduced that the relative accuracy of CRAM may be compromised if a nuclide concentration diminishes significantly during the considered time step. Numerical results are presented for two test cases, the first one representing a small burnup system with 36 nuclides and the second one a full a decay system with 1531 nuclides. (authors)
NASA Astrophysics Data System (ADS)
Aufiero, M.; Cammi, A.; Fiorina, C.; Leppänen, J.; Luzzi, L.; Ricotti, M. E.
2013-10-01
In this work, the Monte Carlo burn-up code SERPENT-2 has been extended and employed to study the material isotopic evolution of the Molten Salt Fast Reactor (MSFR). This promising GEN-IV nuclear reactor concept features peculiar characteristics such as the on-line fuel reprocessing, which prevents the use of commonly available burn-up codes. Besides, the presence of circulating nuclear fuel and radioactive streams from the core to the reprocessing plant requires a precise knowledge of the fuel isotopic composition during the plant operation. The developed extension of SERPENT-2 directly takes into account the effects of on-line fuel reprocessing on burn-up calculations and features a reactivity control algorithm. It is here assessed against a dedicated version of the deterministic ERANOS-based EQL3D procedure (PSI-Switzerland) and adopted to analyze the MSFR fuel salt isotopic evolution. Particular attention is devoted to study the effects of reprocessing time constants and efficiencies on the conversion ratio and the molar concentration of elements relevant for solubility issues (e.g., trivalent actinides and lanthanides). Quantities of interest for fuel handling and safety issues are investigated, including decay heat and activities of hazardous isotopes (neutron and high energy gamma emitters) in the core and in the reprocessing stream. The radiotoxicity generation is also analyzed for the MSFR nominal conditions. The production of helium and the depletion in tungsten content due to nuclear reactions are calculated for the nickel-based alloy selected as reactor structural material of the MSFR. These preliminary evaluations can be helpful in studying the radiation damage of both the primary salt container and the axial reflectors.
Propagation of Nuclear Data Uncertainties for ELECTRA Burn-up Calculations
NASA Astrophysics Data System (ADS)
Sjöstrand, H.; Alhassan, E.; Duan, J.; Gustavsson, C.; Koning, A. J.; Pomp, S.; Rochman, D.; Österlund, M.
2014-04-01
The European Lead-Cooled Training Reactor (ELECTRA) has been proposed as a training reactor for fast systems within the Swedish nuclear program. It is a low-power fast reactor cooled by pure liquid lead. In this work, we propagate the uncertainties in 239Pu transport data to uncertainties in the fuel inventory of ELECTRA during the reactor lifetime using the Total Monte Carlo approach (TMC). Within the TENDL project, nuclear models input parameters were randomized within their uncertainties and 740 239Pu nuclear data libraries were generated. These libraries are used as inputs to reactor codes, in our case SERPENT, to perform uncertainty analysis of nuclear reactor inventory during burn-up. The uncertainty in the inventory determines uncertainties in: the long-term radio-toxicity, the decay heat, the evolution of reactivity parameters, gas pressure and volatile fission product content. In this work, a methodology called fast TMC is utilized, which reduces the overall calculation time. The uncertainty of some minor actinides were observed to be rather large and therefore their impact on multiple recycling should be investigated further. It was also found that, criticality benchmarks can be used to reduce inventory uncertainties due to nuclear data. Further studies are needed to include fission yield uncertainties, more isotopes, and a larger set of benchmarks.
Holly R. Trellue
1998-12-01
Monteburns is a fully automated tool that links the Monte Carlo transport code MCNP with the radioactive decay and burnup code 0RIGEN2. Monteburns produces many criticality and burnup computational parameters based on material feed/removal specifications, power(s), and time intervals. This code processes input from the user indicating the system geometry, initial material compositions, feed/removal, and other code-specific parameters. Results from MCNP, 0RIGEN2, and other calculations are then output successively as the code runs. The principle function of monteburns is to first transfer one-group cross sections and fluxes from MCNP to 0RIGEN2, and then transfer the resulting material compositions (after irradiation and/or decay) from 0RIGEN2 back to MCNP in a repeated, cyclic fashion. The main requirement of the code is that the user have a working MCNP input file and other input parameters; all interaction with 0RIGEN2 and other calculations are performed by monteburns. This report presents the results obtained from the benchmarking of monteburns to measured and previously obtained data from traditional Light Water Reactor systems. The majority of the differences seen between the two were less than five percent. These were primarily a result of variances in cross sections between MCNP, cross section libraries used by other codes, and observed values. With this understanding, this code can now be used with confidence for burnup calculations in three-dimensional systems. It was designed for use in the Accelerator Transmutation of Waste project at Los Alamos National Laboratory but is also being applied to the analysis of isotopic production/destruction of transuranic actinides in a reactor system. The code has now been shown to sufficiently support these calculations.
IN-CORE FUEL MANAGEMENT: PWR Core Calculations Using MCRAC
NASA Astrophysics Data System (ADS)
Petrovi?, B. G.
1991-01-01
The following sections are included: * INTRODUCTION * IN-CORE FUEL MANAGEMENT CALCULATIONS * In-Core Fuel Management * Methodological Problems of In-Core Fuel Management * In-Core Fuel Management Analytical Tools * PENN STATE FUEL MANAGEMENT PACKAGE * Penn State Fuel Management Package (PFMP) * Assembly Data Description (ADD) * Linking PSU-LEOPARD and MCRAC: An Example * MULTICYCLE REACTOR ANALYSIS CODE (MCRAC) * Main Features and Options of MCRAC code * Core geometry * Diffusion equations * 1.5-group model * Multicycle neutronic analysis * Multicycle cost analysis * Criticality search * Power-dependent xenon feedback calculations * Control rod and burnable absorber simulation * Search for LP with flat BOC power distribution * Artificial ADD option * Variable dimensioning technique * RBI version of MCRAC code * Programming changes in PC version * Fuel interchange option * MCRAC Input/Output * General input description * Sample input * Sample output * EXPERIENCE WITH MCRAC CODE * CONCLUSIONS * REFERENCES
Using ORIGEN/KENO to calculate burnup credit for spent-fuel pool criticality analyses
Rombough, C.T.; Martonak, S.H.; Walkin, J.
1994-12-31
Due to delays in the U.S. Department of Energy high-level waste storage program, the spent-fuel pool at the Rancho Seco nuclear power plant was reracked in 1985 so that nearly twice as many fuel assemblies could be stored. Since the fuel assemblies would be much closer together, the racks were impregnated with Boraflex, a compound containing boron to absorb neutrons and reduce the effective multiplication constant, k{sub eff}. In 1993, the U.S. Nuclear Regulatory Commission issued a letter to all licensees using Boraflex to advise them that under certain conditions, the Boraflex absorber may deteriorate and may not provide the necessary negative reactivity. Since the Rancho Seco spent-fuel pool contained only burned fuel assemblies and there were no plans to store any fresh fuel assemblies, it was possible that the pool would remain subcritical if the increase in multiplication caused by the loss of Boraflex were offset by the decrease in multiplication caused by the burnup in the fuel. Consequently, in early 1994, a detailed criticality analysis was performed to accurately account for the actual burnup in the fuel to see if the spent-fuel pool would remain subcritical in the unlikely event that all of the Boraflex absorber were lost.
Fluence-limited burnup as a function of fast reactor core parameters
Kersting, Alyssa (Alyssa Rae)
2011-01-01
The limiting factor in current designs for fast reactors is not only the reactivity, but also the maximum permissible fast-neutron fluence in the cladding, especially for reduced uranium enrichment cores using high-albedo ...
Borodkin, P.G.; Borodkin, G.I.; Khrennikov, N.N.
2011-07-01
This paper deals with calculated and semi-analytical evaluations of VVER-1000 reactor core neutron source distributions and their influence on measurements and calculations of the integral through-vessel neutron leakage. Neutron activation measurements analyzed in the paper were carried out in an ex-vessel air cavity at different nuclear power plant units with VVER-1000 during different fuel cycles. The time-integrated neutron source distributions used for DORT calculations were prepared via two different approaches based on (a) calculated fuel burnup (standard routine procedure) and (b) in-core measurements by means of self-powered detectors (SPDs) and thermocouples (TCs) (new approach). Considering that fuel burnup distributions in operating VVER may be evaluated now by the use of analytical methods (calculations) only, it is necessary to develop new approaches for the testing and correction of calculated evaluations of a neutron source. The results presented in this paper allow one to consider the reverse task of the alternative estimation of fuel burnup distributions. The proposed approach is based on the adjustment (fitting) of time-integrated neutron source distributions, and thus fuel burnup patterns, in some part of the reactor core, taking into account neutron leakage measurements, neutron-physical calculations, and in-core SPD and TC measurement data. (authors)
NASA Astrophysics Data System (ADS)
Sabouri, P.; Bidaud, A.; Dabiran, S.; Lecarpentier, D.; Ferragut, F.
2014-04-01
The development of tools for nuclear data uncertainty propagation in lattice calculations are presented. The Total Monte Carlo method and the Generalized Perturbation Theory method are used with the code DRAGON to allow propagation of nuclear data uncertainties in transport calculations. Both methods begin the propagation of uncertainties at the most elementary level of the transport calculation - the Evaluated Nuclear Data File. The developed tools are applied to provide estimates for response uncertainties of a PWR cell as a function of burnup.
ORPHEE research reactor: 3D core depletion calculation using Monte-Carlo code TRIPOLI-4®
NASA Astrophysics Data System (ADS)
Damian, F.; Brun, E.
2014-06-01
ORPHEE is a research reactor located at CEA Saclay. It aims at producing neutron beams for experiments. This is a pool-type reactor (heavy water), and the core is cooled by light water. Its thermal power is 14 MW. ORPHEE core is 90 cm height and has a cross section of 27x27 cm2. It is loaded with eight fuel assemblies characterized by a various number of fuel plates. The fuel plate is composed of aluminium and High Enriched Uranium (HEU). It is a once through core with a fuel cycle length of approximately 100 Equivalent Full Power Days (EFPD) and with a maximum burnup of 40%. Various analyses under progress at CEA concern the determination of the core neutronic parameters during irradiation. Taking into consideration the geometrical complexity of the core and the quasi absence of thermal feedback for nominal operation, the 3D core depletion calculations are performed using the Monte-Carlo code TRIPOLI-4® [1,2,3]. A preliminary validation of the depletion calculation was performed on a 2D core configuration by comparison with the deterministic transport code APOLLO2 [4]. The analysis showed the reliability of TRIPOLI-4® to calculate a complex core configuration using a large number of depleting regions with a high level of confidence.
G. A. Berna; G. A. Beyer; K. L. Davis; D. D. Lanning
1997-01-01
FRAPCON-3 is a FORTRAN IV computer code that calculates the steady-state response of light water reactor fuel rods during long-term burnup. The code calculates the temperature, pressure, and deformation of a fuel rod as functions of time-dependent fuel rod power and coolant boundary conditions. The phenomena modeled by the code include (1) heat conduction through the fuel and cladding, (2)
Appropriate burnup measurements for transportation burnup credit
Lancaster, D.; Fuentes, E.
1997-04-01
This paper addresses two of the measurement specifications used in analyzing spent fuel packages to gain burnup credit. The philosophy and calculation of rejection criteria and measurement accuracy are discussed. Any assembly for which the declared measured value and reactor record value deviate by more than 10% will be rejected. Measurement accuracy requirements are established for dependent and independent systems. The requirements have been tested and are achievable, ensuring safe operation without extra cost. 6 refs.
Core Burnup Calculation of Plutonium Burning Inert-Matrix Fueled High Temperature Gas Cooled Reactor
Hiroshi AKIE
2007-01-01
For the efficient reduction of excess plutonium amount, Japan Atomic Energy Research Institute (JAERI, now Japan Atomic Energy Agency) has studied a concept of rock-like oxide (ROX) fuel as a kind of inert matrix fuel (IMF). In the JAERI study, ROX fuel is burnt in existing light water reactors (LWRs), while in this study, pebble bed type high temperature gas
Burnup simulations of an inert matrix fuel using a two region, multigroup reactor physics model
Schneider, E. [Dept. of Mechanical Engineering, Univ. of Texas at Austin, 1 Univ. Place C2200, Austin, TX 78712 (United States); Deinert, M.; Bingham Cady, K. [Dept. of Theoretical and Applied Mechanics, Cornell Univ., Ithaca, NY 14853 (United States)
2006-07-01
Determining the time dependent concentration of isotopes in a nuclear reactor core is of fundamental importance to analysis of nuclear fuel cycles and the impact of spent fuels on long term storage facilities. We present a fast, conceptually simple tool for performing burnup calculations applicable to obtaining isotopic balances as a function of fuel burnup. The code (VBUDS: visualization, burnup, depletion and spectra) uses a two region, multigroup collision probability model to determine the energy dependent neutron flux and tracks the buildup and burnout of 24 actinides, as well as fission products. The model has been tested against benchmarked results for LWRs burning UOX and MOX, as well as MONTEBURNS simulations of zirconium oxide based IMF, all with strong fidelity. As an illustrative example, VBUDS burnup calculation results for an IMF fuel are presented in this paper. (authors)
Burnup and neutron economy of accelerator-driven reactors
Takahashi, H.; Yang, W.; An, Y.; Yamazaki, Y. [Brookhaven National Lab., Upton, NY (United States)
1997-12-01
This paper presents an evaluation of fuel burnup in an accelerator driven prototype subcritical reactor in which the fertile material is uranium or thorium and cooling is by sodium or lead. Burnup was calculated by the Monte Carlo method.
PWR AXIAL BURNUP PROFILE ANALYSIS
J.M. Acaglione
2003-09-17
The purpose of this activity is to develop a representative ''limiting'' axial burnup profile for pressurized water reactors (PWRs), which would encompass the isotopic axial variations caused by different assembly irradiation histories, and produce conservative isotopics with respect to criticality. The effect that the low burnup regions near the ends of spent fuel have on system reactivity is termed the ''end-effect''. This calculation will quantify the end-effects associated with Pressurized Water Reactor (PWR) fuel assemblies emplaced in a hypothetical 21 PWR waste package. The scope of this calculation covers an initial enrichment range of 3.0 through 5.0 wt% U-235 and a burnup range of 10 through 50 GWd/MTU. This activity supports the validation of the process for ensuring conservative generation of spent fuel isotopics with respect to criticality safety applications, and the use of burnup credit for commercial spent nuclear fuel. The intended use of these results will be in the development of PWR waste package loading curves, and applications involving burnup credit. Limitations of this evaluation are that the limiting profiles are only confirmed for use with the B&W 15 x 15 fuel assembly design. However, this assembly design is considered bounding of all other typical commercial PWR fuel assembly designs. This calculation is subject to the Quality Assurance Requirements and Description (QARD) because this activity supports investigations of items or barriers on the Q-list (YMP 2001).
Dose Rate Calculations for Rotary Mode Core Sampling Exhauster
FOUST, D.J.
2000-10-26
This document provides the calculated estimated dose rates for three external locations on the Rotary Mode Core Sampling (RMCS) exhauster HEPA filter housing, per the request of Characterization Field Engineering.
Burnup credit for storage and transportation casks
Wells, A.H.
1988-01-01
The application of burnup credit to storage and transportation cask licensing results in a significant improvement in cask capacity and an associated reduction of the cost per kilogram of uranium in the cask contents. The issues for licensing with burnup credit deal primarily with the treatment of fission product poisons and methods of verification of burnup during cask operations. Other issues include benchmarking of cross-section sets and codes and the effect of spatial variation of burnup within an assembly. The licensing of burnup credit for casks will be complex, although the criticality calculations are not themselves difficult. Attention should be directed to the use of fission product poisons and the uncertainties that they introduce. Verification of burnup by measurements will remove some of the concerns for criticality safety. Calculations for burnup credit casks should consider rod-to-rod and axial variations of burnup, as well as variability of burnable poisons it they are used in the assembly. In spite of the complexity of cask burnup credit licensing issues, these issues appear to be resolvable within the current state of the art of criticality safety.
Core Physics and Kinetics Calculations for the Fissioning Plasma Core Reactor
NASA Technical Reports Server (NTRS)
Butler, C.; Albright, D.
2007-01-01
Highly efficient, compact nuclear reactors would provide high specific impulse spacecraft propulsion. This analysis and numerical simulation effort has focused on the technical feasibility issues related to the nuclear design characteristics of a novel reactor design. The Fissioning Plasma Core Reactor (FPCR) is a shockwave-driven gaseous-core nuclear reactor, which uses Magneto Hydrodynamic effects to generate electric power to be used for propulsion. The nuclear design of the system depends on two major calculations: core physics calculations and kinetics calculations. Presently, core physics calculations have concentrated on the use of the MCNP4C code. However, initial results from other codes such as COMBINE/VENTURE and SCALE4a. are also shown. Several significant modifications were made to the ISR-developed QCALC1 kinetics analysis code. These modifications include testing the state of the core materials, an improvement to the calculation of the material properties of the core, the addition of an adiabatic core temperature model and improvement of the first order reactivity correction model. The accuracy of these modifications has been verified, and the accuracy of the point-core kinetics model used by the QCALC1 code has also been validated. Previously calculated kinetics results for the FPCR were described in the ISR report, "QCALC1: A code for FPCR Kinetics Model Feasibility Analysis" dated June 1, 2002.
Bai, D.; Levine, S.L. (Pennsylvania State Univ., University Park (United States)); Luoma, J.; Mahgerefteh, M. (GPU Nuclear Corp., Parsippany, NJ (United States))
1992-01-01
The Three Mile Island unit 1 core reloads have been designed using fast but accurate scoping codes, PSUI-LEOPARD and ADMARC. PSUI-LEOPARD has been normalized to EPRI-CPM2 results and used to calculate the two-group constants, whereas ADMARC is a modern two-dimensional, two-group diffusion theory nodal code. Problems in accuracy were encountered for cycles 8 and higher as the core lifetime was increased beyond 500 effective full-power days. This is because the heavier loaded cores in both {sup 235}U and {sup 10}B have harder neutron spectra, which produces a change in the transport effect in the baffle reflector region, and the burnable poison (BP) simulations were not accurate enough for the cores containing the increased amount of {sup 10}B required in the BP rods. In the authors study, a technique has been developed to take into account the change in the transport effect in the baffle region by modifying the fast neutron diffusion coefficient as a function of cycle length and core exposure or burnup. A more accurate BP simulation method is also developed, using integral transport theory and CPM2 data, to calculate the BP contribution to the equivalent fuel assembly (supercell) two-group constants. The net result is that the accuracy of the scoping codes is as good as that produced by CASMO/SIMULATE or CPM2/SIMULATE when comparing with measured data.
NEPHTIS: Core depletion validation relying on 2D transport core calculations with the APOLLO2 code
Damian, F.; Raepsaet, X.; Groizard, M.; Poinot, C. [DEN/DM2S/SERMA/LCA, CEA Saclay, 91191 Gif-sur-Yvette Cedex (France)
2006-07-01
The CEA, in collaboration with EDF and AREVA-NP, is developing a core modelling tool called NEPHTIS, for Neutronic Process for HTGR Innovating Systems and dedicated at present day to the prismatic block-type HTGR (High Temperature Gas-Cooled Reactors). Due to the lack of usable HTGR experimental results, the confidence in this neutronic computational tool relies essentially on comparisons to reference or best-estimate calculations. In the present analysis, the Aleppo deterministic transport code has been selected as reference for validating core depletion simulations carried out within NEPHTIS. These reference calculations were performed on fully detailed 2D core configurations using the Method of Characteristics. The latter has been validated versus Monte Carlo method for different static core configurations [1], [2] and [3]. All the presented results come from an annular HTGR core loaded with uranium-based fuel (15% enrichment). During the core depletion validation, reactivity, reaction rates distributions and nuclei concentrations have been compared. In addition, the impact of various physical and geometrical parameters such as the core loading (one-through or batch-wise reloading) and the amount of burnable poison has been investigated during the validation phases. The results confirm that NEPHTIS is able to predict the core reactivity with uncertainties of {+-}350 pcm. At the end of the core irradiation, the U-235 consumption is calculated within {+-} 0, 7 % while the plutonium mass discharged from the core is calculated within {+-}1 %. As far as the core power distributions are concerned, small discrepancies ( and < 2.3 %) can be observed on the fuel block-averaged power distribution in the core. (authors)
Berna, G.A.; Beyer, G.A.; Davis, K.L.; Lanning, D.D.
1997-12-01
FRAPCON-3 is a FORTRAN IV computer code that calculates the steady-state response of light water reactor fuel rods during long-term burnup. The code calculates the temperature, pressure, and deformation of a fuel rod as functions of time-dependent fuel rod power and coolant boundary conditions. The phenomena modeled by the code include (1) heat conduction through the fuel and cladding, (2) cladding elastic and plastic deformation, (3) fuel-cladding mechanical interaction, (4) fission gas release, (5) fuel rod internal gas pressure, (6) heat transfer between fuel and cladding, (7) cladding oxidation, and (8) heat transfer from cladding to coolant. The code contains necessary material properties, water properties, and heat-transfer correlations. The codes` integral predictions of mechanical behavior have not been assessed against a data base, e.g., cladding strain or failure data. Therefore, it is recommended that the code not be used for analyses of cladding stress or strain. FRAPCON-3 is programmed for use on both mainframe computers and UNIX-based workstations such as DEC 5000 or SUN Sparcstation 10. It is also programmed for personal computers with FORTRAN compiler software and at least 8 to 10 megabytes of random access memory (RAM). The FRAPCON-3 code is designed to generate initial conditions for transient fuel rod analysis by the FRAPTRAN computer code (formerly named FRAP-T6).
TOPICAL REPORT ON ACTINIDE-ONLY BURNUP CREDIT FOR PWR SPENT NUCLEAR FUEL PACKAGES
DOE
1997-04-01
A methodology for performing and applying nuclear criticality safety calculations, for PWR spent nuclear fuel (SNF) packages with actinide-only burnup credit, is described. The changes in the U-234, U-235, U-236, U-238, Pu-238, Pu-239, Pu-240, Pu-241, Pu-242, and Am-241 concentration with burnup are used in burnup credit criticality analyses. No credit for fission product neutron absorbers is taken. The methodology consists of five major steps. (1) Validate a computer code system to calculate isotopic concentrations of SNF created during burnup in the reactor core and subsequent decay. A set of chemical assay benchmarks is presented for this purpose as well as a method for assessing the calculational bias and uncertainty, and conservative correction factors for each isotope. (2) Validate a computer code system to predict the subcritical multiplication factor, k{sub eff}, of a spent nuclear fuel package. Fifty-seven UO{sub 2}, UO{sub 2}/Gd{sub 2}O{sub 3}, and UO{sub 2}/PuO{sub 2} critical experiments have been selected to cover anticipated conditions of SNF. The method uses an upper safety limit on k{sub eff} (which can be a function of the trending parameters) such that the biased k{sub eff}, when increased for the uncertainty is less than 0.95. (3) Establish bounding conditions for the isotopic concentration and criticality calculations. Three bounding axial profiles have been established to assure the ''end effect'' is accounted for conservatively. (4) Use the validated codes and bounding conditions to generate package loading criteria (burnup credit loading curves). Burnup credit loading curves show the minimum burnup required for a given initial enrichment. The utility burnup record is compared to this requirement after the utility accounts for the uncertainty in its record. Separate curves may be generated for each assembly design, various minimum cooling times and burnable absorber histories. (5) Verify that SNF assemblies meet the package loading criteria and confirm proper assembly selection prior to loading. A measurement of the average assembly burnup is required and that measurement must be within 10% of the utility burnup record for the assembly to be accepted. The measurement device must be accurate to within 10%. Each step is described in detail for use with any computer code system and is then demonstrated with the SCALE 4.2 computer code package using 27BURNUPLIB cross sections.
Effect of Fuel Fraction on Small Modified CANDLE Burn-up Based Gas Cooled Fast Reactors
NASA Astrophysics Data System (ADS)
Ariani, Menik; Su'ud, Zaki; Waris, Abdul; Khairurrijal, Asiah, Nur; Shafii, M. Ali
2010-12-01
A conceptual design study of Gas Cooled Fast Reactors with Modified CANDLE Burn-up has been performed. The objective of this research is to get optimal design parameters of such type reactors. The parameters of nuclear design including the critical condition, conversion ratio, and burn-up level were compared. These parameters are calculated by variation in the fuel fraction 47.5% up to 70%. Two dimensional full core multi groups diffusion calculations was performed by CITATION code. Group constant preparations are performed by using SRAC code system with JENDL-3.2 nuclear data library. In this design the reactor cores with cylindrical cell two dimensional R-Z core models are subdivided into several parts with the same volume in the axial directions. The placement of fuel in core arranged so that the result of plutonium from natural uranium can be utilized optimally for 10 years reactor operation. Modified CANDLE burn-up was established successfully in a core radial width 1.4 m. Total thermal power output for reference core is 550 MW. Study on the effect of fuel to coolant ratio shows that effective multiplication factor (keff) is in almost linear relations with the change of the fuel volume to coolant ratio.
Effect of Fuel Fraction on Small Modified CANDLE Burn-up Based Gas Cooled Fast Reactors
Ariani, Menik; Su'ud, Zaki; Waris, Abdul; Asiah, Nur; Shafii, M. Ali; Khairurrijal
2010-12-23
A conceptual design study of Gas Cooled Fast Reactors with Modified CANDLE Burn-up has been performed. The objective of this research is to get optimal design parameters of such type reactors. The parameters of nuclear design including the critical condition, conversion ratio, and burn-up level were compared. These parameters are calculated by variation in the fuel fraction 47.5% up to 70%. Two dimensional full core multi groups diffusion calculations was performed by CITATION code. Group constant preparations are performed by using SRAC code system with JENDL-3.2 nuclear data library. In this design the reactor cores with cylindrical cell two dimensional R-Z core models are subdivided into several parts with the same volume in the axial directions. The placement of fuel in core arranged so that the result of plutonium from natural uranium can be utilized optimally for 10 years reactor operation. Modified CANDLE burn-up was established successfully in a core radial width 1.4 m. Total thermal power output for reference core is 550 MW. Study on the effect of fuel to coolant ratio shows that effective multiplication factor (k{sub eff}) is in almost linear relations with the change of the fuel volume to coolant ratio.
Value of burnup credit beyond actinides
Lancaster, D.; Fuentes, E.; Kang, Chi
1997-12-01
DOE has submitted a topical report to the NRC justifying burnup credit based only on actinide isotopes (U-234, U-235, U-236, U-238, Pu-238, Pu-239, Pu-240, Pu-241, Pu-242, and Am-241). When this topical report is approved, it will allow a great deal of the commercial spent nuclear fuel to be transported in significantly higher capacity casks. A cost savings estimate for shipping fuel in 32 assembly (burnup credit) casks as opposed to 24 assembly (non-burnup credit) casks was previously presented. Since that time, more detailed calculations have been performed using the methodology presented in the Actinide-Only Burnup Credit Topical Report. Loading curves for derated casks have been generated using actinide-only burnup credit and are presented in this paper. The estimates of cost savings due to burnup credit for shipping fuel utilizing 32, 30, 28, and 24 assembly casks where only the 24 assembly cask does not burnup credit have been created and are discussed. 4 refs., 2 figs.
AFEN-polynomial nodal method for burnup gradient correction
Nam Zin Cho; Do Sam Kim
1996-12-31
In general, neglecting the large intranodal cross-section gradients induced by depletion and feedback causes a nodal method to lose its accuracy in predicting nodal unknowns acceptably. Recently, in context of the analytic function expansion nodal (AFEN) method, Noh and Cho developed a rehomogenization burnup correction model that homogenizes the burnt fuel assembly using equivalence theory. They expanded the fluxes of a burnt fuel assembly in terms of quartic polynomial functions with nonzero current boundary conditions obtained from previous feedback iterations. The fluxes arc used to get additional flux-volume-weighted cross sections and discontinuity factors of the assembly. In this paper, we develop an AFEN-polynomial correction model that directly solves the nodal diffusion equation with spatially varying cross sections due to burnup gradient. In this model, we retain the original analytic basis functions that are derived from the diffusion equation containing only volume-averaged cross sections (burnup dependent), and we add second-order polynomial correction terms in the expansion. Because this model solves the whole core problem without any assumption of nodal quantity and rehomogenization, its accuracy is not affected by the previous feedback calculation.
Randall O. Gauntt; Dana Auburn Powers; Scott G. Ashbaugh; Mark Thomas Leonard; Pamela Longmire
2010-01-01
In this study, risk-significant pressurized-water reactor severe accident sequences are examined using MELCOR 1.8.5 to explore the range of fission product releases to the reactor containment building. Advances in the understanding of fission product release and transport behavior and severe accident progression are used to render best estimate analyses of selected accident sequences. Particular emphasis is placed on estimating the
Recent Developments in No-Core Shell-Model Calculations
Navratil, P; Quaglioni, S; Stetcu, I; Barrett, B R
2009-03-20
We present an overview of recent results and developments of the no-core shell model (NCSM), an ab initio approach to the nuclear many-body problem for light nuclei. In this aproach, we start from realistic two-nucleon or two- plus three-nucleon interactions. Many-body calculations are performed using a finite harmonic-oscillator (HO) basis. To facilitate convergence for realistic inter-nucleon interactions that generate strong short-range correlations, we derive effective interactions by unitary transformations that are tailored to the HO basis truncation. For soft realistic interactions this might not be necessary. If that is the case, the NCSM calculations are variational. In either case, the ab initio NCSM preserves translational invariance of the nuclear many-body problem. In this review, we, in particular, highlight results obtained with the chiral two- plus three-nucleon interactions. We discuss efforts to extend the applicability of the NCSM to heavier nuclei and larger model spaces using importance-truncation schemes and/or use of effective interactions with a core. We outline an extension of the ab initio NCSM to the description of nuclear reactions by the resonating group method technique. A future direction of the approach, the ab initio NCSM with continuum, which will provide a complete description of nuclei as open systems with coupling of bound and continuum states is given in the concluding part of the review.
Yasunori Ohoka; Ismile; Hiroshi Sekimoto
2004-07-01
The CANDLE burnup strategy is a new reactor burnup concept, where the distributions of fuel nuclide densities, neutron flux, and power density move with the same constant speed along the core axis from bottom to top or from top to bottom of the core and without any change in their shapes. It can be applied easily to the block-type high temperature gas cooled reactor using an appropriate burnable poison mixed with uranium oxide fuel. In the present study, the burnup distribution and the temperature distribution in the core are investigated and their effects on the CANDLE burnup core characteristics are studied. In this study, the natural gadolinium is used as the burnable poison. With the fuel enrichment of 15%, the natural gadolinium concentration of 3.0% and the fuel pin pitch of 6.6 cm, the CANDLE burnup is realized with the burning region moving speed of 29 cm/year and the axial half width of power density distribution of 1.5 m for uniform group constant case at 900 K. When the effect of nuclide change by burnup is considered, the burning region speed becomes 25 cm/year and the axial half-width of power density distribution becomes 1.25 m. When the temperature distributions effect is considered, the effects on the core characteristics are smaller than the burnup distribution effect. The maximum fuel temperature of the parallel flow case is higher than the counter flow case. (authors)
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.
Isotopic validation for PWR actinide-only burnup credit using Yankee Rowe data
NONE
1997-11-01
Safety analyses of criticality control systems for transportation packages include an assumption that the spent nuclear fuel (SNF) loaded into the package is fresh or unirradiated. In other words, the spent fuel is assumed to have its original, as-manufactured U-235 isotopic content. The ``fresh fuel`` assumption is very conservative since the potential reactivity of the nuclear fuel is substantially reduced after being irradiated in the reactor core. The concept of taking credit for this reduction in nuclear fuel reactivity due to burnup of the fuel, instead of using the fresh fuel assumption in the criticality safety analysis, is referred to as ``Burnup Credit.`` Burnup credit uses the actual physical composition of the fuel and accounts for the net reduction of fissile material and the buildup of neutron absorbers in the fuel as it is irradiated. Neutron absorbers include actinides and other isotopes generated as a result of the fission process. Using only the change in actinide isotopes in the burnup credit criticality analysis is referred to as ``Actinide-Only Burnup Credit.`` The use of burnup credit in the design of criticality control systems enables more spent fuel to be placed in a package. Increased package capacity results in a reduced number of storage, shipping and disposal containers for a given number of SNF assemblies. Fewer shipments result in a lower risk of accidents associated with the handling and transportation of spent fuel, thus reducing both radiological and nonradiological risk to the public. This paper describes the modeling and the results of comparison between measured and calculated isotopic inventories for a selected number of samples taken from a Yankee Rowe spent fuel assembly.
NASA Astrophysics Data System (ADS)
Sloma, Tanya Noel
When representing the behavior of commercial spent nuclear fuel (SNF), credit is sought for the reduced reactivity associated with the net depletion of fissile isotopes and the creation of neutron-absorbing isotopes, a process that begins when a commercial nuclear reactor is first operated at power. Burnup credit accounts for the reduced reactivity potential of a fuel assembly and varies with the fuel burnup, cooling time, and the initial enrichment of fissile material in the fuel. With regard to long-term SNF disposal and transportation, tremendous benefits, such as increased capacity, flexibility of design and system operations, and reduced overall costs, provide an incentive to seek burnup credit for criticality safety evaluations. The Nuclear Regulatory Commission issued Interim Staff Guidance 8, Revision 2 in 2002, endorsing burnup credit of actinide composition changes only; credit due to actinides encompasses approximately 30% of exiting pressurized water reactor SNF inventory and could potentially be increased to 90% if fission product credit were accepted. However, one significant issue for utilizing full burnup credit, compensating for actinide and fission product composition changes, is establishing a set of depletion parameters that produce an adequately conservative representation of the fuel's isotopic inventory. Depletion parameters can have a significant effect on the isotopic inventory of the fuel, and thus the residual reactivity. This research seeks to quantify the reactivity impact on a system from dominant depletion parameters (i.e., fuel temperature, moderator density, burnable poison rod, burnable poison rod history, and soluble boron concentration). Bounding depletion parameters were developed by statistical evaluation of a database containing reactor operating histories. The database was generated from summary reports of commercial reactor criticality data. Through depletion calculations, utilizing the SCALE 6 code package, several light water reactor assembly designs and in-core locations are analyzed in establishing a combination of depletion parameters that conservatively represent the fuel's isotopic inventory as an initiative to take credit for fuel burnup in criticality safety evaluations for transportation and storage of SNF.
Nuclide analysis in high burnup fuel samples irradiated in Vandellós 2
NASA Astrophysics Data System (ADS)
Zwicky, H. U.; Low, J.; Granfors, M.; Alejano, C.; Conde, J. M.; Casado, C.; Sabater, J.; Lloret, M.; Quecedo, M.; Gago, J. A.
2010-07-01
In the framework of a high burnup fuel demonstration programme, rods with an enrichment of 4.5% 235U were operated to a rod average burnup of about 70 MWd/kgU in the Spanish Vandellós 2 pressurised water reactor. The rods were sent to hot cells and used for different research projects. This paper describes the isotopic composition measurements performed on samples of those rods, and the analysis of the measurement results based on comparison against calculated values. The fraction and composition of fission gases released to the rod free volume was determined for two of the rods. About 8% of Kr and Xe produced by fission were released. From the isotopic composition of the gases, it could be concluded that the gases were not preferentially released from the peripheral part of the fuel column. Local burnup and isotopic content of gamma emitting nuclides were determined by quantitatively evaluating axial gamma scans of the full rods. Nine samples were cut at different axial levels from three of the rods and analysed in two campaigns. More than 50 isotopes of 16 different elements were assessed, most of them by Inductively Coupled Plasma Mass Spectrometry after separation with High Performance Liquid Chromatography. In general, these over 400 data points gave a consistent picture of the isotopic content of irradiated fuel as a function of burnup. Only in a few cases, the analysis provided unexpected results that seem to be wrong, in most cases due to unidentified reasons. Sample burnup analysis was performed by comparing experimental isotopic abundances of uranium and plutonium composition as well as neodymium isotopic concentrations with corresponding CASMO based data. The results were in agreement with values derived independently from gamma scanning and from core design data and plant operating records. Measured isotope abundances were finally assessed using the industry standard SAS2H sequence of the SCALE code system. This exercise showed good agreement between measured and calculated values for most of the analysed isotopes, similar to those reported previously for lower burnup ranges. Thus, it could be concluded, that SAS2H results for high burnup samples are not subject to higher uncertainty and/or different biases than for lower burnup samples, and that the different isotopic experimental measurement methods provide accurate results with acceptable precision.
Inherent safety of minimum-burnup breed and burn reactors
Qvist, S.; Reenspan, E. [Dept. of Nuclear Engineering, Univ. of California, Berkeley, CA 94720-1730 (United States)
2012-07-01
Reactors that aim to sustain the breed and burn (B and B) mode of operation at minimum discharge burnup require excellent neutron economy, Minimum-burnup B and B cores are generally large and feature low neutron leakage probability and a hard neutron spectrum. While highly promising fuel cycles can be achieved with such designs, the very same features are pushing the limits of the core's ability to passively respond safely to unprotected accidents. Low leakage minimum-burnup sodium-cooled B and B cores have a large positive coolant void-worth and coolant temperature reactivity coefficient. In this study, the applicability of major approaches for fast reactor void-worth reduction is evaluated specifically for B and B cores. The design, shuffling scheme and performance of a new metallic-fueled, sodium-cooled minimum burnup B and B core, used as basis for the void-worth reduction analysis, is presented. The analysis shows that reactivity control systems based on passive {sup 6}Li injection during temperature excursions are the only option able to provide negative void-worth without significantly increasing the minimum burnup required for sustaining the B and B mode of operation. A new type of lithium expansion module (LEM) system was developed specifically for B and B cores and its effect on core performance is presented. (authors)
J. O. Barner; M. E. Cunningham; M. D. Freshley; D. D. Lanning
2010-01-01
This is the final report of the High Burnup Effects Program (HBEP). It has been prepared to present a summary, with conclusions, of the HBEP. The HBEP was an international, group-sponsored research program managed by Battelle, Pacific Northwest Laboratories (BNW). The principal objective of the HBEP was to obtain well-characterized data related to fission gas release (FGR) for light water
Advances in core loss calculations for magnetic materials
NASA Technical Reports Server (NTRS)
Triner, J. E.
1982-01-01
A new analytical technique which predicts the basic magnetic properties under various operating conditions encountered in state-of-the-art dc-ac/dc converters is discussed. Using a new flux-controlled core excitation circuit, magnetic core characteristics were developed for constant values of ramp flux (square wave voltage excitation) and frequency. From this empirical data, a mathematical loss characteristics equation is developed to analytically predict the specific core loss of several magnetic materials under various waveform excitation conditions. In addition, these characteristics show the circuit designer for the first time the direct functional relatonships between induction level and specific core loss as a function of the two key dc-dc converter operating parameters of input voltage and duty cycle.
Effect of core polarizability on photoionization cross-section calculations.
NASA Technical Reports Server (NTRS)
Kirkpatrick, R. C.
1972-01-01
Demonstration of the importance of core polarizability in a case where cancellation is only moderate, with suggestion of an improvement to the scaled Thomas-Fermi (STF) wave functions of Stewart and Rotenberg (1965). The inclusion of dipole polarizability of the core for argon is shown to substantially improve the agreement between the theoretical and experimental photoionization cross sections for the ground-state configuration.
Naruk, S.J.
1987-07-01
Minimum offset of 7 km across the Pinaleno Mountains metamorphic core complex is calculated by integrating the shear strains across the exposed width of the mylonite zone. The calculated displacement equals the offset on the associated detachment fault, estimated from offset marker beds. The method of determining displacement by strain integration may be directly applicable to many other metamorphic core complexes.
In-core and ex-core calculations of the VENUS simulated PWR benchmark experiment
Williams, M.L.; Chowdhury, P.; Landesman, M.; Kam, F.B.K.
1985-01-01
The VENUS PWR engineering mockup experiment was established to simulate a beginning-of-life, generic PWR configuration at the zero-power VENUS critical facility located at CEN/SCK, Mol, Belgium. The experimental measurement program consists of (1) gamma scans to determine the core power distribution, (2) in-core and ex-core foil activations, (3) neutron spectrometer measurements, and (4) gamma heating measurements with TLD's. Analysis of the VENUS benchmark has been performed with two-dimensional discrete ordinates transport theory, using the DOT-IV code.
New Soft-Core Potential Function for Molecular Dynamics Based Alchemical Free Energy Calculations
de Groot, Bert
New Soft-Core Potential Function for Molecular Dynamics Based Alchemical Free Energy Calculations accurate free energy calculations based on molecular dynamics simulations. A thermodynamic integration scheme is often used to calculate changes in the free energy of a system by integrating the change
High Burnup Fuel Behavior Modeling
M. Jahingir; R. Rand; R. Stachowski; B. Miles; K. Kusagaya
2007-01-01
This paper discusses the development and qualification of the PRIME03 code to address high burnup mechanisms and to improve uranium utilization in current and new reactor designs. Materials properties and behavioral models have been updated from previous thermal-mechanical codes to reflect the effects of burnup on fuel pellet thermal conductivity, Zircaloy creep, fuel pellet relocation, and fission gas release. These
The ORR Whole-Core LEU Fuel Demonstration
Bretscher, M.M.; Snelgrove, J.L.
1990-01-01
The ORR Whole-Core LEU Fuel Demonstration, conducted as part of the US Reduced Enrichment Research and Test Reactor Program, has been successfully completed. Using commercially-fabricated U{sub 3}Si{sub 2}-Al 20%-enriched fuel elements (4.8 g U/cc) and fuel followers (3.5 g U/cc), the 30-MW Oak Ridge Research Reactor was safely converted from an all-HEU core, through a series of HEU/LEU mixed transition cores, to an all-LEU core. There were no fuel element failures and average discharge burnups were measured to be as high as 50% for the standard elements and 75% for the fuel followers. Experimental results for burnup-dependent critical configurations, cycle-averaged fuel element powers, and fuel-element-averaged {sup 235}U burnups validated predictions based on three-dimensional depletion calculations. Calculated values for plutonium production and isotopic mass ratios as functions of {sup 235}U burnup support the corresponding measured quantities. In general, calculations for reaction rate distributions, control rod worths, prompt neutron decay constants, and isothermal temperature coefficients were found to agree with corresponding measured values. Experimentally determined critical configurations for fresh HEU and LEU cores radially reflected with water and with beryllium are well-predicted by both Monte Carlo and diffusion calculations. 17 refs.
/sup 18/O as a core plus two valence neutrons: A three-body Faddeev calculation
Ueta, K.; Miyake, H.; Mizukami, A.
1983-01-01
The nucleus /sup 18/O is studied assuming a three-body model: two neutrons outside an inert core of /sup 16/O: and solving the Faddeev equations. The calculated spectrum is in good agreement with experiment.
J. W. Sterbentz
1999-08-01
Six uranium isotopes and fourteen fission product isotopes were calculated on a mass basis at end-of-life (EOL) conditions for three fuel rods from different Light Water Breeder Reactor (LWBR) measurements. The three fuel rods evaluated here were taken from an LWBR seed module, a standard blanket module, and a reflector (Type IV) module. The calculated results were derived using a depletion methodology previously employed to evaluate many of the radionuclide inventories for spent nuclear fuels at the Idaho National Engineering and Environmental Laboratory. The primary goal of the calculational task was to further support the validation of this particular calculational methodology and its application to diverse reactor types and fuels. Result comparisons between the calculated and measured mass concentrations in the three rods indicate good agreement for the three major uranium isotopes (U-233, U-234, U-235) with differences of less than 20%. For the seed and standard blanket rod, the U-233 and U-234 differences were within 5% of the measured values (these two isotopes alone represent greater than 97% of the EOL total uranium mass). For the major krypton and xenon fission product isotopes, differences of less than 20% and less than 30% were observed, respectively. In general, good agreement was obtained for nearly all the measured isotopes. For these isotopes exhibiting significant differences, possible explanations are discussed in terms of measurement uncertainty, complex transmutations, etc.
Fuel-Cycle of 'CANDLE' Burnup with Depleted Uranium
Hiroshi, Sekimoto
2006-07-01
A new reactor burnup strategy CANDLE (Constant Axial shape of Neutron flux, nuclide densities and power shape During Life of Energy producing reactor) was proposed, where shapes of neutron flux, nuclide densities and power density distributions remain constant but move upward (or downward) along its core axis. This burnup strategy can derive many merits, especially from safety point of view. The change of excess reactivity along burnup is theoretically zero, and the core characteristics, such as power feedback coefficients and power peaking factor, are not changed along burnup. Application of this burnup strategy to neutron rich fast reactors makes excellent performances. Only natural or depleted uranium is required for the replacing fuels. About 40 % of natural or depleted uranium undergoes fission without the conventional reprocessing and enrichment. If the LWR produced energy of X Joules, the CANDLE reactor can produce about 50 X Joules from the depleted uranium left at the enrichment facility for the LWR fuel. If we can say LWRs have produced energy sufficient for full 20 years, we can produce the energy for 1000 years by using the CANDLE reactors with depleted uranium. We need not mine any uranium ore, and do not need reprocessing facility. The burnup of spent fuel becomes 10 times. Therefore, the spent fuel amount per produced energy is also reduced to one-tenth. The equilibrium core contains a lot of instable materials such as higher actinides and fission products, the enough amounts of which can not be obtained easily. The construction of the initial core is a difficult problem. However, by using enriched uranium substituted for actinides in the equilibrium core, we can construct the initial core whose power profile is similar to the equilibrium one and will reach the equilibrium state without any big change during transient. At present we do not have any material standing for such a high burnup. However, the CANDLE burnup can be realized by employing simple reprocessing, which separates actinides and fission products and replaces the cladding by new one. (author)
Core Calculation of 1 MWatt PUSPATI TRIGA Reactor (RTP) using Monte Carlo MVP Code System
Karim, Julia Abdul
2008-05-20
The Monte Carlo MVP code system was adopted for the Reaktor TRIGA PUSAPTI (RTP) core calculation. The code was developed by a group of researcher of Japan Atomic Energy Agency (JAEA) first in 1994. MVP is a general multi-purpose Monte Carlo code for neutron and photon transport calculation and able to estimate an accurate simulation problems. The code calculation is based on the continuous energy method. This code is capable of adopting an accurate physics model, geometry description and variance reduction technique faster than conventional method as compared to the conventional scalar method. This code could achieve higher computational speed by several factors on the vector super-computer. In this calculation, RTP core was modeled as close as possible to the real core and results of keff flux, fission densities and others were obtained.
Core Calculation of 1 MWatt PUSPATI TRIGA Reactor (RTP) using Monte Carlo MVP Code System
NASA Astrophysics Data System (ADS)
Karim, Julia Abdul
2008-05-01
The Monte Carlo MVP code system was adopted for the Reaktor TRIGA PUSAPTI (RTP) core calculation. The code was developed by a group of researcher of Japan Atomic Energy Agency (JAEA) first in 1994. MVP is a general multi-purpose Monte Carlo code for neutron and photon transport calculation and able to estimate an accurate simulation problems. The code calculation is based on the continuous energy method. This code is capable of adopting an accurate physics model, geometry description and variance reduction technique faster than conventional method as compared to the conventional scalar method. This code could achieve higher computational speed by several factors on the vector super-computer. In this calculation, RTP core was modeled as close as possible to the real core and results of keff flux, fission densities and others were obtained.
Investigation of the Fundamental Constants Stability Based on the Reactor Oklo Burn-Up Analysis
NASA Astrophysics Data System (ADS)
Onegin, M. S.; Yudkevich, M. S.; Gomin, E. A.
2012-12-01
The burn-up of few samples of the natural Oklo reactor zones 3, 5 was calculated using the modern Monte Carlo code. We reconstructed the neutron spectrum in the core by means of the isotope ratios: 147Sm/148Sm and 176Lu/175Lu. These ratios unambiguously determine the water content and core temperature. The isotope ratio of the 149Sm in the sample calculated using this spectrum was compared with experimental one. The disagreement between these two values allows one to limit a possible shift of the low lying resonance of 149Sm. Then, these limits were converted to the limits for the change of the fine structure constant ?. We have found out, that for the rate of ? change, the inequality ?rt? {? }/? ?rt<= 5× 10-18 is fulfilled, which is one order higher than our previous limit.
Designing Critical Experiments in Support of Full Burnup Credit
Mueller, Don; Roberts, Jeremy A
2008-01-01
Burnup credit is the process of accounting for the negative reactivity due to fuel burnup and generation of parasitic absorbers over fuel assembly lifetime. For years, the fresh fuel assumption was used as a simple bound in criticality work for used fuel storage and transportation. More recently, major actinides have been included [1]. However, even this yields a highly conservative estimate in criticality calculations. Because of the numerous economical benefits including all available negative reactivity (i.e., full burnup credit) could provide [2], it is advantageous to work toward full burnup credit. Unfortunately, comparatively little work has been done to include non-major actinides and other fission products (FP) in burnup credit analyses due in part to insufficient experimental data for validation of codes and nuclear data. The Burnup Credit Criticality Experiment (BUCCX) at Sandia National Laboratory was a set of experiments with {sup 103}Rh that have relevance for burnup credit [3]. This work uses TSUNAMI-3D to investigate and adjust a BUCCX model to match isotope-specific, energy-dependent k{sub eff} sensitivity profiles to those of a representative high-capacity cask model (GBC-32) [4] for each FP of interest. The isotopes considered are {sup 149}Sm, {sup 143}Nd, {sup 103}Rh, {sup 133}Cs, {sup 155}Gd, {sup 152}Sm, {sup 99}Tc, {sup 145}Nd, {sup 153}Eu, {sup 147}Sm, {sup 109}Ag, {sup 95}Mo, {sup 150}Sm, {sup 101}Ru, and {sup 151}Eu. The goal is to understand the biases and bias uncertainties inherent in nuclear data, and ultimately, to apply these in support of full burnup credit.
Hartini, Entin, E-mail: entin@batan.go.id; Andiwijayakusuma, Dinan, E-mail: entin@batan.go.id [Center for Development of Nuclear Informatics - National Nuclear Energy Agency, PUSPIPTEK, Serpong, Tangerang, Banten (Indonesia)
2014-09-30
This research was carried out on the development of code for uncertainty analysis is based on a statistical approach for assessing the uncertainty input parameters. In the butn-up calculation of fuel, uncertainty analysis performed for input parameters fuel density, coolant density and fuel temperature. This calculation is performed during irradiation using Monte Carlo N-Particle Transport. The Uncertainty method based on the probabilities density function. Development code is made in python script to do coupling with MCNPX for criticality and burn-up calculations. Simulation is done by modeling the geometry of PWR terrace, with MCNPX on the power 54 MW with fuel type UO2 pellets. The calculation is done by using the data library continuous energy cross-sections ENDF / B-VI. MCNPX requires nuclear data in ACE format. Development of interfaces for obtaining nuclear data in the form of ACE format of ENDF through special process NJOY calculation to temperature changes in a certain range.
NASA Astrophysics Data System (ADS)
Hartini, Entin; Andiwijayakusuma, Dinan
2014-09-01
This research was carried out on the development of code for uncertainty analysis is based on a statistical approach for assessing the uncertainty input parameters. In the butn-up calculation of fuel, uncertainty analysis performed for input parameters fuel density, coolant density and fuel temperature. This calculation is performed during irradiation using Monte Carlo N-Particle Transport. The Uncertainty method based on the probabilities density function. Development code is made in python script to do coupling with MCNPX for criticality and burn-up calculations. Simulation is done by modeling the geometry of PWR terrace, with MCNPX on the power 54 MW with fuel type UO2 pellets. The calculation is done by using the data library continuous energy cross-sections ENDF / B-VI. MCNPX requires nuclear data in ACE format. Development of interfaces for obtaining nuclear data in the form of ACE format of ENDF through special process NJOY calculation to temperature changes in a certain range.
Yamamoto, T.; Suzuki, M.; Ando, Y. [Japan Nuclear Energy Safety Organization, Toranomon Towers Office, 14F, 4-1-28, Toranomon, Minato-ku, Tokyo 105-0001 (Japan)
2012-07-01
After the severe core damage of Fukushima Dai-Ichi Nuclear Power Station, radioactive material leaked from the reactor buildings. As part of monitoring of radioactivity in the site, measurements of radioactivity in soils at three fixed points have been performed for {sup 134}Cs and {sup 137}Cs with gamma-ray spectrometry and for Pu, Pu, and {sup 240}Pu with {alpha}-ray spectrometry. Correlations of radioactivity ratios of {sup 134}Cs to {sup 137}Cs, and {sup 238}Pu to the sum of {sup 239}Pu and {sup 240}Pu with fuel burnup were studied by using theoretical burnup calculations and measurements on isotopic inventories, and compared with the Cs and Pu radioactivity rations in the soils. The comparison indicated that the burnup of the fuel sourcing the radioactivity was from 18 to 38 GWd/t, which corresponded to that of the fuel in the highest power and, therefore, the highest decay heat in operating high-burnup fueled BWR cores. (authors)
NASA Astrophysics Data System (ADS)
Fukushima, M.; Ishikawa, M.; Numata, K.; Jin, T.; Kugo, T.
2014-04-01
Benchmark calculations for reflector effects in fast cores were performed to validate the reliability of scattering data of structural materials in the major evaluated nuclear data libraries, JENDL-4.0, ENDF/B-VII.1 and JEFF-3.1.2. The criticalities of two FCA and two ZPR cores were analyzed by using a continuous energy Monte Carlo calculation code. The ratios of calculation to experimental values were compared between these cores and the sensitivity analyses were performed. From the results, the replacement reactivity from blanket to SS and Na reflector is better evaluated by JENDL-4.0 than by ENDF/B-VII.1 mainly due to the ?bar values of Na and 52Cr.
The burnup dependence of light water reactor spent fuel oxidation
NASA Astrophysics Data System (ADS)
Hanson, Brady Dean
The air oxidation of fragments of Light Water Reactor (LWR) spent fuel with burnup in the range 16-42 MWd/kg M was studied using thermogravimetric analysis. Experiments were conducted in dry air over the temperature range 255-325sp°C. Mass increases were generally followed until the calculated oxygen-to-metal ratio reached 2.7. LWR spent fuel was shown to oxidize via the two step reaction UOsb2-> UOsb {2.4}-> U sb3 Osb8, where the UOsb{2.4} phase is similar to cubic Usb4Osb9. The transition of UOsb2 to UOsb{2.4} was shown to be dependent on the average grain size of the specimen, with smaller-grained fuels oxidizing faster. No correlation with other fuel parameters, such as burnup, was found. The Arrhenius activation energy was calculated as 109 ± 14 kJ molsp{-1}, in agreement with reported values for oxygen diffusion in UOsb{2+x}. The oxidation of UOsb{2.4} to Usb3Osb8 was found to be strongly dependent on both temperature and burnup. At low temperature or high burnup, the fuel fragments displayed a marked resistance to oxidation beyond UOsb{2.4}, in the form of a plateau with nearly constant mass that extended for as long as 3000 hours. Both the duration of the plateau and the time-rate-of-change in the O/M ratio beyond the plateau exhibited identical burnup dependencies within experimental errors. The coefficient for the burnup dependence of the activation energy was determined as 1.2 ± 0.2 kJ molsp{-1} per MWd/kg M. The activation energy extrapolated to zero burnup was shown to agree with the value of 146 ± 10 kJ molsp{-1} reported for the oxidation of unirradiated UOsb2. The correlation of burnup with the kinetics of oxidation, and the fact that burnup is a crude measure of the concentrations of fission products and higher actinides support the conclusion that substitution of fission products and higher actinides into the vacancies in the uranium sublattice of UOsb2 that result from fission stabilizes the fluorite structure with respect to oxidation beyond UOsb{2.4}. Simple estimates show that at least one half of the burnup-dependence of the activation energy may be accounted for by the increase in lattice energy from the contraction produced by substitution of fission products and heavier actinides for uranium vacancies.
Vidal, J. F.; Archier, P.; Calloo, A.; Jacquet, P.; Tommasi, J. [CEA, DEN, DER, Cadarache, F-13108 Saint-Paul-lez-Durance (France); Le Tellier, R. [CEA, DEN, DTN, Cadarache, F-13108 Saint-Paul-lez-Durance (France)
2012-07-01
In the framework of the ASTRID project, sodium cooled fast reactor studies are conducted at CEA in compliance with GEN IV reactors criteria, particularly for safety requirements. An improved safety requires better calculation tools to obtain accurate reactivity effects (especially sodium void effect) and power map distributions. The current calculation route lies on the JEFF3.1.1 library and the classical two-step approach performed with the ECCO module of the ERANOS code system at the assembly level and the Sn SNATCH solver - implemented within the PARIS platform - at the core level. 33-group cross sections used by SNATCH are collapsed from 1968-group self-shielded cross-section with a specific flux-current weighting. Recent studies have shown that this collapsing is non-conservative when dealing with core-reflector interface and can lead to reactivity discrepancies larger than 500 pcm in the case of a steel reflector. Such a discrepancy is due to the flux anisotropy at the interface, which is not taken into account when cross sections are obtained from separate fuel and reflector assembly calculations. A new approach is proposed in this paper. It consists in separating the self-shielding and the flux calculations. The first one is still performed with ECCO on separate patterns. The second one is done with SNATCH on a 1D traverse, representative of the core-reflector interface. An improved collapsing method using angular flux moments is then carried out to collapse the cross sections onto the 33-group structure. In the case of a simplified ZONA2B 2D homogeneous benchmark, results in terms of k{sub eff} and power map are strongly improved for a small increase of the computing time. (authors)
Hay, P.J.; Wadt, W.R.
1985-01-01
Ab initio effective core potentials (ECP's) have been generated to replace the innermost core electron for third-row (K--Au), fourth-row (Rb--Ag), and fifth-row (Cs--Au) atoms. The outermost core orbitals: corresponding to the ns/sup 2/np/sup 6/ configuration for the three rows here: are not replaced by the ECP but are treated on an equal footing with the nd, (n+1)s and (n+1)p valence orbitals. These ECP's have been derived for use in molecular calculations where these outer core orbitals need to be treated explicitly rather than to be replaced by an ECP. The ECP's for the forth and fifth rows also incorporate the mass--velocity and Darwin relativistic effects into the potentials. Analytic fits to the potentials are presented for use in multicenter integral evaluation. Gaussian orbital valence basis sets are developed for the (3s, 3p, 3d, 4s, 4p), (4s, 4p, 4d, 5s, 5p), and (5s, 5p, 5d, 6s, 6p) ortibals of the three respective rows.
Vocadlo, Lidunka
for a partially molten inner core? Lidunka Vocadlo Department of Earth Sciences, University College London, Gower rights reserved. Keywords: iron; iron alloys; Earth's inner core; ab initio calculations; elastic properties There are many unresolved problems concerning our understanding of the Earth's inner core; even
Thermal Analysis for a Heterogeneous VHTR Transmutation Fuel Block as a Function of Burnup
Samuel Bays; Piyush Sabharwall; Stephen Herring; Kevan Weaver
2006-01-01
The VHTR is one of the Generation IV Reactor concepts utilizing TRISO fuel, which enables it to have a higher burnup. Thus, giving higher fuel utilization which is capitalized upon in this work, where one third of the uranium carrying fuel compacts are replaced with transuranic burning fuel for light water reactor waste transmutation. The single fuel block burnup calculation
High Burnup Fuel Behavior Modeling
Jahingir, M.; Rand, R.; Stachowski, R.; Miles, B. [Global Nuclear Fuel-Americas, PO Box 780, Wilmington, NC 28402 (United States); Kusagaya, K. [Global Nuclear Fuel-Japan, 3-1, Uchikawa 2, Yokosuka, Kanagawa, 239-0836 (Japan)
2007-07-01
This paper discusses the development and qualification of the PRIME03 code to address high burnup mechanisms and to improve uranium utilization in current and new reactor designs. Materials properties and behavioral models have been updated from previous thermal-mechanical codes to reflect the effects of burnup on fuel pellet thermal conductivity, Zircaloy creep, fuel pellet relocation, and fission gas release. These new models are based on results of in-pool and post irradiation examination (PIE) of commercial boiling water reactor (BWR) fuel rods at high burnup and results from international experimental programs. The new models incorporated into PRIME03 also address specific high burnup effects associated with formation of pellet rim porosity at high exposure. The PRIME03 code is qualified by comparison of predicted and measured fuel performance parameters for a large number of high, low, and moderate burnup test and commercial reactor rod. The extensive experimental qualification of the PRIME03 prediction capabilities confirms that it is a reliable best-estimate predictor of fuel rod thermal-mechanical performance over a wide range of design and operating conditions. (authors)
Jagannathan, V.; Singh, T.; Pal, U.; Karthikeyan, R.; Sundaram, G.
2006-07-01
India is developing several in-house fuel management codes for the design evaluation of WER-1000 M We reactors, being built at Kudankulam, Tamil Nadu in collaboration with Russian Federation. A lattice burnup code EXCEL provides the few group lattice parameters of various fuel assembly types constituting the core. The core diffusion analyses have been performed by two methods. In the first method the entire fuel assembly is treated as a single homogenized cell. Each fuel assembly cell is divided into 6n{sup 2} triangles, where 'n' is the number of uniform divisions on a side of the hexagon. Regular triangular meshes are used in the active core as well as in surrounding reflector regions. This method is incorporated in the code TRIHEXFA. In the second method a pin by pin description of the core is accomplished by considering the few group lattice parameters generated by EXCEL code for various fuel and non-fuel cells in each fuel assembly. Regular hexagonal cells of one pin pitch are considered in the core and reflector regions. This method is incorporated in HEXPIN code. Both these codes use centre mesh finite difference method (FDM) for regular triangular or hexagonal meshes. It is well known that the large size of the WER fuel assembly, the zigzag structure of the core-baffle zone, the distribution of water tubes of different diameter in this baffle zone and the surrounding steel and water layers of different thickness, all lead to a very complex description of the core-reflector interface. We are analyzing the WER core in fresh state by two other approaches to obtain independent benchmark reference solutions. They are finite element method (FEM) and nodal expansion method (NEM). The few group cross sections of EXCEL are used in the FEM and NEM analyses. The paper would present the comparison of the results of core followup simulations of FD codes with those of FEM and NEM analyses. (authors)
VENUS-2 MOX Core Benchmark: Results of ORNL Calculations Using HELIOS-1.4 - Revised Report
Ellis, RJ
2001-06-01
The Task Force on Reactor-Based Plutonium Disposition (TFRPD) was formed by the Organization for Economic Cooperation and Development/Nuclear Energy Agency (OECD/NEA) to study reactor physics, fuel performance, and fuel cycle issues related to the disposition of weapons-grade (WG) plutonium as mixed-oxide (MOX) reactor fuel. To advance the goals of the TFRPD, 10 countries and 12 institutions participated in a major TFRPD activity: a blind benchmark study to compare code calculations to experimental data for the VENUS-2 MOX core at SCK-CEN in Mol, Belgium. At Oak Ridge National Laboratory, the HELIOS-1.4 code system was used to perform the comprehensive study of pin-cell and MOX core calculations for the VENUS-2 MOX core benchmark study.
Neutronic calculations for the conversion to LEU of a research reactor core
Varvayanni, M.; Catsaros, N.; Stakakis, E.; Grigoriadis, D.
2008-07-15
For a five-year transitional period the Greek Research Reactor (GRR-1) was operating with a mixed core, containing both Low Enrichment (LEU) and High Enrichment (HEU) Uranium MTR- type fuel assemblies. The neutronic study of the GRR-1 conversion to LEU has been performed using a code system comprising the core-analysis code CITATION-LDI2 and the cell-calculation modules XSDRNPM and NITAWL-II of the SCALE code. A conceptual LEU core configuration was defined and analyzed with respect to the three dimensional multi-group neutron fluxes, the power distribution, the control-rod worth and the compliance with pre-defined Operation Limiting Conditions. Perturbation calculations and reactivity feedback computations were also carried out to provide input to a subsequent thermal-hydraulic study. (author)
Electronic Structure Calculations and Adaptation Scheme in Multi-core Computing Environments
Seshagiri, Lakshminarasimhan; Sosonkina, Masha; Zhang, Zhao
2009-05-20
Multi-core processing environments have become the norm in the generic computing environment and are being considered for adding an extra dimension to the execution of any application. The T2 Niagara processor is a very unique environment where it consists of eight cores having a capability of running eight threads simultaneously in each of the cores. Applications like General Atomic and Molecular Electronic Structure (GAMESS), used for ab-initio molecular quantum chemistry calculations, can be good indicators of the performance of such machines and would be a guideline for both hardware designers and application programmers. In this paper we try to benchmark the GAMESS performance on a T2 Niagara processor for a couple of molecules. We also show the suitability of using a middleware based adaptation algorithm on GAMESS on such a multi-core environment.
Jung, Y. S.; Lee, U. C.; Joo, H. G. [Dept. of Nuclear Engineering, Seoul National Univ., 599 Gwanak-ro, Gwanak-gu, Seoul (Korea, Republic of)
2012-07-01
The traditional subgroup method which has been applied for lattice transport calculations has an inherent limitation for non-uniform temperature distributions. As a measure to incorporate temperature dependence into the subgroup formulation, the subgroup level and number density adjustment method have been proposed. In this paper, the temperature dependent subgroup formulations employed for reflecting the non-uniform temperature effects on the resonance spatial self-shielding are examined for the whole core transport calculation with the thermal feedback. For 2D pin-cell problem with non-uniform temperature profiles, the inherent limitation of conventional subgroup method is confirmed. And the improvement in terms of reactivity is observed with the proposed adjustment scheme. For the real PWR core calculation with thermal feedback in the hot-full-power condition, the noticeable correction for the fuel temperature coefficient by about 10% more negative is obtained with the correction schemes. (authors)
Calculation of ex-core physical quantities using the 3D importance functions
NASA Astrophysics Data System (ADS)
Trakas, Christos; De Laubiere, Xavier
2014-06-01
Diverse physical quantities are calculated in engineering studies with penalizing hypotheses to assure the required operation margins for each reactor. Today, these physical quantities are obtained by direct calculations from deterministic or Monte Carlo codes. The related states are critical or sub-critical. The current physical quantities are for example: the SRD counting rates (source range detector) in the sub-critical state, the IRD (intermediary range detector) and PRD (power range detector) counting rates (neutron particles only), the deposited energy in the reflector (neutron + photon particles), the fluence or the DPA (displacement per atom) in the reactor vessel (neutron particles only). The reliability of the proposed methodology is tested in the EPR reactor. The main advantage of the new methodology is the simplicity to obtain the physical quantities by an easy matrix calculation importance linked to nuclear power sources for all the cycles of the reactor. This method also allows to by-pass the direct calculations of the physical quantity of irradiated cores by Monte Carlo Codes, these calculations being impossible today (too many isotopic concentrations / MCNP5 limit). This paper presents the first feasibility study for the physical quantities calculation outside of the core by the importance method instead of the direct calculations used currently by AREVA.
Full Core 3-D Simulation of a Partial MOX LWR Core
S. Bays; W. Skerjanc; M. Pope
2009-05-01
A comparative analysis and comparison of results obtained between 2-D lattice calculations and 3-D full core nodal calculations, in the frame of MOX fuel design, was conducted. This study revealed a set of advantages and disadvantages, with respect to each method, which can be used to guide the level of accuracy desired for future fuel and fuel cycle calculations. For the purpose of isotopic generation for fuel cycle analyses, the approach of using a 2-D lattice code (i.e., fuel assembly in infinite lattice) gave reasonable predictions of uranium and plutonium isotope concentrations at the predicted 3-D core simulation batch average discharge burnup. However, it was found that the 2-D lattice calculation can under-predict the power of pins located along a shared edge between MOX and UO2 by as much as 20%. In this analysis, this error did not occur in the peak pin. However, this was a coincidence and does not rule out the possibility that the peak pin could occur in a lattice position with high calculation uncertainty in future un-optimized studies. Another important consideration in realistic fuel design is the prediction of the peak axial burnup and neutron fluence. The use of 3-D core simulation gave peak burnup conditions, at the pellet level, to be approximately 1.4 times greater than what can be predicted using back-of-the-envelope assumptions of average specific power and irradiation time.
Criticality validation for burnup credit using recycle Pu criticals
Fuentes, E.; Lancaster, D.
1997-04-01
A set of 23 additional critical experiments were analyzed to provide additional input to the criticality validation portion of spent fuel cask analysis. The results of this analyses were combined with the previously analyzed criticals to determine the upper safety limit on k{sub eff}. The combined set of criticals can be used used for criticality validation for burnup credit, and are better suited for the range of isotopics in spent nuclear fuels. A trend observed in the analysis was that the calculated k{sub eff} deviates from the criticals in the positive direction, implying that increased burnup results in increased safety margin. 6 refs., 2 figs., 1 tab.
3D Neutron Transport PWR Full-core Calculation with RMC code
NASA Astrophysics Data System (ADS)
Qiu, Yishu; She, Ding; Fan, Xiao; Wang, Kan; Li, Zeguang; Liang, Jingang; Leroyer, Hadrien
2014-06-01
Nowadays, there are more and more interests in the use of Monte Carlo codes to calculate the detailed power density distributions in full-core reactors. With the Inspur TS1000 HPC Server of Tsinghua University, several calculations have been done based on the EDF 3D Neutron Transport PWR Full-core benchmark through large-scale parallelism. To investigate and compare the results of the deterministic method and Monte Carlo method, EDF R&D and Department of Engineering Physics of Tsinghua University are having a collaboration to make code to code verification. So in this paper, two codes are used. One is the code COCAGNE developed by the EDF R&D, a deterministic core code, and the other is the Monte Carlo code RMC developed by Department of Engineering Physics in Tsinghua University. First, the full-core model is described and a 26-group calculation was performed by these two codes using the same 26-group cross-section library provided by EDF R&D. Then the parallel and tally performance of RMC is discussed. RMC employs a novel algorithm which can cut down most of the communications. It can be seen clearly that the speedup ratio almost linearly increases with the nodes. Furthermore the cell-mapping method applied by RMC consumes little time to tally even millions of cells. The results of the codes COCAGNE and RMC are compared in three ways. The results of these two codes agree well with each other. It can be concluded that both COCAGNE and RMC are able to provide 3D-transport solutions associated with detailed power density distributions calculation in PWR full-core reactors. Finally, to investigate how many histories are needed to obtain a given standard deviation for a full 3D solution, the non-symmetrized condensed 2-group fluxes of RMC are discussed.
Su, Bingjing; Hawari, Ayman, I.
2004-03-30
Modular Pebble Bed Reactor (MPBR) is a high temperature gas-cooled nuclear power reactor currently under study as a next generation reactor system. In addition to its inherently safe design, a unique feature of this reactor is its multi-pass fuel circulation in which the fuel pebbles are randomly loaded and continuously cycled through the core until they reach their prescribed End-of-Life burn-up limit. Unlike the situation with a conventional light water reactor, depending solely on computational methods to perform in-core fuel management for MPBR will be highly inaccurate. An on-line measurement system is needed to accurately assess whether a given pebble has reached its End-of-Life burn-up limit and thereby provide an on-line, automated go/no-go decision on fuel disposition on a pebble-by-pebble basis. This project investigated approaches to analyzing fuel pebbles in real time using gamma spectroscopy and possibly using passive neutron counting of spontaneous fission neutrons to provide the speed, accuracy, and burn-up range required for burnup determination of MPBR. It involved all phases necessary to develop and construct a burn-up monitor, including a review of the design requirements of the system, identification of detection methodologies, modeling and development of potential designs, and finally, the construction and testing of an operational detector system. Based upon the research work performed in this project, the following conclusions are made. In terms of using gamma spectrometry, two possible approaches were identified for burnup assay. The first approach is based on the measurement of the absolute activity of Cs-137. However, due to spectral interference and the need for absolute calibration of the spectrometer, the uncertainty in burnup determination using this approach was found to range from {approx} {+-}40% at beginning of life to {approx} {+-}10% at the discharge burnup. An alternative approach is to use a relative burnup indicator. In this case, a self-calibration method was developed to obtain the spectrometer's relative efficiency curve based upon gamma lines emitted from {sup 140}La. It was found that the ratio of {sup 239}Np/{sup 132}I can be used in burnup measurement with an uncertainty of {approx} {+-}3% throughout the pebble's lifetime. In addition, by doping the fuel with {sup 60}Co, the use of the {sup 60}Co/{sup 134}Cs and {sup 239}Np/{sup 132}I ratios can simultaneously yield the enrichment and burnup of each pebble. A functional gamma-ray spectrometry measurement system was constructed and tested with light water reactor fuels. Experimental results were observed to be consistent with the predictions. On using the passive neutron counting method for the on-line burnup measurement, it was found that neutron emission rate of an irradiated pebble is sensitive to its burnup history and the spectral-averaged cross sections used in the depletion calculations; thus a large uncertainty exists in the correlation between neutron emission and burnup. At low burnup levels, the uncertainty in the neutron emission/burnup correlation is too high and neutron emission rate is too low so that it is impossible to determine a pebble's burnup by on-line neutron counting. At high burnup levels, due to the decreasing of the uncertainty in neutron emission rate and the super-linear feature of the correlation, the uncertainty in burnup determination was found to be {approx}7% at the discharge burnup, which is acceptable for determining whether a pebble should be discharged or not. In terms of neutron detection, because an irradiated pebble is a weak neutron source and a much stronger gamma source, neutron detector system should have high neutron detection efficiency and strong gamma discrimination capability. Of all the commonly used neutron detectors, the He-3 and BF3 detector systems were found to be able to satisfy the requirement on detection efficiency; but their gamma discrimination capability is only marginal for this on-line application. Even with thick gamma shielding, these two types of detectors sha
Lee, C. [Argonne National Laboratory, 9700 S. Cass Avenue, Argonne, IL 60439-4842 (United States); Yang, W. S. [Purdue University, School of Nuclear Engineering, West Lafayette, IN 47907-2017 (United States)
2013-07-01
An improved resonance self-shielding method has been developed to accurately estimate the effective multigroup cross sections for heterogeneous fast reactor assembly and core calculations. In the method, the heterogeneity effect is considered by the use of isotopic escape cross sections while the resonance interference effect is accounted for through the narrow resonance approximation or slowing-down calculations for specific compositions. The isotopic escape cross sections are calculated by solving fixed-source transport equations with the method of characteristics for the whole problem domain. This method requires no pre-calculated resonance integral tables or parameters that are typically necessary in the subgroup method. Preliminary results for multi pin-cell fast reactor problems show that the escape cross sections estimated from the explicit-geometry fixed source calculations produce more accurate eigenvalue and self-shielded effective cross sections than those from conventional one-dimensional geometry models. (authors)
Developing integration and extrapolation methods for no-core configuration interaction calculations
NASA Astrophysics Data System (ADS)
Rakoski, Alexa I.; Caprio, Mark A.
2014-09-01
A study of the states of light nuclei in the no-core shell model is limited by the size of basis for which calculations are possible, but the results for small, calculable bases can be extrapolated to larger basis size. To understand the properties of this extrapolation, one-dimensional models in the harmonic oscillator and Coulomb-Sturmian bases are explored because of the relative ease of calculations. Using a linear algebra approach to solving Schrodinger's equation, the wave function can be determined in these models and observables such as energy and radius can be calculated. However, the integrals required for this process become cumbersome to evaluate using standard numerical methods for large basis size even in the one-dimensional model. Alternate numerical techniques are tested to determine the most effective in extending the calculations to larger basis size, and extrapolation methods within the model are explored. A study of the states of light nuclei in the no-core shell model is limited by the size of basis for which calculations are possible, but the results for small, calculable bases can be extrapolated to larger basis size. To understand the properties of this extrapolation, one-dimensional models in the harmonic oscillator and Coulomb-Sturmian bases are explored because of the relative ease of calculations. Using a linear algebra approach to solving Schrodinger's equation, the wave function can be determined in these models and observables such as energy and radius can be calculated. However, the integrals required for this process become cumbersome to evaluate using standard numerical methods for large basis size even in the one-dimensional model. Alternate numerical techniques are tested to determine the most effective in extending the calculations to larger basis size, and extrapolation methods within the model are explored. Supported by the US DOE under Grant DE-FG02-95ER-40934 and the Research Corporation for Science Advancement under a Cottrell Scholar Award.
NASA Astrophysics Data System (ADS)
Antsiferova, E. V.; Bogdanov, V. V.; Derebenko, E. V.; Lagutina, A. V.; Khmelnikov, E. A.
2006-08-01
The up-to-date development of the armored vehicles conditions complication of armor constructions and increased slope of shell armored plates. Combined strikers (C/S) can be used to destroy armored vehicles. We can increase total weight of the core part to increase the striker's power. However, the increase of core part diameter is limited by body dimensions. Thus, we can increase core part weight by increasing its length. Because of C/S interaction with the barriers at large deviation angles, C/S's mechanical trajectory sparks in the barrier. This results in bending stress which occurs in the core part. Because of large deviation angles, the impact of the side surface of oblong core part against the cavity edge occurs. This increases the probability of core part destruction. The calculation technique for oblong core part penetration into different types of barriers is presented. The large number of factors can be calculated using this technique. It is assumed that the core part is destroyed when the tail part impacts against the cavity in the section where specific impact energy exceeds the critical value. Impact elasticity and destruction at bending stress were selected to be destruction criteria. The following core part destruction scenarios were investigated and calculated: (i) core head part is slightly destroyed but tail part of cylindrical shape penetrates deeper; (ii) core tail part is slightly destroyed but head part penetrates deeper, mass loss is taken into account; and (iii) after the impact, the core part is splitted up into two parts, then both of them penetrate into the barrier, one part is of ogival shape, the other is of cylindrical one. This calculation technique was applied to computational program, then critical angles at which core part side surface is still in contact with cavity surface, and the angles at which core part destruction occurs were calculated. Depths of core part penetration for different destruction scenarios were calculated.
VENUS-2 MOX Core Benchmark: Results of ORNL Calculations Using HELIOS-1.4
Ellis, RJ
2001-02-02
The Task Force on Reactor-Based Plutonium Disposition, now an Expert Group, was set up through the Organization for Economic Cooperation and Development/Nuclear Energy Agency to facilitate technical assessments of burning weapons-grade plutonium mixed-oxide (MOX) fuel in U.S. pressurized-water reactors and Russian VVER nuclear reactors. More than ten countries participated to advance the work of the Task Force in a major initiative, which was a blind benchmark study to compare code benchmark calculations against experimental data for the VENUS-2 MOX core at SCK-CEN in Mol, Belgium. At the Oak Ridge National Laboratory, the HELIOS-1.4 code was used to perform a comprehensive study of pin-cell and core calculations for the VENUS-2 benchmark.
Liquid iron-sulfur alloys at outer core conditions by first-principles calculations
NASA Astrophysics Data System (ADS)
Umemoto, Koichiro; Hirose, Kei; Imada, Saori; Nakajima, Yoichi; Komabayashi, Tetsuya; Tsutsui, Satoshi; Baron, Alfred Q. R.
2014-10-01
We perform first-principles calculations to investigate liquid iron-sulfur alloys (Fe, Fe56S8, Fe52S12, and Fe48S16) under high-pressure and high-temperature (150-300 GPa and 4000-6000 K) conditions corresponding to the Earth's outer core. Considering only the density profile, the best match with the preliminary reference Earth model is by liquid Fe-14 wt % S (Fe50S14), assuming sulfur is the only light element. However, its bulk sound velocity is too high, in particular in the deep outer core, suggesting that another light component such as oxygen is required. An experimental check using inelastic X-ray scattering shows good agreement with the calculations. In addition, a present study demonstrates that the Birch's law does not hold for liquid iron-sulfur alloy, consistent with a previous report on pure liquid iron.
Analysis on burnup step effect for evaluating reactor criticality and fuel breeding ratio
Saputra, Geby; Purnama, Aditya Rizki; Permana, Sidik; Suzuki, Mitsutoshi
2014-09-30
Criticality condition of the reactors is one of the important factors for evaluating reactor operation and nuclear fuel breeding ratio is another factor to show nuclear fuel sustainability. This study analyzes the effect of burnup steps and cycle operation step for evaluating the criticality condition of the reactor as well as the performance of nuclear fuel breeding or breeding ratio (BR). Burnup step is performed based on a day step analysis which is varied from 10 days up to 800 days and for cycle operation from 1 cycle up to 8 cycles reactor operations. In addition, calculation efficiency based on the variation of computer processors to run the analysis in term of time (time efficiency in the calculation) have been also investigated. Optimization method for reactor design analysis which is used a large fast breeder reactor type as a reference case was performed by adopting an established reactor design code of JOINT-FR. The results show a criticality condition becomes higher for smaller burnup step (day) and for breeding ratio becomes less for smaller burnup step (day). Some nuclides contribute to make better criticality when smaller burnup step due to individul nuclide half-live. Calculation time for different burnup step shows a correlation with the time consuming requirement for more details step calculation, although the consuming time is not directly equivalent with the how many time the burnup time step is divided.
NASA Astrophysics Data System (ADS)
Sboev, A. G.; Ilyashenko, A. S.; Vetrova, O. A.
1997-02-01
The method of bucking evaluation, realized in the MOnte Carlo code MCS, is described. This method was applied for calculational analysis of well known light water experiments TRX-1 and TRX-2. The analysis of this comparison shows, that there is no coincidence between Monte Carlo calculations, obtained by different ways: the MCS calculations with given experimental bucklings; the MCS calculations with given bucklings evaluated on base of full core MCS direct simulations; the full core MCNP and MCS direct simulations; the MCNP and MCS calculations, where the results of cell calculations are corrected by the coefficients taking into the account the leakage from the core. Also the buckling values evaluated by full core MCS calculations have differed from experimental ones, especially in the case of TRX-1, when this difference has corresponded to 0.5 percent increase of Keff value.
Non-Invasive Characterization of Burnup for PWR Spent Fuel Rods with Burnups > 80 GWd/t
Caruso, S.; Murphy, M.; Jatuff, F.; Chawla, R.
2006-07-01
High-resolution gamma spectroscopy has been employed for the measurement of {sup 134}Cs/{sup 137}Cs, {sup 154}Eu/{sup 137}Cs and {sup 134}Cs/{sup 154}Eu gamma intensity ratios from spent fuel with the purpose of deriving pin-averaged single-ratio burnup indicators for high and ultra-high burnups. Two UO{sub 2} pressurised water reactor (PWR) fuel rod segments with record burnup levels >80 GWd/t have been experimentally characterised. Additionally, pin cell depletion calculations have been performed for each sample with the deterministic code CASMO-4, using both its JEF2.2- and its ENDF/B-IV-based libraries, for three different descriptions of the fuel rod irradiation histories, in order to test the sensitivity of the results to neutron cross sections and to the depletion model employed. Measured and calculated ratios have then been compared. It is shown that the {sup 134}Cs/{sup 137}Cs ratio, frequently used as burnup monitor, is considerably less accurate for values exceeding 50 GWd/t; discrepancies of up to {approx}25% are found between measured and calculated values. The ratios built with the {sup 154}Eu concentration show much larger discrepancies, essentially because this isotope is rather poorly predicted as revealed by the use of different basic cross section data. (authors)
Numerical Toy-Model Calculation of the Nucleon Spin Autocorrelation Function in a Supernova Core
Georg Raffelt; Guenter Sigl
1998-08-30
We develop a simple model for the evolution of a nucleon spin in a hot and dense nuclear medium. A given nucleon is limited to one-dimensional motion in a distribution of external, spin-dependent scattering potentials. We calculate the nucleon spin autocorrelation function numerically for a variety of potential densities and distributions which are meant to bracket realistic conditions in a supernova core. For all plausible configurations the width of the spin-density structure function is found to be less than the temperature. This is in contrast with a naive perturbative calculation based on the one-pion exchange potential which overestimates the width and thus suggests a large suppression of the neutrino opacities by nucleon spin fluctuations. Our results suggest that it may be justified to neglect the collisional broadening of the spin-density structure function for the purpose of estimating the neutrino opacities in the deep inner core of a supernova. On the other hand, we find no indication that processes such as axion or neutrino pair emission, which depend on nucleon spin fluctuations, are substantially suppressed beyond the multiple-scattering effect already discussed in the literature. Aside from these practical conclusions, our model reveals a number of interesting and unexpected insights. For example, the spin-relaxation rate saturates with increasing potential strength only if bound states are not allowed to form by including a repulsive core. There is no saturation with increasing density of scattering potentials until localized eigenstates of energy begin to form.
Calculation of scattering characteristic of complex target on multi-core platform
NASA Astrophysics Data System (ADS)
Guo, Xing; Wu, Zhensen; Linghu, Longxiang
2013-09-01
The scattering characteristic of complex target from terrestrial and celestial background radiation has been widely used in such engineering fields as remote sensing, feature extraction, tracking and recognition of target thus having been an attractive field for many scientists for decades. In our method, the model of target is constructed using 3DMAX and the surface is divided into triangle facets firstly. Bidirectional Reflectance Distribution Function (BRDF) is introduced and MODTRAN is applied to calculate background radiation for a given time at a given place. Finally the scattering of each facet is added up to get the scattering of the target. As the background radiance comes in all directions and in a wide spectrum and the complex target always consists of thousands of facets, in general it takes hours to complete the calculation. Consequently this limits its use in the real time applications. Recent years have seen the continual development of multi-core CPU. As a result parallel programming on multi-cores has been more and more popular. In this paper, the openMP, Intel CILK ++, Intel Threading Building Blocks (TBB) are used separately to leverage the processing power of multi-cores processors. Our experiments are conducted on a DELL desktop based on an Intel I7- 2600K CPU running at 3.40 GHz with 8 cores and 16.0 GB RAM. The Intel Composer 2013 is employed to build the program. Also in OpenMP implementation, gcc is used. The results demonstrate that highest speedups for three parallel models are 5.06X, 5.02X, 5.15X respectively.
Review of Technical Studies in the United States in Support of Burnup Credit Regulatory Guidance
Wagner, John C [ORNL] [ORNL; Parks, Cecil V [ORNL] [ORNL; Mueller, Don [ORNL] [ORNL; Gauld, Ian C [ORNL] [ORNL
2010-01-01
Taking credit for the reduction in reactivity associated with fuel depletion can enable more cost-effective, higher-density storage, transport, disposal, and reprocessing of spent nuclear fuel (SNF) while maintaining sufficient subcritical margin to establish an adequate safety basis. Consequently, there continues to be considerable interest in the United States (U.S.), as well as internationally, in the increased use of burnup credit in SNF operations, particularly related to storage, transport, and disposal of commercial SNF. This interest has motivated numerous technical studies related to the application of burnup credit, both domestically and internationally, as well as the design of SNF storage, transport and disposal systems that rely on burnup credit for maintaining subcriticality. Responding to industry requests and needs, the U.S. Nuclear Regulatory Commission (NRC) initiated a burnup credit research program in 1999, with support from the Oak Ridge National Laboratory (ORNL), to develop regulatory guidance and the supporting technical bases for allowing and expanding the use of burnup credit in pressurized-water reactor SNF storage and transport applications. Although this NRC research program has not been continuous since its inception, considerable progress has been achieved in many key areas in terms of increased understanding of relevant phenomena and issues, availability of relevant information and data, and subsequently updated regulatory guidance for expanded use of burnup credit. This paper reviews technical studies performed by ORNL for the U.S. NRC burnup credit research program. Examples of topics include reactivity effects associated with reactor operating characteristics, fuel assembly characteristics, burnable absorbers, control rods, spatial burnup distributions, cooling time, and assembly misloading; methods and data for validation of isotopic composition predictions; methods and data for validation of criticality calculations; and operational issues and data related to assembly burnup confirmation. The objective of this paper is to summarize the work and significant accomplishments, with references to the technical reports and publications for complete details, and provide a useful resource to others in the burnup credit community.
NASA Astrophysics Data System (ADS)
Espel, Federico Puente
The main objective of this PhD research is to develop a high accuracy modeling tool using a Monte Carlo based coupled system. The presented research comprises the development of models to include the thermal-hydraulic feedback to the Monte Carlo method and speed-up mechanisms to accelerate the Monte Carlo criticality calculation. Presently, deterministic codes based on the diffusion approximation of the Boltzmann transport equation, coupled with channel-based (or sub-channel based) thermal-hydraulic codes, carry out the three-dimensional (3-D) reactor core calculations of the Light Water Reactors (LWRs). These deterministic codes utilize nuclear homogenized data (normally over large spatial zones, consisting of fuel assembly or parts of fuel assembly, and in the best case, over small spatial zones, consisting of pin cell), which is functionalized in terms of thermal-hydraulic feedback parameters (in the form of off-line pre-generated cross-section libraries). High accuracy modeling is required for advanced nuclear reactor core designs that present increased geometry complexity and material heterogeneity. Such high-fidelity methods take advantage of the recent progress in computation technology and coupled neutron transport solutions with thermal-hydraulic feedback models on pin or even on sub-pin level (in terms of spatial scale). The continuous energy Monte Carlo method is well suited for solving such core environments with the detailed representation of the complicated 3-D problem. The major advantages of the Monte Carlo method over the deterministic methods are the continuous energy treatment and the exact 3-D geometry modeling. However, the Monte Carlo method involves vast computational time. The interest in Monte Carlo methods has increased thanks to the improvements of the capabilities of high performance computers. Coupled Monte-Carlo calculations can serve as reference solutions for verifying high-fidelity coupled deterministic neutron transport methods with detailed and accurate thermal-hydraulic models. The development of such reference high-fidelity coupled multi-physics scheme is described in this dissertation on the basis of MCNP5, NEM, NJOY and COBRA-TF (CTF) computer codes. This work presents results from studies performed and implemented at the Pennsylvania State University (PSU) on both accelerating Monte Carlo criticality calculations by using hybrid nodal diffusion Monte Carlo schemes and thermal-hydraulic feedback modeling in Monte Carlo core calculations. The hybrid MCNP5/CTF/NEM/NJOY coupled code system is proposed and developed in this dissertation work. The hybrid coupled code system contains a special interface developed to update the required MCNP5 input changes to account for dimension and density changes provided by the thermal-hydraulics feedback module. The interface has also been developed to extract the flux and reaction rates calculated by MCNP5 to later transform the data into the power feedback needed by CTF (axial and radial peaking factors). The interface is contained in a master program that controls the flow of the calculations. Both feedback modules (thermal-hydraulic and power subroutines) use a common internal interface to further accelerate the data exchange. One of the most important steps to correctly include the thermal hydraulic feedback into MCNP5 calculations begins with temperature dependent cross section libraries. If the cross sections used for the calculations are not at the correct temperature, the temperature feedback cannot be included into MCNP5 (referred to the effect of temperature on cross sections: Doppler boarding of resolve and unresolved resonances, thermal scattering and elastic scattering). The only method of considering the temperature effects on cross sections is through the generation (or as introduced in this dissertation through a novel interpolation mechanism) of continuous energy temperature-dependent cross section libraries. An automated methodology for generation of continuous energy temperature-dependent cross section libraries has been developed
Investigation of Burnup Credit Issues in BWR Fuel
Broadhead, B.L.; DeHart, M.D.
1999-09-20
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.
Wadt, W.R.; Hay, P.J.
1985-01-01
A consistent set of ab initio effective core potentials (ECP) has been generated for the main group elements from Na to Bi using the procedure originally developed by Kahn. The ECP's are derived from all-electron numerical Hartree--Fock atomic wave functions and fit to analytical representations for use in molecular calculations. For Rb to Bi the ECP's are generated from the relativistic Hartree--Fock atomic wave functions of Cowan which incorporate the Darwin and mass--velocity terms. Energy-optimized valence basis sets of (3s3p) primitive Gaussians are presented for use with the ECP's. Comparisons between all-electron and valence-electron ECP calculations are presented for NaF, NaCl, Cl/sub 2/, Cl/sub 2//sup -/, Br/sub 2/, Br/sub 2//sup -/, and Xe/sub 2//sup +/. The results show that the average errors introduced by the ECP's are generally only a few percent.
Automatic whole core depletion and criticality calculations by MCNPX 2.7.0
Kalcheva, S.; Koonen, E. [SCKCEN, BR2 Reactor Dept., Boeretang 200, 2400 Mol (Belgium)
2012-07-01
Different approaches to perform automatic whole core criticality and depletion calculations in a research reactor using MCNPX 2.7.0 are presented. An approximate method is to use the existing symmetries of the burned fuel material distribution in the core, i.e., the axial, radial and azimuth symmetries around the core center, in order to significantly reduce the computation time. In this case it is not necessary to give a unique material number to each burn up cell. Cells having similar burn up and power, achieved during similar irradiation history at same initial fuel composition, will experience similar composition evolution and can therefore be given the same material number. To study the impact of the number of unique burn up materials on the computation time and utilized RAM memory, several MCNPX models have been developed. The paper discusses the accuracy of the model on comparison with measurements of BR2 operation cycles in function of the number of unique burn up materials and the impact of the used Q-value (MeV/fission) of the recoverable fission energy. (authors)
SENSITIVITY AND UNCERTAINTY ANALYSIS OF COMMERCIAL REACTOR CRITICALS FOR BURNUP CREDIT
Radulescu, Georgeta; Mueller, Don; Wagner, John C
2009-01-01
The purpose of this study is to provide insights into the neutronic similarities that may exist between a generic cask containing typical spent nuclear fuel assemblies and commercial reactor critical (CRC) state-points. Forty CRC state-points from five pressurized-water reactors were selected for the study and the type of CRC state-points that may be applicable for validation of burnup credit criticality safety calculations for spent fuel transport/storage/disposal systems are identified. The study employed cross-section sensitivity and uncertainty analysis methods developed at Oak Ridge National Laboratory and the TSUNAMI set of tools in the SCALE code system as a means to investigate system similarity on an integral and nuclide-reaction specific level. The results indicate that, except for the fresh fuel core configuration, all analyzed CRC state-points are either highly similar, similar, or marginally similar to a generic cask containing spent nuclear fuel assemblies with burnups ranging from 10 to 60 GWd/MTU. Based on the integral system parameter, C{sub k}, approximately 30 of the 40 CRC state-points are applicable to validation of burnup credit in the generic cask containing typical spent fuel assemblies with burnups ranging from 10 to 60 GWd/MTU. The state-points providing the highest similarity (C{sub k} > 0.95) were attained at or near the end of a reactor cycle. The C{sub k} values are dominated by neutron reactions with major actinides and hydrogen, as the sensitivities of these reactions are much higher than those of the minor actinides and fission products. On a nuclide-reaction specific level, the CRC state-points provide significant similarity for most of the actinides and fission products relevant to burnup credit. A comparison of energy-dependent sensitivity profiles shows a slight shift of the CRC K{sub eff} sensitivity profiles toward higher energies in the thermal region as compared to the K{sub eff} sensitivity profile of the generic cask. Parameters representing coverage of the application by the CRCs on an energy-dependent, nuclide-reaction specific level (i.e., effectiveness of the CRCs for validating the cross sections as used in the application) were also examined. Based on the CRCs with C{sub k} > 0.8 and an assumed relative standard deviation for uncovered covariance data of 25%, the relative standard deviation of K{sub eff} due to uncovered sensitivity data varies from 0.79% to 0.95% for cask burnups ranging from 10 to 60 GWd/MTU. As expected, this uncertainty in K{sub eff} is largely dominated by noncoverage of sensitivities from major actinides and hydrogen. The contributions from fission products and minor actinides are very small and comparable to statistical uncertainties in K{sub eff} results. These results (again, assuming a 25% uncertainty for uncovered covariance data) indicate that there could be approximately 1% uncertainty in the calculated application K{sub eff} due to incomplete neutronic testing (validation) of the software by the CRCs. However, this conclusion also assumes all other uncertainties in the complex CRC configurations (e.g., isotopic compositions of burned fuel, operation history, data) are well known. Thus, an evaluation of the uncertainties in the CRC configurations is needed prior to the use of CRCs for code validation (i.e., quantifying code bias and bias uncertainty).
NASA Technical Reports Server (NTRS)
Tuma, Margaret L.; Weisshaar, Andreas; Li, Jian; Beheim, Glenn
1995-01-01
To determine the feasibility of coupling the output of a single-mode optical fiber into a single-mode rib waveguide in a temperature varying environment, a theoretical calculation of the coupling efficiency between the two was investigated. Due to the complex geometry of the rib guide, there is no analytical solution to the wave equation for the guided modes, thus, approximation and/or numerical techniques must be utilized to determine the field patterns of the guide. In this study, three solution methods were used for both the fiber and guide fields; the effective-index method (EIM), Marcatili's approximation, and a Fourier method. These methods were utilized independently to calculate the electric field profile of each component at two temperatures, 20 C and 300 C, representing a nominal and high temperature. Using the electric field profile calculated from each method, the theoretical coupling efficiency between an elliptical-core optical fiber and a rib waveguide was calculated using the overlap integral and the results were compared. It was determined that a high coupling efficiency can be achieved when the two components are aligned. The coupling efficiency was more sensitive to alignment offsets in the y direction than the x, due to the elliptical modal field profile of both components. Changes in the coupling efficiency over temperature were found to be minimal.
Calculation of core-level excitation in some MAX-phase compounds
NASA Astrophysics Data System (ADS)
Wang, Liaoyuan; Rulis, Paul; Ching, W. Y.
2013-07-01
We report first-principles spectroscopic calculation of core level excitations in five MAX-phase compounds. The spectra of Ti-K edges in Ti2AlC and Ti2AlN, C-K edge in Ti2AlC, N-K edge in Ti2AlN, and Nb-K edge in Nb2AlC are calculated and found to be in good agreement with reported experimental measurements. Based on this agreement, the Al-K and Al-L3 edges in the same five phases plus the Cr-K and C-K edges in Cr2AlC and the C-K edge in Nb2AlC are calculated as theoretical predictions. We further analyze the anisotropy in the calculated spectra to gain additional insights on the structure-properties relationships in these MAX-phase compounds. These results are further discussed in the context of the local atomic environments of the M, A, and X elements in MAX-phase compounds and in relation to their fundamental electronic structures.
NASA Astrophysics Data System (ADS)
Serrano-Purroy, D.; Clarens, F.; González-Robles, E.; Glatz, J. P.; Wegen, D. H.; de Pablo, J.; Casas, I.; Giménez, J.; Martínez-Esparza, A.
2012-08-01
Two weak points in Performance Assessment (PA) exercises regarding the alteration of Spent Nuclear Fuel (SNF) are the contribution of the so-called Instant Release Fraction (IRF) and the effect of High Burn-Up Structure (HBS). This manuscript focuses on the effect of HBS in matrix (long term) and instant release of a Pressurised Water Reactor (PWR) SNF irradiated in a commercial reactor with a mean Burn-Up (BU) of 60 GWd/tU. In order to study the HBS contribution, two samples from different radial positions have been prepared. One from the centre of the SNF, labelled CORE, and one from the periphery, enriched with HBS and labelled OUT. Static leaching experiments have been carried out with two synthetic leaching solutions: bicarbonate (BIC) and Bentonitic Granitic Groundwater (BGW), and in all cases under oxidising conditions. IRF values have been calculated from the determined Fraction of Inventory in Aqueous Phase (FIAP). In all studied cases, some radionuclides (RN): Rb, Sr and Cs, have shown higher release rates than uranium, especially at the beginning of the experiment, and have been considered as IRF. Redox sensitive RN like Mo and Tc have been found to dissolve slightly faster than uranium and further studies might be needed to confirm if they can also be considered part of the IRF. Most of the remaining studied RN, mainly actinides and lanthanides, have been found to dissolve congruently with the uranium matrix. Finally, Zr, Ru and Rh presented lower release rates than the matrix. Higher matrix release has been determined for CORE than for OUT samples showing that the formation of HBS might have a protective effect against the oxidative corrosion of the SNF. On the contrary, no significant differences have been observed between the two studied leaching solutions (BIC and BGW). Two different IRF contributions have been determined. One corresponding to the fraction of inventory segregated in the external open grain boundaries, directly available to water and very significant during the first days of the experiment; and a second one corresponding to a less accessible, most probably located at the internal grain boundaries, one order of magnitude lower than the first one at equal given dissolution times but of much longer period of incidence. Unlike matrix release results, higher Cs IRF release was found for OUT than for CORE sample. This effect can be attributed to thermal migration of Cs to the periphery of the fuel during irradiation. In the case of Rb no clear differences were observed between CORE and OUT showing equilibrium between the opposing thermal migration and matrix effects. Finally, Sr CORE/OUT release ratio showed similar behaviour to matrix release, thus proving no significant thermal migration during irradiation.
The REBUS Experimental Programme for Burn-Up Credit
D'hondt, Pierre; Van der Meer, Klaas; Baeten, Peter [SCK.CEN, Boeretang 200, 2400 Mol (Belgium); Marloye, Daniel; Lance, Benoit; Basselier, Jacques [Belgonucleaire (Belgium)
2002-07-01
An international programme called REBUS (Reactivity tests for a direct evaluation of the Burn-Up credit on Selected irradiated LWR fuel bundles) for the investigation of the burn-up credit has been initiated by the Belgian Nuclear Research Center SCK.CEN and Belgonucleaire with the support of USNRC, EdF from France, VGB, representing German nuclear utilities and NUPEC, representing the Japanese industry. The programme aims to establish a neutronic benchmark for reactor physics codes. This benchmark will qualify the codes to perform calculations of the burn-up credit. The benchmark exercise investigates the following fuel types with associated burn-up: - Reference 3.3% enriched UO{sub 2} fuel; - Fresh commercial PWR UO{sub 2} fuel; - Irradiated commercial PWR UO{sub 2} fuel (51 GWd/tM); - Fresh PWR MOX fuel; - Irradiated PWR MOX fuel (20 GWd/tM). Reactivity effects are measured in the critical facility VENUS. Fission rate and flux distributions in the experimental bundles will be determined. The accumulated burn-up of all rods is measured non-destructively in a relative way by gross gamma-scanning, while some rods are examined by gamma-spectrometry for an absolute determination of the burn-up. Some rods will be analyzed destructively with respect to accumulated burn-up, actinides content and TOP-19 fission products (i.e. those non-gaseous fission products that have most implications on the reactivity). Additionally some irradiated rods have undergone a profilometry and length determination. The experimental implementation of the programme has started in 2000 with major changes in the VENUS critical facility. Gamma scans, profilometry, length determination and gamma-spectrometry measurements on the MOX fuel have been performed. In the course of October 2001 the first fresh fuel configuration will be investigated. In the same period the commercial irradiated fuel will arrive at the SCK.CEN hot cells and will be re-fabricated into fuel rodlets of 1 meter length. (authors)
Comparative sodium void effects for different advanced liquid metal reactor fuel and core designs
Dobbin, K.D.; Kessler, S.F.; Nelson, J.V.; Gedeon, S.R.; Omberg, R.P.
1991-07-01
An analysis of metal-, oxide, and nitride-fueled advanced liquid metal reactor cores was performed to investigate the calculated differences in sodium void reactivity, and to determine the relationship between sodium void reactivity and burnup reactivity swing using the three fuel types. The results of this analysis indicate that nitride fuel has the least positive sodium void reactivity for any given burnup reactivity swing. Thus, it appears that a good design compromise between transient overpower and loss of flow response is obtained using nitride fuel. Additional studies were made to understand these and other nitride advantages. 8 refs., 5 figs., 3 tabs.
The burnup dependence of light water reactor spent fuel oxidation
Brady Dean Hanson
1998-01-01
The air oxidation of fragments of Light Water Reactor (LWR) spent fuel with burnup in the range 16-42 MWd\\/kg M was studied using thermogravimetric analysis. Experiments were conducted in dry air over the temperature range 255-325sp°C. Mass increases were generally followed until the calculated oxygen-to-metal ratio reached 2.7. LWR spent fuel was shown to oxidize via the two step reaction
Thermal-hydraulic calculations for the conversion to LEU of a research reactor core
Grigoriadis, D.; Varvayanni, M.; Catsaros, N.; Stakakis, E.
2008-07-15
The thermal-hydraulic analysis performed for the needs of the conversion of the open pool 5MW Greek Research Reactor (GRR-1) to a pure Low Enrichment (LEU) configuration is presented. The methodology was based on a complete set of neutronic calculations performed for the new core configuration, in compliance with pre-defined Operation Limiting Conditions. The hottest channel analysis approach was adopted, and peaking factors were used to account for fabrication or measuring uncertainties. Calculations were carried out using the numerical codes NATCON, PLTEMP and PARET provided by Argonne National Laboratory (ANL). Two main different classes of conditions were considered, namely i) steady state normal operating conditions and ii) transient cases related to accidental events including reactivity feedback effects. For steady state operating conditions the behaviour of the new configuration was examined both for forced and natural convection cooling modes. Transient calculations considered several initiating events including reactivity insertion accidents (slow or fast reactivity insertion) and total or partial loss-of-flow accidents, i.e. in accordance to guidelines provided by the IAEA for research Reactors. (author)
Burnup analysis of rock-like oxide fuel disks irradiated in the Japan Research Reactor No. 3
NASA Astrophysics Data System (ADS)
Nakano, Y.; Akie, H.; Magara, M.; Takano, H.
1999-08-01
Burnup analysis of rock-like oxide (ROX) fuel disks has been carried out and the results have been compared with measured values. Two kinds of ROX disks: zirconia and thoria, were fabricated and irradiated in an irradiation hole of the Japan Research Reactor No. 3 (JRR-3). After irradiation, several post-irradiation examinations (PIE) were performed. Computer codes used for the calculations were the SRAC and the MVP-BURN codes. Firstly, the neutron spectrum in the irradiation hole was calculated using the SRAC code system. Fixed source problems were solved to obtain the neutron spectra and effective cross-sections of the disks and burnup calculations were performed. The calculated results of burnup, isotopic abundance of plutonium and production of americium and curium were compared with measurement values. Calculations overestimate the measured burnup by 7 ˜ 15% and both codes largely underestimate the measured production of americium and curium isotopes. The calculated plutonium abundance agrees moderately well with the measured values.
Accident source terms for light-water nuclear power plants using high-burnup or MOX fuel.
Salay, Michael; Gauntt, Randall O.; Lee, Richard Y.; Powers, Dana Auburn; Leonard, Mark Thomas
2011-01-01
Representative accident source terms patterned after the NUREG-1465 Source Term have been developed for high burnup fuel in BWRs and PWRs and for MOX fuel in a PWR with an ice-condenser containment. These source terms have been derived using nonparametric order statistics to develop distributions for the timing of radionuclide release during four accident phases and for release fractions of nine chemical classes of radionuclides as calculated with the MELCOR 1.8.5 accident analysis computer code. The accident phases are those defined in the NUREG-1465 Source Term - gap release, in-vessel release, ex-vessel release, and late in-vessel release. Important differences among the accident source terms derived here and the NUREG-1465 Source Term are not attributable to either fuel burnup or use of MOX fuel. Rather, differences among the source terms are due predominantly to improved understanding of the physics of core meltdown accidents. Heat losses from the degrading reactor core prolong the process of in-vessel release of radionuclides. Improved understanding of the chemistries of tellurium and cesium under reactor accidents changes the predicted behavior characteristics of these radioactive elements relative to what was assumed in the derivation of the NUREG-1465 Source Term. An additional radionuclide chemical class has been defined to account for release of cesium as cesium molybdate which enhances molybdenum release relative to other metallic fission products.
Varivtsev, A. V. Zhemkov, I. Yu.
2014-12-15
The application of the improved method for calculating the radiation heat generation in the elements of an experimental device located at the periphery of the BOR-60 reactor core results in a significant reduction in the discrepancies between the calculated and the experimental data. This allows us to conclude that the improved method has an advantage over the one used earlier.
Ab Initio No-Core Shell Model Calculations Using Realistic Two- and Three-Body Interactions
Navratil, P; Ormand, W E; Forssen, C; Caurier, E
2004-11-30
There has been significant progress in the ab initio approaches to the structure of light nuclei. One such method is the ab initio no-core shell model (NCSM). Starting from realistic two- and three-nucleon interactions this method can predict low-lying levels in p-shell nuclei. In this contribution, we present a brief overview of the NCSM with examples of recent applications. We highlight our study of the parity inversion in {sup 11}Be, for which calculations were performed in basis spaces up to 9{Dirac_h}{Omega} (dimensions reaching 7 x 10{sup 8}). We also present our latest results for the p-shell nuclei using the Tucson-Melbourne TM three-nucleon interaction with several proposed parameter sets.
Jung, Y. S.; Joo, H. G.; Yoon, J. I.
2013-07-01
The nTRACER direct whole core transport code employing the planar MOC solution based 3-D calculation method, the subgroup method for resonance treatment, the Krylov matrix exponential method for depletion, and a subchannel thermal/hydraulic calculation solver was developed for practical high-fidelity simulation of power reactors. Its accuracy and performance is verified by comparing with the measurement data obtained for three pressurized water reactor cores. It is demonstrated that accurate and detailed multi-physic simulation of power reactors is practically realizable without any prior calculations or adjustments. (authors)
Calculs d'assemblages de REP en environnement
NASA Astrophysics Data System (ADS)
Leroyer, Hadrien
Pressurized Water Reactors (PWR) are the most common nuclear reactor used today. The core of a PWR is composed of approximately 200 assemblies immersed in pressurized light water, which can be Uranium Oxyde assemblies (UOX) or Mixed Oxyde assemblies (MOX) coming from the reprocessings of used UOX. Electro-nuclear industries want to calculate the neutron flux inside these reactors, by solving the neutron transport equation, because it controls the dynamic of the core. Actually, the computers' power available today does not allow for a solution to the transport equation over the whole core, in three dimensions, with burnup. This is why reactor physicists use several approximations in order to obtain a solution for the neutron flux. This implies defining a pertinent calculation scheme. Generally, the calculation scheme requires homogenized macroscopic cross sections libraries generated using infinite lattice calculations on assemblies. Several parameters are used for the tabulation of these libraries including the burnup, the temperature and density of the coolant and the fuel, the concentration of boron and of xenon-135. Then the code evaluates the flux distribution in the finite reactor with the diffusion equation using cross sections interpolated from these libraries. However, the infinite lattice hypothesis may not be valid for highly heterogeneous cores, for example a core with burnup gradients or MOX / UOX interfaces. The purpose of this study is to evaluate the physical impact of heterogeneous environment on PWR assemblies. We first define a reference calculation scheme for a 3 x 3 assembly cluster, taking all heterogeneous environment effect into account, with the lattice cell code DRAGON. We later compare this reference with infinite lattice calculations, or with other calculation schemes closer to a full reactor calculation code. Those comparisons will allow us to explain physically the effects of the heterogeneous environment, and also to evaluate the errors in the reactor code committed when this effect is not taken into account. Finally, we will propose solutions to this issue.
Model biases in high-burnup fast reactor simulations
Touran, N.; Cheatham, J.; Petroski, R.
2012-07-01
A new code system called the Advanced Reactor Modeling Interface (ARMI) has been developed that loosely couples multiscale, multiphysics nuclear reactor simulations to provide rapid, user-friendly, high-fidelity full systems analysis. Incorporating neutronic, thermal-hydraulic, safety/transient, fuel performance, core mechanical, and economic analyses, ARMI provides 'one-click' assessments of many multi-disciplined performance metrics and constraints that historically require iterations between many diverse experts. The capabilities of ARMI are implemented in this study to quantify neutronic biases of various modeling approximations typically made in fast reactor analysis at an equilibrium condition, after many repetitive shuffles. Sensitivities at equilibrium that result in very high discharge burnup are considered ( and >20% FIMA), as motivated by the development of the Traveling Wave Reactor. Model approximations discussed include homogenization, neutronic and depletion mesh resolution, thermal-hydraulic coupling, explicit control rod insertion, burnup-dependent cross sections, fission product model, burn chain truncation, and dynamic fuel performance. The sensitivities of these approximations on equilibrium discharge burnup, k{sub eff}, power density, delayed neutron fraction, and coolant temperature coefficient are discussed. (authors)
NASA Astrophysics Data System (ADS)
Andersen, J. N.; Hennig, D.; Lundgren, E.; Methfessel, M.; Nyholm, R.; Scheffler, M.
1994-12-01
High resolution measurements are reported of the surface core-level shift of the 3d level for the Rh(111), Rh(110), Pd(111), Pd(110), and Ag(111) single-crystal surfaces. These measurements and earlier ones for the Mo(110), Rh(100), and Pd(100) surfaces are analyzed by ab initio calculations of the surface core-level shift. The calculations are found to reproduce well the trends of the experimental shifts with the 4d metal and with the crystal plane. The comparison between these experimental and theoretical results demonstrates the importance of proper inclusion of final-state effects for accurate calculations of surface core-level shifts. A core hole in a surface atom is found to be better screened than one in a bulk atom for the 4d metals to the left of Pd in the Periodic Table. The use of the Z+1 approximation to describe the core hole is investigated both by explicit use of this approximation and by performing calculations for 1s and 3d core holes, respectively. The Z+1 approximation is found to be well obeyed in the case of Ag whereas for the rest of the 4d transition metals it is less precise, introducing errors of typically 0.1 eV.
ATR PDQ and MCWO Fuel Burnup Analysis Codes Evaluation
G.S. Chang; P. A. Roth; M. A. Lillo
2009-11-01
The Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL) is being studied to determine the feasibility of converting it from the highly enriched Uranium (HEU) fuel that is currently uses to low enriched Uranium (LEU) fuel. In order to achieve this goal, it would be best to qualify some different computational methods than those that have been used at ATR for the past 40 years. This paper discusses two methods of calculating the burnup of ATR fuel elements. The existing method, that uses the PDQ code, is compared to a modern method that uses A General Monte Carlo N-Particle Transport Code (MCNP) combined with the Origen2.2 code. This modern method, MCNP with ORIGEN2.2 (MCWO), is found to give excellent agreement with the existing method (PDQ). Both of MCWO and PDQ are also in a very good agreement to the 235U burnup data generated by an analytical method.
ATR WG-MOX Fuel Pellet Burnup Measurement by Monte Carlo - Mass Spectrometric Method
Chang, Gray Sen I
2002-10-01
This paper presents a new method for calculating the burnup of nuclear reactor fuel, the MCWO-MS method, and describes its application to an experiment currently in progress to assess the suitability for use in light-water reactors of Mixed-OXide (MOX) fuel that contains plutonium derived from excess nuclear weapons material. To demonstrate that the available experience base with Reactor-Grade Mixed uranium-plutonium OXide (RGMOX) can be applied to Weapons-Grade (WG)-MOX in light water reactors, and to support potential licensing of MOX fuel made from weapons-grade 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. Fuel burnup is an important parameter needed for fuel performance evaluation. For the irradiated MOX fuel’s Post-Irradiation Examination, the 148Nd method is used to measure the burnup. The fission product 148Nd is an ideal burnup indicator, when appropriate correction factors are applied. In the ATR test environment, the spectrum-dependent and burnup-dependent correction factors (see Section 5 for detailed discussion) can be substantial in high fuel burnup. The validated Monte Carlo depletion tool (MCWO) used in this study can provide a burnup-dependent correction factor for the reactor parameters, such as capture-to-fission ratios, isotopic concentrations and compositions, fission power, and spectrum in a straightforward fashion. Furthermore, the correlation curve generated by MCWO can be coupled with the 239Pu/Pu ratio measured by a Mass Spectrometer (in the new MCWO-MS method) to obtain a best-estimate MOX fuel burnup. A Monte Carlo - MCWO method can eliminate the generation of few-group cross sections. The MCWO depletion tool can analyze the detailed spatial and spectral self-shielding effects in UO2, WG-MOX, and reactor-grade mixed oxide (RG-MOX) fuel pins. The MCWO-MS tool only needs the MS-measured 239Pu/Pu ratio, without the measured isotope 148Nd concentration data, to determine the burnup accurately. MCWO-MS not only provided linear heat generation rate, Pu isotopic composition versus burnup, and burnup distributions within the WG-MOX fuel capsules, but also correctly pointed out the inconsistency in the large difference in burnups obtained by the 148Nd method.
Recommendations for Addressing Axial Burnup in the PWR Burnup Credit Analyses
Wagner, J.C.
2002-10-23
This report presents studies performed to support the development of a technically justifiable approach for addressing the axial-burnup distribution in pressurized-water reactor (PWR) burnup-credit criticality safety analyses. The effect of the axial-burnup distribution on reactivity and proposed approaches for addressing the axial-burnup distribution are briefly reviewed. A publicly available database of profiles is examined in detail to identify profiles that maximize the neutron multiplication factor, k{sub eff}, assess its adequacy for PWR burnup credit analyses, and investigate the existence of trends with fuel type and/or reactor operations. A statistical evaluation of the k{sub eff} values associated with the profiles in the axial-burnup-profile database was performed, and the most reactive (bounding) profiles were identified as statistical outliers. The impact of these bounding profiles on k{sub eff} is quantified for a high-density burnup credit cask. Analyses are also presented to quantify the potential reactivity consequence of loading assemblies with axial-burnup profiles that are not bounded by the database. The report concludes with a discussion on the issues for consideration and recommendations for addressing axial burnup in criticality safety analyses using burnup credit for dry cask storage and transportation.
Burnup credit issues in transportation and storage
Brady, M. C.; Sanders, T. L.; Seager, K. D.; Lake, W. H.
1992-01-01
Reliance on the reduced reactivity of spent fuel for criticality control during transportation and storage is referred to as burnup credit. This concept has attracted international interest and is being actively pursued in the United States in the development of a new generation of transport casks. An overview of the US experience in developing a methodology to implement burnup credit in an integrated approach to transport cask design is presented in this paper. Specifically, technical issues related to the analysis, validation and implementation of burnup credit are identified and discussed.
Felfli, Z
2015-01-01
Core-polarization interactions are investigated in low-energy electron elastic scattering from the atoms In,Sn,Eu,Au and At through the calculation of their electron affinities. The complex angular momentum method wherein is embedded the vital electron-electron correlations is used. The core-polarization effects are studied through the well investigated rational function approximation of the Thomas-Fermi potential,which can be analytically continued into the complex plane. The EAs are extracted from the large resonance peaks in the calculated low-energy electron atom scattering total cross sections and compared with those from measurements and sophisticated theoretical methods. It is concluded that when the electron-electron correlation effects and core polarization interactions are accounted for adequately the importance of relativity on the calculation of the electron affinities of atoms can be assessed. For At, relativistic effects are estimated to contribute a maximum of about 3.6 percent to its (non-rela...
FaCE: a tool for three body Faddeev calculations with core excitation
NASA Astrophysics Data System (ADS)
Thompson, I. J.; Nunes, F. M.; Danilin, B. V.
2004-08-01
FaCE is a self contained program, with namelist input, that solves the three body Faddeev equations. It enables the inclusion of excitation of one of the three bodies, whilst the other two remain inert. It is particularly useful for obtaining the binding energies and bound state structure compositions of light exotic nuclei treated as three-body systems, given the three effective two body interactions. A large variety of forms for these interactions may be defined, and supersymmetric transformations of these potentials may be calculated whenever two body states need to be removed due to Pauli blocking. Program summaryTitle of program: FaCE (Faddeev with Core Excitation) Catalogue identifier: ADTW Program summary URL:http://cpc.cs.qub.ac.uk/summaries/ADTW Program obtainable from: CPC Program Library, Queen's University of Belfast, N. Ireland Computers: The code is designed to run on any Unix/Linux workstation or PC. Operating systems: Linux or UNIX Program language used: Fortran-90 Numerical libraries used: Source code for 6 routines from the NAG and BLAS libraries is included to enable independent compilation. Memory required to execute with typical data: 9 Mbytes of RAM memory and 12 MB of hard disk space. No. of bits in a word: 32 or 64 No. of bytes in distributed program, including test data, etc.: 116 514 No. of lines in distributed program, including test data, etc.: 15 574 Distribution format: tar gzip file Nature of physical problem: The program calculates eigenenergies and eigenstates for the three body problem by solving the Faddeev equations. Method of solution: Given the two body effective potentials it performs the supersymmetric transformation in case where there are forbidden states to be removed. The three body wavefunction is expanded in hyperspherical coordinates, the hyper-angular part is a series of Jacobi polynomials and the hyper-radial part is written in terms of a Laguerre basis. Within this basis the three body matrix elements are calculated and the full three body Hamiltonian matrix is completed. The diagonalization process is performed after various reductions (isospin, orthonormal and Feshbach) to determine the energies. Finally the three body wavefunction is reconstructed and other bound state observables are calculated. Typical running time: 6 s on a 1.7 GHz Intel P4-processor machine.
Randy G. Lott
2003-01-01
OAK (B204) The corrosion behavior of nuclear fuel cladding is a key factor limiting the performance of nuclear fuel elements, improved cladding alloys, which resist corrosion and radiation damage, will facilitate higher burnup core designs. The objective of this project is to understand the mechanisms by which alloy composition, heat treatment and microstructure affect corrosion rate. This knowledge can be
Characterization and modeling of high burn-up mixed oxide fuel
NASA Astrophysics Data System (ADS)
Teague, Melissa Christine
Currently, fast reactor performance is largely constrained by the limitations of the materials involved in these reactors. The fuel is particularly limiting due to fission gas generation, changes in thermal conductivity, microstructure changes within the fuel, fuel swelling, and fuel-cladding chemical interaction (FCCI). Highly irradiated fuel is radially inhomogeneous in composition, microstructure, and temperature. In this work, high burn-up mixed oxide fuel with local burn-ups of 3.4-23.7% FIMA were destructively examined as part of a research project to understand the performance of oxide fuel at extreme burn-ups. Optical metallography, transmission electron microscopy and electron back-scatter diffraction were performed to further study the microstructure and chemical composition of the irradiated fuel. The optical micrographs were used to generate finite-element meshes in order to model the effective thermal conductivity of the irradiated fuel as a function of burn-up, radial position, and temperature. The fuel-to-cladding gap closed significantly in samples with burn-ups below 7-9% FIMA. Samples with burn-ups in excess of 7-9% FIMA had a reopening of the fuel-to-cladding gap and evidence of joint oxide-gain formation. Additionally, high burn-up structure was observed in the two highest burn-up samples (23.7 and 21.3% FIMA). The microstructural modeling of the effective thermal conductivity found close (10-20%) agreement between the calculated effective thermal conductivities and the semi-empirical based analytical models, validating the finite-element mesoscale approach to microstructural modeling of effective thermal conductivities in irradiated fuel.
Liquid iron alloys at outer core conditions by first-principles calculation
NASA Astrophysics Data System (ADS)
Umemoto, K.; Hirose, K.
2014-12-01
Since the density of the outer core deduced from seismic data is about 10% lower than that of pure iron at core pressures and temperatures (P-T), it is widely believed that the outer core includes one or more light elements. The light element in the core, however, has not yet been identified. Comparison of the density and sound velocity of liquid iron alloys with observations, such as the PREM, is a promising way to determine the species and quantity of light alloying component(s) in the outer core. Here we report the results of a first-principles molecular dynamics study on liquid iron alloyed with different concentrations of sulfur and hydrogen, in order to understand the effects of these impurities on the liquid density and sound velocity under outer core P-T conditions. We discuss the composition of the Earth's outer core based on a comparison of the present results with the PREM density and velocity profiles.
NASA Astrophysics Data System (ADS)
Liu, Lang
2015-05-01
The unitary correlation operator method (UCOM) and the similarity renormalization group theory (SRG) are compared and discussed in the framework of the no-core Monte Carlo shell model (MCSM) calculations for 3H and 4He. The treatment of spurious center-of-mass motion by Lawson's prescription is performed in the MCSM calculations. These results with both transformed interactions show good suppression of spurious center-of-mass motion with proper Lawson's prescription parameter ?c.m. values. The UCOM potentials obtain faster convergence of total energy for the ground state than that of SRG potentials in the MCSM calculations, which differs from the cases in the no-core shell model calculations (NCSM). These differences are discussed and analyzed in terms of the truncation scheme in the MCSM and NCSM, as well as the properties of the potentials of SRG and UCOM. Supported by Fundamental Research Funds for the Central Universities (JUSRP1035), National Natural Science Foundation of China (11305077)
P. Maris; M. A. Caprio; J. P. Vary
2015-01-30
The emergence of rotational bands is observed in no-core configuration interaction (NCCI) calculations for the Be isotopes (7<=A<=12), as evidenced by rotational patterns for excitation energies, electromagnetic moments, and electromagnetic transitions. Yrast and low-lying excited bands are found. The results indicate well-developed rotational structure in NCCI calculations, using the JISP16 realistic nucleon-nucleon interaction within finite, computationally-accessible configuration spaces.
MCWO - Linking MCNP And ORIGEN2 For Fuel Burnup Analysis
Gray S Chang
2005-04-01
The UNIX BASH (Bourne Again Shell) script MCWO has been developed at the Idaho National Engineering and Environment Laboratory (INEEL) to couple the Monte Carlo transport code MCNP with the depletion and buildup code ORIGEN2. MCWO is a fully automated tool that links the Monte Carlo transport code MCNP with the radioactive decay and burnup code ORIGEN2. MCWO can handle a large number of fuel burnup and material loading specifications, Advanced Test Reactor (ATR) powers, and irradiation time intervals. The program processes input from the user that specifies the system geometry, initial material compositions, feed/removal specifications, and other code-specific parameters. Calculated results from MCNP, ORIGEN2, and data process module calculations are then output successively as the code runs. The principal function of MCWO is to transfer one-group cross-section and flux values from MCNP to ORIGEN2, and then transfer the resulting material compositions (after irradiation and/or decay) from ORIGEN2 back to MCNP in a repeated, cyclic fashion. The basic requirement of the code is that the user have a working MCNP input file and other input parameters; all interaction with ORIGEN2 and other calculations are performed by UNIX BASH script MCWO. This paper presents the MCWO-calculated results of the RERTR-1 and -2 , and the Weapons-Grade Mixed Oxiide fuel (Wg-MOX) fuel experiments in ATR and compares the MCWO-calculated results with the measured data.
All-electron Bethe-Salpeter calculations for shallow-core x-ray absorption near-edge structures
NASA Astrophysics Data System (ADS)
Olovsson, W.; Tanaka, I.; Mizoguchi, T.; Puschnig, P.; Ambrosch-Draxl, C.
2009-01-01
X-ray absorption near-edge structure spectra are calculated by fully solving the electron/core-hole Bethe-Salpeter equation (BSE) in an all-electron framework. We study transitions from shallow core states, including the Mg L2,3 edge in MgO, the Li K edge in the Li halides LiF, LiCl, LiBr, and LiI, as well as Li2O . We illustrate the advantage of the many-body approach over a core-hole supercell calculation. Both schemes lead to strongly bound excitons, but the nonlocal treatment of the electron-hole interaction in the BSE turns out to be crucial for an agreement with experiment.
Krasheninnikov, Arkady V.
2014-01-01
PHYSICAL REVIEW B 89, 035120 (2014) Electronic stopping power from first-principles calculations electronic stopping power Se of energetic ions in graphitic targets from first principles. By treating core into the dependence of the electronic stopping power Se on projectile velocity have been obtained with the explicit
Development of burnup dependent fuel rod model in COBRA-TF
NASA Astrophysics Data System (ADS)
Yilmaz, Mine Ozdemir
The purpose of this research was to develop a burnup dependent fuel thermal conductivity model within Pennsylvania State University, Reactor Dynamics and Fuel Management Group (RDFMG) version of the subchannel thermal-hydraulics code COBRA-TF (CTF). The model takes into account first, the degradation of fuel thermal conductivity with high burnup; and second, the fuel thermal conductivity dependence on the Gadolinium content for both UO2 and MOX fuel rods. The modified Nuclear Fuel Industries (NFI) model for UO2 fuel rods and Duriez/Modified NFI Model for MOX fuel rods were incorporated into CTF and fuel centerline predictions were compared against Halden experimental test data and FRAPCON-3.4 predictions to validate the burnup dependent fuel thermal conductivity model in CTF. Experimental test cases from Halden reactor fuel rods for UO2 fuel rods at Beginning of Life (BOL), through lifetime without Gd2O3 and through lifetime with Gd 2O3 and a MOX fuel rod were simulated with CTF. Since test fuel rod and FRAPCON-3.4 results were based on single rod measurements, CTF was run for a single fuel rod surrounded with a single channel configuration. Input decks for CTF were developed for one fuel rod located at the center of a subchannel (rod-centered subchannel approach). Fuel centerline temperatures predicted by CTF were compared against the measurements from Halden experimental test data and the predictions from FRAPCON-3.4. After implementing the new fuel thermal conductivity model in CTF and validating the model with experimental data, CTF model was applied to steady state and transient calculations. 4x4 PWR fuel bundle configuration from Purdue MOX benchmark was used to apply the new model for steady state and transient calculations. First, one of each high burnup UO2 and MOX fuel rods from 4x4 matrix were selected to carry out single fuel rod calculations and fuel centerline temperatures predicted by CTF/TORT-TD were compared against CTF /TORT-TD /FRAPTRAN predictions. After confirming that the new fuel thermal conductivity model in CTF worked and provided consistent results with FRAPTRAN predictions for a single fuel rod configuration, the same type of analysis was carried out for a bigger system which is the 4x4 PWR bundle consisting of 15 fuel pins and one control guide tube. Steady- state calculations at Hot Full Power (HFP) conditions for control guide tube out (unrodded) were performed using the 4x4 PWR array with CTF/TORT-TD coupled code system. Fuel centerline, surface and average temperatures predicted by CTF/TORT-TD with and without the new fuel thermal conductivity model were compared against CTF/TORT-TD/FRAPTRAN predictions to demonstrate the improvement in fuel centerline predictions when new model was used. In addition to that constant and CTF dynamic gap conductance model were used with the new thermal conductivity model to show the performance of the CTF dynamic gap conductance model and its impact on fuel centerline and surface temperatures. Finally, a Rod Ejection Accident (REA) scenario using the same 4x4 PWR array was run both at Hot Zero Power (HZP) and Hot Full Power (HFP) condition, starting at a position where half of the control rod is inserted. This scenario was run using CTF/TORT-TD coupled code system with and without the new fuel thermal conductivity model. The purpose of this transient analysis was to show the impact of thermal conductivity degradation (TCD) on feedback effects, specifically Doppler Reactivity Coefficient (DRC) and, eventually, total core reactivity.
Electronic Structure Calculations and Adaptation Scheme in Multi-core Computing Environments
Lakshminarasimhan Seshagiri; Masha Sosonkina; Zhao Zhang
2009-01-01
Multi-core processing environments have become the norm in the generic computing environment and are being considered for\\u000a adding an extra dimension to the execution of any application. The T2 Niagara processor is a very unique environment where\\u000a it consists of eight cores having a capability of running eight threads simultaneously in each of the cores. Applications\\u000a like General Atomic and
Neutron Spectrum Effects on Burnup, Reactivity, and Isotopics in UOâ\\/HâO Lattices
Xu Zhiwen; Michael J. Driscoll; Mujid S. Kazimi
2002-01-01
To provide guidance for future light water reactor core design and fuel management strategies, the effects of the moderator-to-fuel ratio on burnup, core endurance, and waste disposal characteristics have been investigated. The analysis is based on a unit cell model of the standard four-loop Westinghouse pressurized water reactor (PWR) with varied water density, rod diameter, and lattice pitch. Two state-of-the-art
A simple novel analysis procedure for IVR calculation in core-molten severe accident
Y. P. Zhang; S. Z. Qiu; G. H. Su; W. X. Tian
In-Vessel Retention (IVR) of core melt is a key severe accident management strategy adopted by operating nuclear power plants and advanced light water reactors (ALWRs), AP600, AP1000, etc. External Reactor Vessel Cooling (ERVC), which involves flooding the reactor cavity to submerge the reactor vessel in an attempt to cool core debris relocated to the vessel low head, is a novel
NASA Astrophysics Data System (ADS)
Vocadlo, L.; Martorell, B.; Brodholt, J. P.; Wood, I. G.
2014-12-01
Seismically determined S-wave velocities in the Earth's inner core are observed to be much lower (10-30%) than those generally inferred from mineral physics. This is a remarkably large discrepancy - mineralogical models for the mantle and the outer core match the observed velocities to around 1%. In no other large volume of the Earth does such a difference exist. There have been a number of arguments put forward over the years to account for the difference, but none have been universally accepted and our inability to explain the seismic velocities of the inner core remains an uncomfortable truth. Here, we present results from ab initio molecular dynamics calculations performed at 360 GPa and core temperatures on hcp and fcc iron, and on fcc-Fe alloyed with nickel and hcp-Fe alloyed with silicon. The calculated shear modulus, and therefore seismic velocities, of pure hcp-Fe reduces dramatically just prior to melting, providing an elegant explanation for the observed velocities. Calculations on fcc-Fe show no such strong reduction in VS, with a transformation to an hcp-type structure prior to melting; addition of 6.5 atm% and 13 atm% Ni to fcc-Fe raises the temperature of this transition. When silicon is added to hcp-Fe, the pre-melting behaviour is found to be very similar to that of pure hcp-Fe with a strong nonlinear shear weakening just before melting and a corresponding reduction in VS. Because temperatures range from T/Tm = 1 at the inner-outer core boundary to T/Tm ? 0.99 at the centre, this strong nonlinear effect on VS should occur in the inner core, providing a compelling explanation for the low VS observed.
NASA Astrophysics Data System (ADS)
Yoshii, Noriyuki; Okazaki, Susumu
2007-03-01
In our previous analysis of the structural stability of a sodium dodecyl sulfate (SDS) micelle based on molecular dynamics calculation, vacancies were found in the center of the micelles [N. Yoshii and S. Okazaki, Chem. Phys. Lett.425, 58 (2006)]. It is very interesting to clarify whether a water molecule is expected in the vacancy in thermodynamic equilibrium at room temperature. In order to investigate the stability of water in the core of micelle, free energy of transfer of water from bulk to the core has been calculated for the SDS micelle in water for two micelle sizes, N =61 and 121, at temperature T =300 K and pressure P =1 atm. The calculated free energy of transfer, ?Gc ?b, from the bulk to the core is about 28±4 kJ/mol and 26±4 kJ/mol for the micelle of the size N=61 and 121, respectively, where the corresponding Boltzmann factor, exp(-?Gc ?b/kT), is in the order of one over several ten thousands. Thus, a water molecule hardly permeates into the core of the micelle.
Mo-99 production at the Annular Core Research Reactor - recent calculative results
Parma, E.J.
1997-11-01
Significant progress has been made over the past year in understanding the chemistry and processing challenges associated with {sup 99}Mo production using Cintichem type targets. Targets fabricated at Los Alamos National Laboratory have been successfully irradiated in fuel element locations at the Annular Core Research Reactor (ACRR) and processed at the Sandia Hot Cell Facility. The next goal for the project is to remove the central cavity experiment tube from the reactor core, allowing for the irradiation of up to 37 targets. After the in-core work is complete, the reactor will be capable of producing significant quantities of {sup 99}Mo.
Production of {sup 99}Mo at the annular core research reactor-recent calculative results
Parma, E.J. [Sandia National Labs., Albuquerque, NM (United States)
1997-12-01
Significant progress has been made over the past year in understanding the chemistry and processing challenges associated with {sup 99}Mo production using Cintichem-type targets. Targets fabricated at Los Alamos National Laboratory have been successfully irradiated in fuel element locations at the Annular Core Research Reactor (ACRR) and processed at the Sandia Hot Cell Facility. The next goal for the project is to remove the central cavity experiment tube from the reactor core, allowing for the irradiation of up to 37 targets. After the in-core work is complete, the reactor will be capable of producing significant quantities of {sup 99}Mo.
Mai, V. T.; Fujii, T.; Wada, K.; Kitada, T.; Takaki, N.; Yamaguchi, A.; Watanabe, H.; Unesaki, H.
2012-07-01
Considering the importance of thorium data and concerning about the accuracy of Th-232 cross section library, a series of experiments of thorium critical core carried out at KUCA facility of Kyoto Univ. Research Reactor Inst. have been analyzed. The core was composed of pure thorium plates and 93% enriched uranium plates, solid polyethylene moderator with hydro to U-235 ratio of 140 and Th-232 to U-235 ratio of 15.2. Calculations of the effective multiplication factor, control rod worth, reactivity worth of Th plates have been conducted by MVP code using JENDL-4.0 library [1]. At the experiment site, after achieving the critical state with 51 fuel rods inserted inside the reactor, the measurements of the reactivity worth of control rod and thorium sample are carried out. By comparing with the experimental data, the calculation overestimates the effective multiplication factor about 0.90%. Reactivity worth of the control rods evaluation using MVP is acceptable with the maximum discrepancy about the statistical error of the measured data. The calculated results agree to the measurement ones within the difference range of 3.1% for the reactivity worth of one Th plate. From this investigation, further experiments and research on Th-232 cross section library need to be conducted to provide more reliable data for thorium based fuel core design and safety calculation. (authors)
A variational transport theory method for two-dimensional reactor core calculations
NASA Astrophysics Data System (ADS)
Mosher, Scott W.
It seems very likely that the next generation of reactor analysis methods will be based largely on neutron transport theory, at both the assembly and core levels. Significant progress has been made in recent years toward the goal of developing a transport method that is applicable to large, heterogeneous coarse-meshes. Unfortunately, the major obstacle hindering a more widespread application of transport theory to large-scale calculations is still the computational cost. In this dissertation, a variational heterogeneous coarse-mesh transport method has been extended from one to two-dimensional Cartesian geometry in a practical fashion. A generalization of the angular flux expansion within a coarse-mesh was developed. This allows a far more efficient class of response functions (or basis functions) to be employed within the framework of the original variational principle. New finite element equations were derived that can be used to compute the expansion coefficients for an individual coarse-mesh given the incident fluxes on the boundary. In addition, the non-variational method previously used to converge the expansion coefficients was developed in a new and more thorough manner by considering the implications of the fission source treatment imposed by the response expansion. The new coarse-mesh method was implemented for both one and two-dimensional (2-D) problems in the finite-difference, multigroup, discrete ordinates approximation. An efficient set of response functions was generated using orthogonal boundary conditions constructed from the discrete Legendre polynomials. Several one and two-dimensional heterogeneous light water reactor benchmark problems were studied. Relatively low-order response expansions were used to generate highly accurate results using both the variational and non-variational methods. The expansion order was found to have a far more significant impact on the accuracy of the results than the type of method. The variational techniques provide better accuracy, but at substantially higher computational costs. The non-variational method is extremely robust and was shown to achieve accurate results in the 2-D problems, as long as the expansion order was not very low.
Toru Yamamoto; Motomu Suzuki; Yoshihira Ando; Hiroaki Nagano
2012-01-01
The measured isotopic compositions of fuel samples taken from high-burnup spent PWR MOX and UO2 assemblies in the MALIBU program has been analyzed by lattice physics codes. The measured isotopes were U, Np, Pu, Am, and Cm isotopes and about 30 major fission product nuclides. The codes used in the present study were a continuous-energy Monte Carlo burnup calculation code
Yerokhin, V A
2012-01-01
Large-scale relativistic configuration-interaction calculation of energy levels of core-excited states of lithium-like ions is presented. Quantum electrodynamic, nuclear recoil, and frequency-dependent Breit corrections are included in the calculation. The approach is consistently applied for calculating all $n=2$ core-excited states for all lithium-like ions starting from argon ($Z = 18$) and ending with krypton ($Z = 36$). The results obtained are supplemented with systematical estimations of calculation errors and omitted effects.
ISOTOPIC MODEL FOR COMMERCIAL SNF BURNUP CREDIT
A.H. Wells
2004-11-17
The purpose of this report is to demonstrate a process for selecting bounding depletion parameters, show that they are conservative for pressurized water reactor (PWR) and boiling water reactor (BWR) spent nuclear fuel (SNF), and establish the range of burnup for which the parameters are conservative. The general range of applicability is for commercial light water reactor (LWR) SNF with initial enrichments between 2.0 and 5.0 weight percent {sup 235}U and burnups between 10 and 50 gigawatt-day per metric ton of uranium (GWd/MTU).
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.
Burnup verification tests with the FORK measurement system-implementation for burnup credit
Ewing, R.I.
1994-08-01
Verification measurements may be used to help ensure nuclear criticality safety when burnup credit is applied to spent fuel transport and storage systems. The FORK system measures the passive neutron and gamma-ray emission from spent fuel assemblies while in the storage pool. It was designed at Los Alamos National Laboratory for the International Atomic Energy Agency safeguards program and is well suited to verify burnup and cooling time records at commercial Pressurized Water Reactor (PWR) sites. This report deals with the application of the FORK system to burnup credit operations.
1984-01-01
BEAGL-01 is a computer program for calculating the conditions in a light water reactor core at steady state and while undergoing a transient. It couples a neutron kinetics model in R,Z geometry with a two-phase thermal-hydraulics model for multiple parallel channels. Appropriate applications include reactivity insertion accidents in both pressurized and boiling water reactors and transients in boiling water reactors
The calculation of a collapse of iron-oxygen stellar core in one-group energy approximation
A. G. Aksenov; D. K. Nadyozhin
1998-01-01
Results of the calculation of a gravitational collapse of the iron-oxygen core of a star with mass 2M_solar by the new method proposed by one of the authors (Aksenov 1998a) are presented. This method incorporates the integration of hydrodynamic equations of matter, equations of -processes and thermonuclear reactions kinetics, and the solution of transfer equations for different kinds of neutrino
Dependence of transuranic content in spent fuel on fuel burnup
Reese, Drew A. (Drew Amelia)
2007-01-01
As the increasing demand for nuclear energy results in larger spent fuel volume, implementation of longer fuel cycles incorporating higher burnup are becoming common. Understanding the effect of higher burnup on the spent ...
Hay, P.J.; Wadt, W.R.
1985-01-01
Ab initio effective core potentials (ECP's) have been generated to replace the Coulomb, exchange, and core-orthogonality effects of the chemically inert core electron in the transition metal atoms Sc to Hg. For the second and third transition series relative ECP's have been generated which also incorporate the mass--velocity and Darwin relativistic effects into the potential. The ab initio ECP's should facilitate valence electron calculations on molecules containing transition-metal atoms with accuracies approaching all-electron calculations at a fraction of the computational cost. Analytic fits to the potentials are presented for use in multicenter integral evaluation. Gaussian orbital valence basis sets are developed for the (3d,4s,4p), (4d,5s,5p), and (5d,6s,6p) orbitals of the first, second, and third transition series atoms, respectively. All-electron and valence-electron atomic excitation energies are also compared for the low-lying states of Sc--Hg, and the valence-electron calculations are found to reproduce the all-electron excitation energies (typically within a few tenths of an eV).
Sensitivity and Uncertainty Analysis to Burnup Estimates on ADS using the ACAB Code
Cabellos, O.; Sanz, J.; Rodriguez, A.; Gonzalez, E.; Embid, M.; Alvarez, F.; Reyes, S.
2005-05-24
Within the scope of the Accelerator Driven System (ADS) concept for nuclear waste management applications, the burnup uncertainty estimates due to uncertainty in the activation cross sections (XSs) are important regarding both the safety and the efficiency of the waste burning process. We have applied both sensitivity analysis and Monte Carlo methodology to actinides burnup calculations in a lead-bismuth cooled subcritical ADS. The sensitivity analysis is used to identify the reaction XSs and the dominant chains that contribute most significantly to the uncertainty. The Monte Carlo methodology gives the burnup uncertainty estimates due to the synergetic/global effect of the complete set of XS uncertainties. These uncertainty estimates are valuable to assess the need of any experimental or systematic re-evaluation of some uncertainty XSs for ADS.
An Investigation of Irradiation Performance of High Burnup HTGR Fuel
Kazuhiro SAWA; Kazuo MINATO
1999-01-01
In order to investigate fuel behavior under high burnup irradiation condition of high temperature gas-cooled reactor (HTGR), an irradiation test was performed. An irradiation was carried out as a part of a cooperative effort between the US DOE and the Japan Atomic Energy Research Institute. The fuel for the irradiation test was called high burnup fuel, whose target burnup and
Gauntt, Randall O.; Ross, Kyle W. (Los Alamos National Laboratory, Los Alamos, NM); Smith, James Dean; Longmire, Pamela
2010-04-01
The Oak Ridge National Laboratory computer code, ORIGEN2.2 (CCC-371, 2002), was used to obtain the elemental composition of irradiated low-enriched uranium (LEU)/mixed-oxide (MOX) pressurized-water reactor fuel assemblies. Described in this report are the input parameters for the ORIGEN2.2 calculations. The rationale for performing the ORIGEN2.2 calculation was to generate inventories to be used to populate MELCOR radionuclide classes. Therefore the ORIGEN2.2 output was subsequently manipulated. The procedures performed in this data reduction process are also described herein. A listing of the ORIGEN2.2 input deck for two-cycle MOX is provided in the appendix. The final output from this data reduction process was three tables containing the radionuclide inventories for LEU/MOX in elemental form. Masses, thermal powers, and activities were reported for each category.
Mo99 production at the Annular Core Research Reactor - recent calculative results
Parma
1997-01-01
Significant progress has been made over the past year in understanding the chemistry and processing challenges associated with â¹â¹Mo production using Cintichem type targets. Targets fabricated at Los Alamos National Laboratory have been successfully irradiated in fuel element locations at the Annular Core Research Reactor (ACRR) and processed at the Sandia Hot Cell Facility. The next goal for the project
NASA Astrophysics Data System (ADS)
Chi, C.-C.; Hsiao, C.-H.; Skoropata, E.; van Lierop, J.; Ouyang, Chuenhou Hao
2015-05-01
Significant efforts towards understanding bi-magnetic core-shell nanoparticles are underway currently as they provide a pathway towards properties unavailable with single-phased systems. Recently, we have demonstrated that the magnetism of ?-Fe2O3/CoO core-shell nanoparticles, in particular, at high temperatures, originates essentially from an interfacial doped iron-oxide layer that is formed by the migration of Co2+ from the CoO shell into the surface layers of the ?-Fe2O3 core [Skoropata et al., Phys. Rev. B 89, 024410 (2014)]. To examine directly the nature of the intermixed layer, we have used high-resolution transmission electron microscopy (HRTEM) and first-principles calculations to examine the impact of the core-shell intermixing at the atomic level. By analyzing the HRTEM images and energy dispersive spectra, the level and nature of intermixing was confirmed, mainly as doping of Co into the octahedral site vacancies of ?-Fe2O3. The average Co doping depths for different processing temperatures (150 °C and 235 °C) were 0.56 nm and 0.78 nm (determined to within 5% through simulation), respectively, establishing that the amount of core-shell intermixing can be altered purposefully with an appropriate change in synthesis conditions. Through first-principles calculations, we find that the intermixing phase of ?-Fe2O3 with Co doping is ferromagnetic, with even higher magnetization as compared to that of pure ?-Fe2O3. In addition, we show that Co doping into different octahedral sites can cause different magnetizations. This was reflected in a change in overall nanoparticle magnetization, where we observed a 25% reduction in magnetization for the 235 °C versus the 150 °C sample, despite a thicker intermixed layer.
Stout, R.B.; Merckx, K.R.; Holm, J.S.
1981-01-01
This study calculates the reduced uranium requirements and the economic incentives for increasing the burnup of current design LWR fuels from the current range of 25 to 35 MWD/Kg to a range of 45 to 55 MWD/Kg. The changes in fuel management strategies which may be required to accommodate these high burnup fuels and longer fuel cycles are discussed. The material behavior problems which may present obstacles to achieving high burnup or to license fuel are identified and discussed. These problems are presented in terms of integral fuel response and the informational needs for commercial and licensing acceptance. Research and development programs are outlined which are aimed at achieving a licensing position and commercial acceptance of high burnup fuels.
NASA Astrophysics Data System (ADS)
Botazzoli, Pietro; Luzzi, Lelio; Brémier, Stephane; Schubert, Arndt; Van Uffelen, Paul; Walker, Clive T.; Haeck, Wim; Goll, Wolfgang
2011-12-01
The TRANSURANUS burn-up model (TUBRNP) calculates the local concentration of the actinides, the main fission products, and 4He as a function of the radial position across a fuel rod. In this paper, the improvements in the helium production model as well as the extensions in the simulation of 238-242Pu, 241Am, 243Am and 242-245Cm isotopes are described. Experimental data used for the extended validation include new EPMA measurements of the local concentrations of Nd and Pu and recent SIMS measurements of the radial distributions of Pu, Am and Cm isotopes, both in a 3.5% enriched commercial PWR UO 2 fuel with a burn-up of 80 and 65 MWd/kgHM, respectively. Good agreement has been found between TUBRNP and the experimental data. The analysis has been complemented by detailed neutron transport calculations (VESTA code), and also revealed the need to update the branching ratio for the 241Am(n,?) 242mAm reaction in typical PWR conditions.
NASA Astrophysics Data System (ADS)
Pinto, R. M.; Dias, A. A.; Coreno, M.; de Simone, M.; Giuliano, B. M.; Santos, J. P.; Costa, M. L.
2011-11-01
The relative populations of the 1H- and 2H-tautomer of gas-phase 5-methyltetrazole (5MTZ) have been assessed through core-level photoelectron spectroscopy, and compared with the results obtained from GAUSSIAN-n (Gn, n = 1, 2 and 3) and Complete Basis Set methods (CBS-4M and CBS-Q). The C 1s and N 1s core-electron binding energies (CEBEs) for each ionization site of both tautomers have been computed using the ?self-consistent-field (?SCF) approach. The C 1s and N 1s XPS spectra, obtained at 313 K, yield a 1H/2H tautomer ratio of ca. 0.16/0.84 and 0.21/0.79, respectively.
Production of â¹â¹Mo at the annular core research reactor-recent calculative results
Parma
1997-01-01
Significant progress has been made over the past year in understanding the chemistry and processing challenges associated with â¹â¹Mo production using Cintichem-type targets. Targets fabricated at Los Alamos National Laboratory have been successfully irradiated in fuel element locations at the Annular Core Research Reactor (ACRR) and processed at the Sandia Hot Cell Facility. The next goal for the project is
Guerin, P.; Baudron, A. M.; Lautard, J. J.
2006-07-01
This paper describes a new technique for determining the pin power in heterogeneous core calculations. It is based on a domain decomposition with overlapping sub-domains and a component mode synthesis technique for the global flux determination. Local basis functions are used to span a discrete space that allows fundamental global mode approximation through a Galerkin technique. Two approaches are given to obtain these local basis functions: in the first one (Component Mode Synthesis method), the first few spatial eigenfunctions are computed on each sub-domain, using periodic boundary conditions. In the second one (Factorized Component Mode Synthesis method), only the fundamental mode is computed, and we use a factorization principle for the flux in order to replace the higher order Eigenmodes. These different local spatial functions are extended to the global domain by defining them as zero outside the sub-domain. These methods are well-fitted for heterogeneous core calculations because the spatial interface modes are taken into account in the domain decomposition. Although these methods could be applied to higher order angular approximations - particularly easily to a SPN approximation - the numerical results we provide are obtained using a diffusion model. We show the methods' accuracy for reactor cores loaded with UOX and MOX assemblies, for which standard reconstruction techniques are known to perform poorly. Furthermore, we show that our methods are highly and easily parallelizable. (authors)
Simulating the Dynamics of Earth's Core: Using NCCS Supercomputers Speeds Calculations
NASA Technical Reports Server (NTRS)
2002-01-01
If one wanted to study Earth's core directly, one would have to drill through about 1,800 miles of solid rock to reach liquid core-keeping the tunnel from collapsing under pressures that are more than 1 million atmospheres and then sink an instrument package to the bottom that could operate at 8,000 F with 10,000 tons of force crushing every square inch of its surface. Even then, several of these tunnels would probably be needed to obtain enough data. Faced with difficult or impossible tasks such as these, scientists use other available sources of information - such as seismology, mineralogy, geomagnetism, geodesy, and, above all, physical principles - to derive a model of the core and, study it by running computer simulations. One NASA researcher is doing just that on NCCS computers. Physicist and applied mathematician Weijia Kuang, of the Space Geodesy Branch, and his collaborators at Goddard have what he calls the,"second - ever" working, usable, self-consistent, fully dynamic, three-dimensional geodynamic model (see "The Geodynamic Theory"). Kuang runs his model simulations on the supercomputers at the NCCS. He and Jeremy Bloxham, of Harvard University, developed the original version, written in Fortran 77, in 1996.
González-Cataldo, F. [Grupo de NanoMateriales, Departamento de Física, Facultad de Ciencias, Universidad de Chile, Casilla 653, Santiago (Chile); Wilson, Hugh F.; Militzer, B., E-mail: fgonzalez@lpmd.cl [Department of Earth and Planetary Science, University of California Berkeley, Berkeley, CA 94720 (United States)
2014-05-20
By combining density functional molecular dynamics simulations with a thermodynamic integration technique, we determine the free energy of metallic hydrogen and silica, SiO{sub 2}, at megabar pressures and thousands of degrees Kelvin. Our ab initio solubility calculations show that silica dissolves into fluid hydrogen above 5000 K for pressures from 10 and 40 Mbars, which has implications for the evolution of rocky cores in giant gas planets like Jupiter, Saturn, and a substantial fraction of known extrasolar planets. Our findings underline the necessity of considering the erosion and redistribution of core materials in giant planet evolution models, but they also demonstrate that hot metallic hydrogen is a good solvent at megabar pressures, which has implications for high-pressure experiments.
Study on physics characteristics of Th-U fuel in long-cycle core
Yu, G.; Wang, K. [Dept. of Engineering Physics, Tsinghua Univ., Beijing, 100084 (China)
2006-07-01
Based on the investigation of recent development of the long-life small nuclear power systems in the world, this paper puts forward a new concept of building a long-life reactor core with Thorium-Uranium fuel and Lead-Bismuth coolant and conducts a series of research on the characteristics of the Th-U fuel. In the physics design of a long-life reactor, it is important to keep a small reactivity swing with burnup and a deep burnup core, as well as negative temperature and void reactivity coefficient. The program MCBurn are used in the calculation of a pin cell model with different initial driver fuel, p/d ratio, enrichment and fuel type to obtain some physical results such as neutron spectrum, the reactivity swing with burnup and initial conversion ratio, etc. Based on the analysis of these results, this paper concludes some physical requirements of a long-life reactor core and constructs a preliminary core design with some safety parameters such as void reactivity coefficient and other results. (authors)
Neutron Spectrum Effects on Burnup, Reactivity, and Isotopics in UO{sub 2}/H{sub 2}O Lattices
Xu Zhiwen; Driscoll, Michael J.; Kazimi, Mujid S. [Massachusetts Institute of Technology (United States)
2002-07-15
To provide guidance for future light water reactor core design and fuel management strategies, the effects of the moderator-to-fuel ratio on burnup, core endurance, and waste disposal characteristics have been investigated. The analysis is based on a unit cell model of the standard four-loop Westinghouse pressurized water reactor (PWR) with varied water density, rod diameter, and lattice pitch. Two state-of-the-art computer codes, CASMO-4 and MOCUP (MCNP+ORIGEN), have been used. Considering the entire range of moderation (from fast spectra to overthermalized spectra), the results show that higher reactivity-limited burnup is achievable by either a wetter lattice or a much drier lattice than normal. In particular, epithermal lattices are distinctly inferior performers. Current PWR lattices are about the optimum in terms of highest fuel endurance. However, wetter lattices produce less plutonium with a degraded plutonium isotopic mix with respect to weapons usability. Neptunium-237 content is only mildly affected by the hydrogen-to-heavy-metal ratio. High burnup is significantly beneficial to reducing plutonium production per unit energy and to making its isotopic mix less attractive as a weapon material. In particular, the {sup 238}Pu to {sup 239}Pu ratio increases approximately as the 2.5 power of burnup for a fixed initial enrichment. Based on this neutronics study, wetter lattices are recommended for future high-burnup applications.
Multidimensional transition-state theory calculations for nuclear dynamics of core-excited molecules
Carniato, Stephane; Gallet, Jean-Jacques [Laboratoire de Chimie Physique, Matiere et Rayonnement, UMR 7614, Universite Pierre et Marie Curie, 11, rue Pierre et Marie Curie, 75231 Paris Cedex 05 (France); Ilakovac, Vita [Laboratoire de Chimie Physique, Matiere et Rayonnement, UMR 7614, Universite Pierre et Marie Curie, 11, rue Pierre et Marie Curie, 75231 Paris Cedex 05 (France); Universite de Cergy Pontoise, 95031, Cergy Pontoise Cedex (France); Kukk, Edwin [Department of Physical Sciences, University of Oulu, P.O. Box 3000, FIN-90014, Oulu (Finland); Luo Yi [Theoretical Chemistry, Royal Institute of Technology, AlbaNova, S-106 91 Stockholm (Sweden)
2004-09-01
We have extended the transition-state theory to describe the dynamics of core hole excitation. This allows us to interpret the abnormally bland near-edge x-ray absorption fine structure spectrum of the gas phase benzonitrile molecule at the N 1s edge. We have brought to light different paths for the two most intensive resonances, going from the linear to the bent structure. The profile of each resonance consists of two different vibrational progressions corresponding to stretching modes and a broad continuum of bending excited states.
Three-Dimensional Full Core Power Calculations for Pressurized Water Reactors
Evans, Thomas M [ORNL] [ORNL; Davidson, Gregory G [ORNL] [ORNL; Slaybaugh, Rachel N [ORNL] [ORNL
2010-01-01
We have implemented a new multilevel parallel decomposition in the Denovo discrete ordinates radiation transport code. In concert with Krylov subspace iterative solvers, the multilevel decomposition allows concurrency over energy in addition to space-angle. The original space-angle partitioning in Denovo placed an eective limit on the scalability of the transport solver that was highly dependent on the problem size. The added phase-space concurrency combined with the high-performance Krylov solvers has enabled weak scaling to 100K cores on the Jaguar XT5 supercomputer. Furthermore, the multilevel decomposition provides enough concurrency to scale to exascale computing and beyond.
Reactor physics calculations for {sup 99}Mo production at the annular core research reactor
Parma, E.J. [Sandia National Labs., Albuquerque, NM (United States)
1995-12-31
The Isotope Production and Distribution Program at the U.S. Department of Energy has designated Sandia National Laboratories (SNL) as the most appropriate facility for the production of {sup 99}Mo, a radioisotope whose daughter, {sup 99m}Tc, is used in more than 36,000 medical procedures per day in the United States and is considered to be a vital medical diagnostic and treatment tool. The isotope would be produced at SNL using the annular core research reactor (ACRR) facility and collocated hot cell facility. The {sup 99}Mo would be produced using the fission process by irradiating {open_quotes}targets{close_quotes} coated with {sup 235}U in the form of highly enriched U{sub 3}O{sub 8}. After {approximately}7 days of continuous irradiation in the ACRR, a target would be re- moved from the reactor core for processing. The isotope would be extracted by chemically precipitating the molybdenum using the {open_quotes}Cintichem{close_quotes} process and would be shipped to the various pharmaceutical companies by commercial or chartered airline.
Some Nuclear Calculations of U-235-D2O Gaseous-Core Cavity Reactors
NASA Technical Reports Server (NTRS)
Ragsdale, Robert G.; Hyland, Robert E.
1961-01-01
The results of a multigroup, diffusion theory study of spherical gaseous-core cavity reactors are presented in this report. The reactor cavity of gaseous U235 is enclosed by a region of hydrogen gas and is separated from an external D2O moderator-reflector by a zirconium structural shell. Some cylindrical reactors are also investigated. A parametric study of spherical reactors indicates that, for the range of variables studied, critical mass increases as: (1) Fuel region is compressed within the reactor cavity, (2) moderator thickness is decreased, (3) structural shell thickness is increased, and (4) moderator temperature is increased. A buckling analogy is used to estimate the critical mass of fully reflected cylindrical reactors from spherical results without fuel compression. For a reactor cavity of a 120-centimeter radius uniformly filled with fuel, no structural shell, a moderator temperature of 70 F, and a moderator thickness of 100 centimeters, the critical mass of a spherical reactor is 3.1 kilograms while that of a cylinder with a length-to-diameter ratio of 1.0 (L/D = 1) is approximately 3.8 kilograms and, with L/D = 2, 5.9 kilograms. For the range of variables considered for U235-D2O gaseous-core cavity reactors, the systems are characterized by 95 to 99 percent thermal absorptions, with the flux reaching a maximum in the moderator about 10 to 15 centimeters from the reactor cavity.
Ab initio shell-model calculation for ^{18}O in a restricted no-core model space
S. Fujii; B. R. Barrett
2009-02-12
We perform an ab initio shell-model calculation for ^{18}O in a restricted no-core model space, microscopically deriving a two-body effective interaction and introducing a minimal refinement of one-body energies in the spsd or spsdpf model space. Low-lying energy levels, except for the experimental 0_{2}^{+} and 2_{3}^{+} states, are better described in the spsdpf space than in the spsd space. The structure of low-lying energy levels is discussed with an emphasis on many-particle many-hole states beyond the four-particle two-hole configuration.
NASA Astrophysics Data System (ADS)
Matthiä, Daniel; Meier, Matthias M.; Reitz, Günther
2014-03-01
The increased radiation exposure at aviation altitudes is of public interest as well as of legal relevance in many countries. The dose rates that are elevated compared to sea level are mainly caused by galactic cosmic ray particles interacting with the atmosphere and producing a complex radiation field at aviation altitudes. The intensity and composition of this radiation field mainly depend on altitude, geomagnetic shielding, and primary particle intensity. In this work, we present a model based on Monte Carlo simulations, which retrospectively estimates secondary particle fluence as well as ambient dose equivalent rates and effective dose rates at any point in the atmosphere. This model will be used as the physical core in the Professional Aviation Dose Calculator (PANDOCA) software developed by the German Aerospace Center (Deutsches Zentrum für Luft- und Raumfahrt) for the calculation of route doses in aviation. The calculations are based on galactic cosmic ray spectra taking into account primary nuclei from hydrogen to iron by direct transport calculations of hydrogen and helium nuclei and approximating heavier nuclei by the number of protons equaling the corresponding atomic number. A comparison to experimental data recorded on several flights with a tissue equivalent proportional counter shows a very good agreement between model calculations and measurements.
Calculation of the yields of water radiolysis products from fast-neutrons in a vver core
A. V. Gordeev; B. G. Ershov
1992-01-01
The great demands to the reactor parameters made it necessary to improve the calculations in physics and heat engineering and to solve many complicated problems, including the analysis of emergency situations, which can be mastered only with modern computers. Owing to the complexity of the plant's function, a full-scale test-stand prototype was built and new solutions were developed on it;
Adaptive Calculation of a Collapsing Molecular Cloud Core: The Jeans Condition
NASA Astrophysics Data System (ADS)
Sigalotti, Leonardo Di G.; Klapp, Jaime
In 1997 Truelove et al. introduced the Jeans condition to determine what level of spatial resolution is needed to avoid artificial fragmentation during protostellar collapse calculations. They first found using a Cartesian code based on an adaptive mesh refinement (AMR) technique that a Gaussian cloud model collapsed isothermally to form a singular filament rather than a binary or quadruple protostellar system as predicted by previous calculations. Recently Boss et al. in 2000 using a different hydrodynamics code with high spatial resolution reproduced the filamentary collapse solution of Truelove et al., implying that high resolution coupled with the Jeans condition is necessary to perform reliable calculations of the isothermal protostellar collapse. Here we recalculate the isothermal Gaussian cloud model of Truelove et al. and Boss et al. using a completely different code based on zooming coordinates to achieve the required high spatial resolution. We follow the collapse through 7 orders of magnitude increase in density and reproduce the filamentary solution. With the zooming coordinates, we are allowed to perform an adaptive calculation with a much lower computational cost than the AMR technique and other grid redefinition methods.
Parametric Study of the Effect of Burnable Poison Rods for PWR Burnup Credit
Wagner, J.C.
2001-09-28
The Interim Staff Guidance on burnup credit (ISG-8) issued by the United States Nuclear Regulatory Commission's (U.S. NRC) Spent Fuel Project Office recommends restricting the use of burnup credit to assemblies that have not used burnable absorbers. This recommended restriction eliminates a large portion of the currently discharged spent fuel assemblies from cask loading, and thus severely limits the practical usefulness of burnup credit. In the absence of readily available information on burnable poison rod (BPR) design specifications and usage in U.S. pressurized-water-reactors (PWRs), and the subsequent reactivity effect of BPR exposure on discharged spent nuclear fuel (SNF), NRC staff has indicated a need for additional information in these areas. In response, this report presents a parametric study of the effect of BPR exposure on the reactivity of SNF for various BPR designs, fuel enrichments, and exposure conditions, and documents BPR design specifications. Trends in the reactivity effects of BPRs are established with infinite pin-cell and assembly array calculations with the SCALE and HELIOS code packages, respectively. Subsequently, the reactivity effects of BPRs for typical initial enrichment and burnup combinations are quantified based on three-dimensional (3-D) KENO V.a Monte Carlo calculations with a realistic rail-type cask designed for burnup credit. The calculations demonstrate that the positive reactivity effect due to BPR exposure increases nearly linearly with burnup and is dependent on the number, poison loading, and design of the BPRs and the initial fuel enrichment. Expected typical reactivity increases, based on one-cycle BPR exposure, were found to be less than 1% {Delta}k. Based on the presented analysis, guidance is offered on an appropriate approach for calculating bounding SNF isotopic data for assemblies exposed to BPRs. Although the analyses do not address the issue of validation of depletion methods for assembly designs with BPRs, they do demonstrate that the effect of BPRs is generally well behaved and that independent codes and cross-section libraries predict similar results. The report concludes with a discussion of the issues for consideration and recommendations for inclusion of SNF assemblies exposed to BPRs in criticality safety analyses using burnup credit for dry cask storage and transport.
NASA Astrophysics Data System (ADS)
Wenzel, Jan; Holzer, Andre; Wormit, Michael; Dreuw, Andreas
2015-06-01
The extended second order algebraic-diagrammatic construction (ADC(2)-x) scheme for the polarization operator in combination with core-valence separation (CVS) approximation is well known to be a powerful quantum chemical method for the calculation of core-excited states and the description of X-ray absorption spectra. For the first time, the implementation and results of the third order approach CVS-ADC(3) are reported. Therefore, the CVS approximation has been applied to the ADC(3) working equations and the resulting terms have been implemented efficiently in the adcman program. By treating the ? and ? spins separately from each other, the unrestricted variant CVS-UADC(3) for the treatment of open-shell systems has been implemented as well. The performance and accuracy of the CVS-ADC(3) method are demonstrated with respect to a set of small and middle-sized organic molecules. Therefore, the results obtained at the CVS-ADC(3) level are compared with CVS-ADC(2)-x values as well as experimental data by calculating complete basis set limits. The influence of basis sets is further investigated by employing a large set of different basis sets. Besides the accuracy of core-excitation energies and oscillator strengths, the importance of cartesian basis functions and the treatment of orbital relaxation effects are analyzed in this work as well as computational timings. It turns out that at the CVS-ADC(3) level, the results are not further improved compared to CVS-ADC(2)-x and experimental data, because the fortuitous error compensation inherent in the CVS-ADC(2)-x approach is broken. While CVS-ADC(3) overestimates the core excitation energies on average by 0.61% ± 0.31%, CVS-ADC(2)-x provides an averaged underestimation of -0.22% ± 0.12%. Eventually, the best agreement with experiments can be achieved using the CVS-ADC(2)-x method in combination with a diffuse cartesian basis set at least at the triple-? level.
Wenzel, Jan; Holzer, Andre; Wormit, Michael; Dreuw, Andreas
2015-06-01
The extended second order algebraic-diagrammatic construction (ADC(2)-x) scheme for the polarization operator in combination with core-valence separation (CVS) approximation is well known to be a powerful quantum chemical method for the calculation of core-excited states and the description of X-ray absorption spectra. For the first time, the implementation and results of the third order approach CVS-ADC(3) are reported. Therefore, the CVS approximation has been applied to the ADC(3) working equations and the resulting terms have been implemented efficiently in the adcman program. By treating the ? and ? spins separately from each other, the unrestricted variant CVS-UADC(3) for the treatment of open-shell systems has been implemented as well. The performance and accuracy of the CVS-ADC(3) method are demonstrated with respect to a set of small and middle-sized organic molecules. Therefore, the results obtained at the CVS-ADC(3) level are compared with CVS-ADC(2)-x values as well as experimental data by calculating complete basis set limits. The influence of basis sets is further investigated by employing a large set of different basis sets. Besides the accuracy of core-excitation energies and oscillator strengths, the importance of cartesian basis functions and the treatment of orbital relaxation effects are analyzed in this work as well as computational timings. It turns out that at the CVS-ADC(3) level, the results are not further improved compared to CVS-ADC(2)-x and experimental data, because the fortuitous error compensation inherent in the CVS-ADC(2)-x approach is broken. While CVS-ADC(3) overestimates the core excitation energies on average by 0.61% ± 0.31%, CVS-ADC(2)-x provides an averaged underestimation of -0.22% ± 0.12%. Eventually, the best agreement with experiments can be achieved using the CVS-ADC(2)-x method in combination with a diffuse cartesian basis set at least at the triple-? level. PMID:26049476
Assessment of Reactivity Margins and Loading Curves for PWR Burnup Credit Cask Designs
Wagner, J.C.
2002-12-17
This report presents studies to assess reactivity margins and loading curves for pressurized water reactor (PWR) burnup-credit criticality safety evaluations. The studies are based on a generic high-density 32-assembly cask and systematically vary individual calculational (depletion and criticality) assumptions to demonstrate the impact on the predicted effective neutron multiplication factor, k{sub eff}, and burnup-credit loading curves. The purpose of this report is to provide a greater understanding of the importance of input parameter variations and quantify the impact of calculational assumptions on the outcome of a burnup-credit evaluation. This study should provide guidance to regulators and industry on the technical areas where improved information will most enhance the estimation of accurate subcritical margins. Based on these studies, areas where future work may provide the most benefit are identified. The report also includes an evaluation of the degree of burnup credit needed for high-density casks to transport the current spent nuclear fuel inventory. By comparing PWR discharge data to actinide-only based loading curves and determining the number of assemblies that meet the loading criteria, this evaluation finds that additional negative reactivity (through either increased credit for fuel burnup or cask design/utilization modifications) is necessary to accommodate the majority of current spent fuel assemblies in high-capacity casks. Assemblies that are not acceptable for loading in the prototypic high-capacity cask may be stored or transported by other means (e.g., lower capacity casks that utilize flux traps and/or increased fixed poison concentrations or high-capacity casks with design/utilization modifications).
Benefits of the delta K of depletion benchmarks for burnup credit validation
Lancaster, D.; Machiels, A.
2012-07-01
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{sub 2} critical experiments to determine the criticality safety limits on the neutron multiplication factor, k{sub 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)
Tikhomirov, A. V.; Ponomarenko, G. L. [OKB GIDROPRESS, Podolsk (Russian Federation)
2012-07-01
An additional verification of bundled software (BS) SAPFIR-95 and amp;RC [1] and code KORSAR/GP [2] was performed. Both software products were developed in A.P. Alexandrov NITI and certified by ROSTEKHNADZOR of RF for numeric simulation of stationary, transitional and emergency conditions of VVER reactors. A benchmark model for neutronics calculations was created within the limits of this work. The cold subcritical state of VVER - 1000 reactor stationary fuelling was simulated on the basis of FA with an increased height of the fuel column (TVS-2M) considering detailed presentation of radial and front neutron reflectors. A case of passing of pure condensate slug through the core in initially deep subcritical state during start of the first RCP set after refueling was considered as an examined condition of reactor operation. A relatively small size of the slug, its spatial position near the reflectors (lower and lateral), as well as failure of the inserted control rods of the control and protection system (CPS CR) to reach the lower limit of the fuel column stipulate for methodical complexity of a correct calculation of the neutron multiplication constant (K{sub eff}) using engineering codes. Code RC was used as a test program in the process of reactor calculated 3-D modeling. Code MCNP5 [3] was used as the precision program, which solves the equation of neutrons transfer by Monte-Carlo method and which was developed in the US (Los-Alamos). As a result of comparative calculations dependency of K{sub eff} on two parameters was evaluated - boron acid concentration (Cb) and CPS CR position. Reactivity effect was evaluated, which is implemented as a result of failure of all CPS control rods to reach the lower fuel limit calculated using the engineering codes mentioned above. (authors)
A new boiling water reactor core concept for a next-generation light water reactor
J. Yamashita; A. Nishimura; T. Mochida; O Yokomizo
1991-01-01
In this paper a boiling water reactor (BWR) core concept that meets various requirements for a next-generation light water reactor is proposed. This BWR core can be operated as either a high-burnup core or a high-conversion core simply by replacing the fuel assemblies and control rods. The high-burnup core is suitable for a once-through nuclear fuel cycle and has a
ABRAC: A microcomputer-based Fortran code for multi-cyle burnup
Olson, A.P.
1990-01-01
Pressurized-water reactors have reactor physics and fuel management characteristics which are very amenable to simplified analysis. Given models which account for the dominant features of core and fuel performance, it is possible to rapidly perform relatively accurate scoping studies of many years of reactor operation in just a few hours on a modern (386-class) microcomputer. Models are described for burnup-dependent cross-section generation, for burnup of fuel under irradiation, and for computation of radial power distributions in hexagonal geometry assuming hexagonal fuel assemblies. Comparisons with more elaborate methods are given in order to validate the models, and to quantify the accuracy of the results. 16 refs., 5 figs., 5 tabs.
Porter, Troy A
2013-01-01
Cosmic dust particles effectively attenuate starlight. Their absorption of starlight produces emission spectra from the near- to far-infrared, which depends on the sizes and properties of the dust grains, and spectrum of the heating radiation field. The near- to mid-infrared is dominated by the emissions by very small grains. Modeling the absorption of starlight by these particles is, however, computationally expensive and a significant bottleneck for self-consistent radiation transport codes treating the heating of dust by stars. In this paper, we summarize the formalism for computing the stochastic emissivity of cosmic dust, which was developed in earlier works, and present a new library HEATCODE implementing this formalism for the calculation for arbitrary grain properties and heating radiation fields. Our library is highly optimized for general-purpose processors with multiple cores and vector instructions, with hierarchical memory cache structure. The HEATCODE library also efficiently runs on co-processo...
Niskanen, Johannes; Arul Murugan, N; Rinkevicius, Zilvinas; Vahtras, Olav; Li, Cui; Monti, Susanna; Carravetta, Vincenzo; Agren, Hans
2013-01-01
We report hybrid density functional theory-molecular mechanics (DFT/MM) calculations performed for glycine in water solution at different pH values. In this paper, we discuss several aspects of the quantum mechanics-molecular mechanics (QM/MM) simulations where the dynamics and spectral binding energy shifts are computed sequentially, and where the latter are evaluated over a set of configurations generated by molecular or Car-Parrinello dynamics simulations. In the used model, core ionization takes place in glycine as a quantum mechanical (QM) system modeled with DFT, and the solution is described with expedient force fields in a large molecular mechanical (MM) volume of water molecules. The contribution to the core electronic binding energy from all interactions within and between the two (DFT and MM) parts is accounted for, except charge transfer and dispersion. While the obtained results were found to be in qualitative agreement with experiment, their precision must be qualified with respect to the problem of counter ions, charge transfer and optimal division of QM and MM parts of the system. Results are compared to those of a recent study [Ottoson et al., J. Am. Chem. Soc., 2011, 133, 3120]. PMID:23160171
S?4 Reactor: Operating Lifetime and Estimates of Temperature and Burnup Reactivity Coefficients
NASA Astrophysics Data System (ADS)
King, Jeffrey C.; El-Genk, Mohamed S.
2006-01-01
The S?4 reactor has a sectored, Mo-14%Re solid core for avoidance of single point failures in reactor cooling and Closed Brayton Cycle (CBC) energy conversion. The reactor is loaded with UN fuel, cooled with a He-Xe gas mixture at ~1200 K and operates at steady thermal power of 550 kW. Following a launch abort accident, the axial and radial BeO reflectors easily disassemble upon impact so that the bare reactor is subcriticial when submerged in wet sand or seawater and the core voids are filled with seawater. Spectral Shift Absorber (SSA) additives have been shown to increase the UN fuel enrichment and significantly reduce the total mass of the reactor. This paper investigates the effects of SSA additions on the temperature and burnup reactivity coefficients and the operational lifetime of the S?4 reactor. SSAs slightly decrease the temperature reactivity feedback coefficient, but significantly increase the operating lifetime by decreasing the burnup reactivity coefficient. With no SSAs, fuel enrichment is only 58.5 wt% and the estimated operating lifetime is the shortest (7.6 years) with the highest temperature and burnup reactivity feedback coefficients (-0.2709 ¢/K and -1.3470 $/atom%). With europium-151 and gadolinium-155 additions, the enrichment (91.5 and 94 wt%) and operating lifetime (9.9 and 9.8 years) of the S?4 reactor are the highest while the temperature and burnup reactivity coefficients (-0.2382 and -0.2447 ¢/K -0.9073 and 0.8502 $/atom%) are the lowest.
Lashkari, A; Khalafi, H; Kazeminejad, H
2013-05-01
In this work, kinetic parameters of Tehran research reactor (TRR) mixed cores have been calculated. The mixed core configurations are made by replacement of the low enriched uranium control fuel elements with highly enriched uranium control fuel elements in the reference core. The MTR_PC package, a nuclear reactor analysis tool, is used to perform the analysis. Simulations were carried out to compute effective delayed neutron fraction and prompt neutron lifetime. Calculation of kinetic parameters is necessary for reactivity and power excursion transient analysis. The results of this research show that effective delayed neutron fraction decreases and prompt neutron lifetime increases with the fuels burn-up. Also, by increasing the number of highly enriched uranium control fuel elements in the reference core, the prompt neutron lifetime increases, but effective delayed neutron fraction does not show any considerable change. PMID:24976672
Ji-Young Lee; Jung-Hwan Chang; Do-Hyun Kang; Sung-Il Kim; Jung-Pyo Hong
2006-01-01
Transverse flux linear motor (TFLM) is getting to be widely used in low speed and high power systems. For easy fabrication, solid core is more useful than laminated core. However, eddy current losses are increased in solid core, and it makes efficiency decreased. Some experiment results show the fact. Therefore, estimation of eddy current losses is important to select core
ATR WG-MOX Fuel Pellet Burnup Measurement by Monte Carlo - Mass Spectrometric Method
G. S. Chang; Gray Sen I
2002-01-01
This paper presents a new method for calculating the burnup of nuclear reactor fuel, the MCWO-MS method, and describes its application to an experiment currently in progress to assess the suitability for use in light-water reactors of Mixed-OXide (MOX) fuel that contains plutonium derived from excess nuclear weapons material. To demonstrate that the available experience base with Reactor-Grade Mixed uranium-plutonium
Analyzing the rod drop accident in a BWR with high burnup fuel
D. J. Diamond; L. Neymotin
1997-01-01
The response of fuel in a boiling water reactor to the rod drop accident (RDA) was studied using the RAMONA-4B computer code. Calculations of this design-basis event has been done conservatively because there was margin to the fuel failure criterion of 170 cal\\/g. Because high burnup fuel may fail at much lower fuel enthalpies, the best-estimate of the enthalpy and
Effect of microconstants uncertainty on transplutonium actinides burnup in molten salt reactors
A. A. Myasnikov
2007-01-01
An evaluation method and results for the error due to microconstants uncertainties in the calculation of neptunium and transplutonium\\u000a actinide burnup in a molten-salt reactor are presented. The method developed treats the characteristics of a reactor in an\\u000a equilibrium state and assumes that Np, Am, Cm, and other transplutonium elements as well as material for maintaining criticality\\u000a are fed continually
DANDE: a linked code system for core neutronics/depletion analysis
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.
Saavedra, Steven F [ORNL; Charlton, William S [Texas A& M University; Solodov, Alexander A [ORNL; Ehinger, Michael H [ORNL
2010-01-01
Globally, there exists a long history in reprocessing in evaluation of the shipper/receiver difference (SRD) on spent nuclear fuel (SNF) received and processed. Typically, the declared shipper s values for uranium and plutonium in SNF (based on calculations involving the initial manufacturer s data and reactor operating history) are used as the input quantities to the head-end process of the facility. Problems have been encountered when comparing these values with measured results of the input accountability tank contents. A typical comparison yields a systematic bias indicated as a loss of 5 7 percent of the plutonium (Pu) and approximately 1 percent for the uranium (U). Studies suggest that such deviation can be attributed to the non-linear nature of the axial burnup values of the SNF. Oak Ridge National Laboratory and Texas A&M University are co-investigating the development of a new method, via Nondestructive Assay (NDA) techniques, to improve the accuracy in burnup and Pu content quantification. Two major components have been identified to achieve this objective. The first component calculates a measurement-based burnup profile along the axis of a fuel rod. Gamma-ray data is collected at numerous locations along the axis of the fuel rod using a High Purity Germanium (HPGe) detector designed for a wide range of gamma-ray energies. Using two fission products, 137Cs and 134Cs, the burnup is calculated at each measurement location and a profile created along the axis of the rod based on the individual measurement locations. The second component measures the U/Pu ratio using an HPGe detector configured for relatively low-energy gamma-rays including x-rays. Fluorescence x-rays from U and Pu are measured and compared to the U/Pu ratio determined from a destructive analysis of the sample. This will be used to establish a relationship between the measured and actual values. This relationship will be combined with the burnup analysis results to establish a relationship between fission product activity and Pu content. It is anticipated that this new method will allow receiving facilities to make a limited number of NDA, gamma-ray, measurements to confirm the shipper declared values for burnup and Pu content thereby improving the SRD.
Fission gas release during power bumping at high burnup
M. E. Cunningham; M. D. Freshley; D. D. Lanning
1991-01-01
Research to define the behavior of Zircaloy-clad light water reactor fuel irradiated to high burnup levels has been conducted by the High Burnup Effects Program (HBEP), an international, group-sponsored research effort conducted by Battelle, Pacific Northwest Laboratories. One activity conducted by the HBEP was to investigate fission gas release (FGR) from commercially irradiated pressurized water reactor (PWR) and boiling water
Microstructural analysis of LWR spent fuels at high burnup
NASA Astrophysics Data System (ADS)
Thomas, L. E.; Beyer, C. E.; Chariot, L. A.
1992-06-01
The microstructural changes in commercial light-water reactor (LWR) fuels irradiated to average burnups near 50 MWd/kgM were studied by analytical transmission electron microscopy and Auger electron spectrometry. Several poorly understood aspects of the fuel behavior were examined, including (a) precipitation of the fission gases in dense, highly pressurized inclusions, (b) apparent solution of Cs, Ba, Zr, and Te in the UO 2 matrix, and (c) the "rim effect" involving restructuring of the enhanced burnup region at the fuel outer edges. Initial observations of the high-burnup rim showed an extremely fine-grained structure formed by recrystallization of the original UO 2. The restructuring is a burnup-induced instability of the UO 2, possibly driven by the stored energy of fission products in solution, and is expected to extend across LWR fuel pellets irradiated to higher burnups.
Uncertainties in the effects of burnup and their impact on criticality safety licensing criteria
Carlson, R.W.; Fisher, L.E.
1990-07-13
Current criteria for criticality safety for spent fuel shipping and storage casks are conservative because no credit is permitted for the effects of burnup of the fuel inside the cask. Cask designs that will transport and store large numbers of fuel assemblies (20 or more) must devote a substantial part of their payload to criticality control measures if they are to meet this criteria. The Department of Energy is developing the data necessary to support safety analyses that incorporate the effects of burnup for the next generation of spent fuel shipping casks. The efforts described here are devoted to the development of acceptance criteria that will be the basis for accepting safety analyses. Preliminary estimates of the uncertainties of the effects of burnup have been developed to provide a basis for the consideration of critically safety criteria. The criticality safety margins in a spent fuel shipping or storage cask are dominated by the portions of a fuel assembly that are in low power regions of a reactor core, and the reactor operating conditions are very different from spent fuel storage or transport cask conditions. Consequently, the experience that has been gathered during years of reactor operation does not apply directly to the prediction of criticality safety margins for spent fuel shipping or storage casks. The preliminary estimates of the uncertainties presented in this paper must be refined by both analytical and empirical studies that address both the magnitude of the uncertainties and their interdependence. 9 refs., 5 figs.
Preparation of higher-actinide burnup and cross section samples. [LMFBR
Adair, H.L.; Kobisk, E.H.; Quinby, T.C.; Thomas, D.K.; Dailey, J.M.
1981-01-01
A joint research program involving the United States and the United Kingdom was instigated about four years ago for the purpose of studying burnup of higher actinides using in-core irradiation in the fast reactor at Dounreay, Scotland. Simultaneously, determination of cross sections of a wide variety of higher actinide isotopes was proposed. Coincidental neutron flux and energy spectral measurements were to be made using vanadium encapsulated dosimetry materials in the immediate region of the burnup and cross section samples. The higher actinide samples chosen for the burnup study were /sup 241/Am and /sup 244/Cm in the forms of Am/sub 2/O/sub 3/, Cm/sub 2/O/sub 3/, and Am/sub 6/ Cm(RE)/sub 7/O/sub 21/, where (RE) represents a mixture of lanthanide sesquioxides. It is the purpose of this paper to describe technology development and its application in the preparation of the fuel specimens and the cross section specimens that are being used in this cooperative program.
NASA Astrophysics Data System (ADS)
Giglio, J. J.; Cummings, D. G.; Michlik, M. M.; Goodall, P. S.; Johnson, S. G.
1997-02-01
The determination of burnup, an indicator of fuel cycle efficiency, has been accomplished by the determination of 139La by high-resolution inductively coupled plasma atomic emission spectroscopy (HR-ICP-AES). Solutions of digested samples of reactor fuel rods were introduced into a shielded glovebox housing an inductively-coupled plasma (ICP) and the resulting atomic emission transmitted to a high-resolution spectrometer by a 31 m fiber optic bundle. Total and isotopic U determination by thermal ionization mass spectrometry (TIMS) is presented to allow for the calculation of burnup for the samples. This method of burnup determination reduces the time, material, sample handling and waste generation associated with typical burnup determinations which require separation of lanthanum from the other fission products with high specific activities.
NASA Astrophysics Data System (ADS)
Chung, Wen-Hung; Wang, Chia-Ching; Tsai, Dah-Shyang; Jiang, Jyh-Chiang; Cheng, Yu-Chang; Fan, Liang-Jen; Yang, Yaw-Wen; Huang, Ying-Sheng
2010-01-01
Deoxygenation of the IrO 2(1 1 0) surface is investigated at 403-493 K, using the core-level spectroscopy and density functional theory (DFT) calculation. The Ir-4f 7/2 signals of 1f-cus-Ir with and without on-top oxygen (O top) emerge as surface features of the baked-out surface, whose positive and negative shifts in binding energy are in line with the DFT computation results. Progressively increasing the reduction temperature, the 1f-cus-Ir feature quickly disappears and the signal of 2f-cus-Ir emerges at 403 K. Meanwhile the feature of 1f-cus-Ir + O top diminishes but persists when the Ir metal signal is evident. The intriguing coexistence of 1f-cus-Ir + O top and Ir metal at 433-443 K is elucidated in the theoretical pathway study. DFT calculation reveals that O 2 desorption via pairing two neighboring O top atoms is the rate-determining step of surface deoxygenation. Under the UHV conditions, O top is replenished via migration of the surface oxygen species, including the threefold coordinated oxygen (O 3f) of a reduced surface. Hence the O top atom is an active and long-lived surface species, which does not vanish until O 3f is consumed and surface Ir begins to cluster. Under the realistic pressure conditions, O top can also be refreshed via the dissociative adsorption of gas-phase oxygen. In either pathway, O top is a critical intermediary of IrO 2(1 1 0) oxidation catalysis.
Needs of reliable nuclear data and covariance matrices for Burnup Credit in JEFF-3 library
NASA Astrophysics Data System (ADS)
Chambon, A.; Santamarina, A.; Riffard, C.; Lavaud, F.; Lecarpentier, D.
2013-03-01
Burnup Credit (BUC) is the concept which consists in taking into account credit for the reduction of nuclear spent fuel reactivity due to its burnup. In the case of PWR-MOx spent fuel, studies pointed out that the contribution of the 15 most absorbing, stable and non-volatile fission products selected to the credit is as important as the one of the actinides. In order to get a "best estimate" value of the keff, biases of their inventory calculation and individual reactivity worth should be considered in criticality safety studies. This paper enhances the most penalizing bias towards criticality and highlights possible improvements of nuclear data for the 15 FPs of PWRMOx BUC. Concerning the fuel inventory, trends in function of the burnup can be derived from experimental validation of the DARWIN-2.3 package (using the JEFF-3.1.1/SHEM library). Thanks to the BUC oscillation programme of separated FPs in the MINERVE reactor and fully validated scheme PIMS, calculation over experiment ratios can be accurately transposed to tendencies on the FPs integral cross sections.
Methodology for the Weapons-Grade MOX Fuel Burnup Analysis in the Advanced Test Reactor
G. S. Chang
2005-08-01
A UNIX BASH (Bourne Again SHell) script CMO has been written and validated at the Idaho National Laboratory (INL) to couple the Monte Carlo transport code MCNP with the depletion and buildup code ORIGEN-2 (CMO). The new Monte Carlo burnup analysis methodology in this paper consists of MCNP coupling through CMO with ORIGEN-2, and is therefore called the MCWO. MCWO is a fully automated tool that links the Monte Carlo transport code MCNP with the radioactive decay and burnup code ORIGEN-2. MCWO is capable of handling a large number of fuel burnup and material loading specifications, Advanced Test Reactor (ATR) lobe powers, and irradiation time intervals. MCWO processes user input that specifies the system geometry, initial material compositions, feed/removal specifications, and other code-specific parameters. Calculated results from MCNP, ORIGEN-2, and data process module calculations are output in succession as MCWO executes. The principal function of MCWO is to transfer one-group cross-section and flux values from MCNP to ORIGEN-2, and then transfer the resulting material compositions (after irradiation and/or decay) from ORIGEN-2 back to MCNP in a repeated, cyclic fashion. The basic requirements of MCWO are a working MCNP input file and some additional input parameters; all interaction with ORIGEN-2 as well as other calculations are performed by CMO. This paper presents the MCWO-calculated results for the Reduced Enrichment Research and Test Reactor (RERTR) experiments RERTR-1 and RERTR-2 as well as the Weapons-Grade Mixed Oxide (WG-MOX) fuel testing in ATR. Calculations performed for the WG-MOX test irradiation, which is managed by the Oak Ridge National Laboratory (ORNL), supports the DOE Fissile Materials Disposition Program (FMDP). The MCWO-calculated results are compared with measured data.
Development of HELIOS/CAPP code system for the analysis of block type VHTR cores
Lee, H. C.; Han, T. Y.; Jo, C. K.; Noh, J. M. [Korea Atomic Energy Research Inst., 989-111, Daedeok-Daero, Yuseong-Gu, Daejeon (Korea, Republic of)
2012-07-01
In this paper, the HELIOS/CAPP code system developed for the analysis of block type VHTR cores is presented and verified against several VHTR core configurations. Verification results shows that HELIOS code predicts less negative MTC and RTC than McCARD code does and thus HELIOS code overestimates the multiplication factors at the states with high moderator and reflector temperature especially when the B{sub 4}C BP is loaded. In the depletion calculation for the VHTR single cell fuel element, the error of HELIOS code increases as burnup does. It is ascribed to the fact that HELIOS code treats some fission product nuclides with large resonances as non-resonant nuclides. In the 2-D core depletion calculation, a relatively large reactivity error is observed in the case with BP loading while the reactivity error in the case without BP loading is less than 300 pcm. (authors)
Navarro, J.; Aryaeinejad, R.; Nigg, D.W. [Idaho National Laboratory, P. O. Box 1625, Idaho Falls, ID 83415 (United States)
2011-07-01
The goal of this work was to perform a feasibility study and establish measurement techniques to determine the burnup of the Advanced Test Reactor (ATR) fuels at the Idaho National Laboratory (INL). Three different detectors of high purity germanium (HPGe), lanthanum bromide (LaBr{sub 3}), and high pressure xenon (HPXe) in two detection system configurations of below and above the water pool were used in this study. The last two detectors were used for the first time in fuel burnup measurements. The results showed that a better quality spectra can be achieved with the above the water pool configuration. Both short and long cooling time fuels were investigated in order to determine which measurement technique, absolute or fission product ratio, is better suited in each scenario and also to establish what type of detector should be used in each case for the best burnup measurement. The burnup and cooling time calibrations were established using experimental absolute activities or isotopic ratios and ORIGEN burnup calculations. A method was developed to do burnup and cooling time calibrations using fission isotopes activities without the need to know the exact geometry. (authors)
Interpretation of High-Burnup Fuel Annealing Tests
Paul BLAIR; Grigori KHVOSTOV; Antonino ROMANO; Christian HELLWIG; Rakesh CHAWLA
2008-01-01
The growth of the porosity in high-burnup fuel is of particular interest when considering the effect of fission gas retention within the high-burnup structure (HBS). A mechanistic model of porosity growth under annealing conditions for light water reactor (LWR) UO2 fuel with typical stereological parameters of the HBS has been developed. The model takes into account both multipore and surface
Transverse buckling effects on solitary burn-up waves
Xue-Nong Chen; Werner Maschek
2005-01-01
A three-dimensional one-group diffusion model with explicit effects of burnup and feedback is studied for a so-called “candle reactor”. By a perturbation method the problem is reduced to a one-dimensional one, for which a solitary wave solution was obtained by van Dam (2000) [Self-stabilizing criticality waves. Annals of Nuclear Energy 27, 1505]. Therefore, such a travelling burn-up wave exists as
Portable gamma-ray holdup and attributes measurements of high- and variable-burnup plutonium
Wenz, T.R.; Russo, P.A.; Miller, M.C.; Menlove, H.O. (Los Alamos National Lab., NM (United States)); Takahashi, S.; Yamamoto, Y.; Aoki, I. (Power Reactor and Nuclear Fuel Development Corp., Tokai, Ibaraki (Japan))
1991-01-01
High burnup-plutonium holdup has been assayed quantitatively by low resolution gamma-ray spectrometry. The assay was calibrated with four plutonium standards representing a range of fuel burnup and {sup 241}Am content. Selection of a calibration standard based on its qualitative spectral similarity to gamma-ray spectra of the process material is partially responsible for the success of these holdup measurements. The spectral analysis method is based on the determination of net counts in a single spectral region of interest (ROI). However, the low-resolution gamma-ray assay signal for the high-burnup plutonium includes unknown amounts of contamination from {sup 241}Am. For most needs, the range of calibration standards required for this selection procedure is not available. A new low-resolution gamma-ray spectral analysis procedure for assay of {sup 239}Pu has been developed. The procedure uses the calculated isotope activity ratios and the measured net counts in three spectral ROIs to evaluate and remove the {sup 241}Am contamination from the {sup 239}Pu assay signal on a spectrum-by-spectrum basis. The calibration for the new procedure requires only a single plutonium standard. The procedure also provides a measure of the burnup and age attributes of holdup deposits. The new procedure has been demonstrated using portable gamma-ray spectroscopy equipment for a wide range of plutonium standards and has also been applied to the assay of {sup 239}Pu holdup in a mixed oxide fuel fabrication facility. 10 refs., 5 figs., 3 tabs.
NASA Astrophysics Data System (ADS)
Li, Qingxiang; Gandhi, Om P.
2005-01-01
Compliance testing of electronic article surveillance (EAS) devices requires that induced current densities in central nervous system (CNS) tissues, i.e. brain and the spinal cord, be less than the prescribed safety limits. Even though ferromagnetic cores are mostly used for activation/deactivation of embedded magnetic tags, assumed equivalent air-core coils with guessed increased number of ampere turns have always been used to calculate the magnetic fields for the proximal region to which a customer is exposed. We show that at low frequencies up to several kilohertz, duality of electric and magnetic circuits may be exploited such that the shaped high reluctance core is modelled as though it was a higher conductivity electric circuit of the corresponding shape. The proposed procedure is tested by examples of two magnetic cores typical of countertop activation/deactivation devices. The equivalent exposure magnetic fields obtained from the dual electric fields are shown to be in excellent agreement (within ±5%) with those measured for these ferromagnetic EAS devices. The previously proposed impedance method is then used to calculate the induced current densities for a 1.974 × 1.974 × 2.93 mm resolution anatomic model of a human. For the two considered EAS systems using excitation currents of 5000 A turns at 200 Hz, the maximum 1 cm2 area-averaged induced current densities in the CNS tissues are calculated and found to be less than the ICNIRP safety limits.
Dieudonne, C.; Dumonteil, E.; Malvagi, F.; Diop, C. M.
2013-07-01
For several years, Monte Carlo burnup/depletion codes have appeared, which couple a Monte Carlo code to simulate the neutron transport to a deterministic method that computes the medium depletion due to the neutron flux. Solving Boltzmann and Bateman equations in such a way allows to track fine 3 dimensional effects and to get rid of multi-group hypotheses done by deterministic solvers. The counterpart is the prohibitive calculation time due to the time-expensive Monte Carlo solver called at each time step. Therefore, great improvements in term of calculation time could be expected if one could get rid of Monte Carlo transport sequences. For example, it may seem interesting to run an initial Monte Carlo simulation only once, for the first time/burnup step, and then to use the concentration perturbation capability of the Monte Carlo code to replace the other time/burnup steps (the different burnup steps are seen like perturbations of the concentrations of the initial burnup step). This paper presents some advantages and limitations of this technique and preliminary results in terms of speed up and figure of merit. Finally, we will detail different possible calculation scheme based on that method. (authors)
Segev, M.; Galperin, A.; Schwageraus, E. [Ben Gurion University of the Negev (Israel)
2000-07-15
Shortly after the loading of a pressurized water reactor (PWR) core, the axial power distribution in fresh fuel has already attained the characteristic, almost flat shape, typical of burned fuel. At beginning of cycle (BOC), however, the axial distribution is centrally peaked. In assemblies hosting uniform burnable boron rods, this BOC peaking is even more pronounced. A reduction in the axial peaking is today often achieved by shortening the burnable boron rods by some 30 cm at each edge.It is shown that a two-zone grading of the boron rod leads, in a representative PWR cycle, to a reduction of the hot-spot temperature of {approx}70 deg. C, compared with the nongraded case. However, with a proper three-zone grading of the boron rod, an additional 20 deg. C may be cut off the hot-spot temperature. Further, with a slightly skewed application of this three-zone grading, an additional 50 deg. C may be cut off.The representative PWR cycle studied was cycle 11 of the Indian Point 2 station, with a simplification in the number of fuel types and in the burnup distribution. The analysis was based on a complete three-dimensional burnup calculation. The code system was ELCOS, with BOXER as an assembly code for the generation of burnup-dependent cross sections and SILWER as a three-dimensional core code with thermal-hydraulic feedback.
Burn-up and neutron economy of accelerator-driven reactor
Takahashi, H.; Yang, W.; An, Y.; Yamazaki, Y.
1997-07-01
It is desirable to have only a small reactivity change in the large burn-up of a solid fuel fast reactor, so that the number of replacements or shuffling of the fuel can be reduced, and plant factor accordingly increased. Also, this reduces the number of control rods needed for the change in burn-up reactivity. In subcritical operation, power controlled by beam power is suggested, but this practice is not as economical as the use of control rods and makes more careful operation of the accelerator is required due to changes in the wake field. In subcritical operation, even a slightly subcritical one, the safety problems associated with a hard neutron spectrum can be alleviated. Neutron leakage from a flattened core, which is needed for operation of the critical fast reactor can be lessen by using the non flat core which has good neutron economy. For generating nuclear energy, it is essential to have a high neutron economy, although breeding the fuel is not welcomed in the present political climate, as is needed for transmuting long lived fission products. In contrast to the breeder, the accelerator driven reactor can separate the energy production from fuel production and processing. Thus, it is suited for non-proliferation of nuclear material by prohibiting the processing and production of fuel in the unrestricted area so this can be only done in international controlled areas which are restricted and remote.
Analysis of MNSR core composition changes using the codes WIMSD-4 and CITATION.
Haj Hassan, H; Ghazi, N; Hainoun, A
2008-10-01
The codes WIMSD/4 and BORGES--part of the MTR-PC code package--have been applied to prepare the microscopic cross-section library for the main elements of miniature neutron source reactor (MNSR) core for six neutron energy groups. The generated library has been utilized by the 3D code CITATION to perform the calculation of fuel burn-up including the identification of main fission products and their impacts on the multiplication factor. In this regard some modifications have been introduced to the subroutine NUCY in CITATION to incorporate estimating the concentration of the related actinides and fission products. The burn-up results have indicated that the core life-time of MNSR is being mainly estimated by Sm(149) followed by Gd(157) and Cd(113). The accumulation of these fission products during 100 continuous operation days caused a reduction of about 4.3 mk for the excess reactivity. This result seems to be in good agreement with the available empirical value of 3.5 mk, which relates to the whole discontinuous operation period of the reactor since its start up to now. The calculation procedure simulates the sporadic operation with an equivalent continuous operation period. This approximation is valid for the long-lived fission products that mainly dictate the core life-time. However, it is an overestimation for the concentration of short-lived radioactive products like Xe(135). PMID:18547812
Verification of the ORIGEN2 code analysis for the TMI-2 reactor core
Akers, D.W.; Schnitzler, B.G.
1988-01-01
Accurate definition of the fission product inventories produced in the TMI-2 reactor prior to the accident on March 29, 1979 are of considerable interest to many organizations including the Department of Energy which is shipping the damage reactor core to the Idaho National Engineering Laboratory and conducting the TMI-2 reactor examination program, and General Public Utilities which is defueling the reactor. Numerous fission product inventory calculations have been performed for the TMI-2 core, including an ORIGEN2 analysis by EG G which uses 1239 nodes to define burnup in the reactor core. To provide a verification of the predicted fission product inventories, a measurement study was performed using pellets from various core regions. Measurements were performed for transuranics, burnup monitors, noble gases, principal gamma ray emitters, {sup 129}I, and {sup 90}Sr. Comparisons between the experimental results and the code analyses are presented with an evaluation of the associated uncertainties. Also, a discussion is presented of the probable causes of the observed differences between the code and measured values. 10 refs., 1 fig., 2 tabs.
Terry, William Knox; Gougar, Hans D; Ougouag, Abderrafi Mohammed-El-Ami
2002-07-01
A new deterministic method has been developed for the neutronics analysis of a pebble-bed reactor (PBR). The method accounts for the flow of pebbles explicitly and couples the flow to the neutronics. The method allows modeling of once-through cycles as well as cycles in which pebbles are recirculated through the core an arbitrary number of times. This new work is distinguished from older methods by the systematically semi-analytical approach it takes. In particular, whereas older methods use the finite-difference approach (or an equivalent one) for the discretization and the solution of the burnup equation, the present work integrates the relevant differential equation analytically in discrete and complementary sub-domains of the reactor. Like some of the finite-difference codes, the new method obtains the asymptotic fuel-loading pattern directly, without modeling any intermediate loading pattern. This is a significant advantage for the design and optimization of the asymptotic fuel-loading pattern. The new method is capable of modeling directly both the once-through-then-out fuel cycle and the pebble recirculating fuel cycle. Although it currently includes a finite-difference neutronics solver, the new method has been implemented into a modular code that incorporates the framework for the future coupling to an efficient solver such as a nodal method and to modern cross section preparation capabilities. In its current state, the deterministic method presented here is capable of quick and efficient design and optimization calculations for the in-core PBR fuel cycle. The method can also be used as a practical "scoping" tool. It could, for example, be applied to determine the potential of the PBR for resisting nuclear-weapons proliferation and to optimize proliferation-resistant features. However, the purpose of this paper is to show that the method itself is viable. Refinements to the code are under way, with the objective of producing a powerful reactor physics analysis tool for PBRs.
Radulescu, Georgeta [ORNL; Gauld, Ian C [ORNL; Ilas, Germina [ORNL; Wagner, John C [ORNL
2014-01-01
This paper describes a depletion code validation approach for criticality safety analysis using burnup credit for actinide and fission product nuclides in spent nuclear fuel (SNF) compositions. The technical basis for determining the uncertainties in the calculated nuclide concentrations is comparison of calculations to available measurements obtained from destructive radiochemical assay of SNF samples. Probability distributions developed for the uncertainties in the calculated nuclide concentrations were applied to the SNF compositions of a criticality safety analysis model by the use of a Monte Carlo uncertainty sampling method to determine bias and bias uncertainty in effective neutron multiplication factor. Application of the Monte Carlo uncertainty sampling approach is demonstrated for representative criticality safety analysis models of pressurized water reactor spent fuel pool storage racks and transportation packages using burnup-dependent nuclide concentrations calculated with SCALE 6.1 and the ENDF/B-VII nuclear data. The validation approach and results support a recent revision of the U.S. Nuclear Regulatory Commission Interim Staff Guidance 8.
Engel, Eberhard
Relevance of core-valence interaction for electronic structure calculations with exact exchange E-60438 Frankfurt/Main, Germany Received 6 August 2009; published 23 October 2009 The role of the core-valence interaction in electronic structure calculations with the exact exchange of density-functional theory
The role of ORIGEN-S in the design of burnup credit spent fuel casks
Brady, M.C.
1991-12-31
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.
The role of ORIGEN-S in the design of burnup credit spent fuel casks
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.
Cerne, S.P.; Hermann, O.W.; Westfall, R.M.
1987-10-01
This study presents the reactivity loss of spent PWR fuel due to burnup in terms of the infinite lattice multiplications factor, k/sub infinity/. Calculations were performed using the SAS2 and CSAS1 control modules of the SCALE system. The k/sub infinity/ values calculated for all combinations of six enrichments, seven burnups, and five cooling times. The results are presented as a primary function of enrichment in both tabular and graphic form. An equation has been developed to estimate the tabulated values of k/sub infinity/'s by specifying enrichment, cooling time, and burnup. Atom densities for fresh fuel, and spent fuel at cooling times of 2, 10, and 20 years are included. 13 refs., 8 figs., 8 tabs.
An empirical formulation to describe the evolution of the high burnup structure
NASA Astrophysics Data System (ADS)
Lemes, Martín; Soba, Alejandro; Denis, Alicia
2015-01-01
In the present work the behavior of fuel pellets for LWR power reactors in the high burnup range (average burnup higher than about 45 MWd/kgU) is analyzed. For extended irradiation periods, a considerable Pu concentration is reached in the pellet periphery (rim zone), that contributes to local burnup. Gradually, a new microstructure develops in that ring, characterized by small grains and large pores as compared with those of the original material. In this region Xe is absent from the solid lattice (although it continues to be dissolved in the rest of the pellet). The porous microstructure in the pellet edge causes local changes in the mechanical and thermal properties, thus affecting the overall fuel behavior. It is generally accepted that the evolution of porosity in the high burnup structure (HBS) is determinant of the retention capacity of the fission gases rejected from the fuel matrix. This is the reason why, during the latest years a considerable effort has been devoted to characterizing the parameters that influence porosity. Although the mechanisms governing the microstructural transformation have not been completely elucidated yet, some empirical expressions can be given, and this is the intention of the present work, for representing the main physical parameters. Starting from several works published in the open literature, some mathematical expressions were developed to describe the behavior and progress of porosity at local burnup values ranging from 60 to 300 MWd/kgU. The analysis includes the interactions of different orders between pores, the growth of the pore radius by capturing vacancies, the evolution of porosity, pore number density and overpressure within the closed pores, the inventory of fission gas dissolved in the matrix and retained in the pores. The model is mathematically expressed by a system of non-linear differential equations. In the present work, results of this calculation scheme are compared with experimental data available in the open literature and with simulations performed by other authors. The results of these separate tests are quite satisfactory so, the next step will be the incorporation of this model as a new subroutine of the DIONISIO code, to expand the application range of this general fuel behavior simulation tool.
NASA Astrophysics Data System (ADS)
Muratov, V. G.; Lopatkin, A. V.
An important aspect in the verification of the engineering techniques used in the safety analysis of MOX-fuelled reactors, is the preparation of test calculations to determine nuclide composition variations under irradiation and analysis of burnup problem errors resulting from various factors, such as, for instance, the effect of nuclear data uncertainties on nuclide concentration calculations. So far, no universally recognized tests have been devised. A calculation technique has been developed for solving the problem using the up-to-date calculation tools and the latest versions of nuclear libraries. Initially, in 1997, a code was drawn up in an effort under ISTC Project No. 116 to calculate the burnup in one VVER-1000 fuel rod, using the MCNP Code. Later on, the authors developed a computation technique which allows calculating fuel burnup in models of a fuel rod, or a fuel assembly, or the whole reactor. It became possible to apply it to fuel burnup in all types of nuclear reactors and subcritical blankets.
Fission gas release during power bumping at high burnup
NASA Astrophysics Data System (ADS)
Cunningham, M. E.; Freshley, M. D.; Lanning, D. D.
1993-03-01
Research to define the behavior of Zircaloy-clad light-water reactor fuel irradiated to high burnup levels was conducted by the High Burnup Effects Program (HBEP). One activity conducted by the HBEP was to "bump" the power level of irradiated, commercial light-water reactor fuel rods to design limit linear heat generation rates at end-of-life. These bumping irradiations simulated end-of-life design limit linear heat generation rates and provided data on the effects of short-term, high power irradiations at high burnup applicable to the design and operating constraints imposed by maximum allowable fuel rod internal gas pressure limits. Based on net fission gas release during the bumping irradiations, it was observed that higher burnup rods had greater rod-average fractional fission gas release than lower burnup rods at equal bumping powers. It was also observed that a hold period of 48 hours at the peak power was insufficient to achieve equilibrium fission gas release. Finally, differences in the prebump location of fission gas, i.e., within the UO 2 matrix or at grain boundaries, affected the fission gas release during the bumping irradiations.
Murphy, B.D.; Kravchenko, J.; Lazarenko, A.; Pavlovitchev, A.; Sidorenko, V.; Chetverikov, A.
2000-03-01
The HELIOS reactor-physics computer program system was used to simulate the burnup of UO{sub 2} fuel in three VVER reactors. The manner in which HELIOS was used in these simulations is described. Predictions of concentrations for actinides up to {sup 244}Cm and for isotopes of neodymium were compared with laboratory-measured values. Reasonable agreement between calculated and measured values was seen for experimental samples from a fuel rod in the interior of an assembly.
B. R. Bergel’son; A. S. Gerasimov; T. S. Zaritskaya; G. V. Tikhomirov
2007-01-01
The residual energy release and radiotoxicity of spent high burnup VV?R-1000 fuel during long-term storage is investigated\\u000a as a function of time. The contributions of ?, ?, ?radiation and radiotoxicity-the maximum admissible activity of nuclides\\u000a in air and water-are taken into account in the calculations of the energy release. The data presented can be used to develop\\u000a methods for handling
Comments on fission-gas release from fuel at high burnup in Vol. 19, No. 6. [Water cooled reactors
H. Ocken; J. T. A. Roberts
1979-01-01
Meyer, Beyer, and Voglewede have proposed that an enhancement factor be applied to existing vendor models when fission-gas release (FGR) at burnups greater than 20,000 MWd\\/metric ton is calculated for licensing purposes. This enhancement factor is derived from FGR data obtained from liquid-metal-cooled fast breeder reactor (LMFBR) fuel. The analysis assumes that the intrinsic source of the high FGR measured
Benchmark of Atucha-2 PHWR RELAP5-3D control rod model by Monte Carlo MCNP5 core calculation
Pecchia, M.; D'Auria, F.; Mazzantini, O.
2012-07-01
Atucha-2 is a Siemens-designed PHWR reactor under construction in the Republic of Argentina. Its geometrical complexity and peculiarities require the adoption of advanced Monte Carlo codes for performing realistic neutronic simulations. Therefore core models of Atucha-2 PHWR were developed using MCNP5. In this work a methodology was set up to collect the flux in the hexagonal mesh by which the Atucha-2 core is represented. The scope of this activity is to evaluate the effect of obliquely inserted control rod on neutron flux in order to validate the RELAP5-3D{sup C}/NESTLE three dimensional neutron kinetic coupled thermal-hydraulic model, applied by GRNSPG/UNIPI for performing selected transients of Chapter 15 FSAR of Atucha-2. (authors)
Zhang, D.; Rahnema, F. [Georgia Institute of Technology, 770 State Street NW, Atlanta, GA 30332-0745 (United States)
2013-07-01
The coarse mesh transport method (COMET) is a highly accurate and efficient computational tool which predicts whole-core neutronics behaviors for heterogeneous reactor cores via a pre-computed eigenvalue-dependent response coefficient (function) library. Recently, a high order perturbation method was developed to significantly improve the efficiency of the library generation method. In that work, the method's accuracy and efficiency was tested in a small PWR benchmark problem. This paper extends the application of the perturbation method to include problems typical of the other water reactor cores such as BWR and CANDU bundles. It is found that the response coefficients predicted by the perturbation method for typical BWR bundles agree very well with those directly computed by the Monte Carlo method. The average and maximum relative errors in the surface-to-surface response coefficients are 0.02%-0.05% and 0.06%-0.25%, respectively. For CANDU bundles, the corresponding quantities are 0.01%-0.05% and 0.04% -0.15%. It is concluded that the perturbation method is highly accurate and efficient with a wide range of applicability. (authors)
Performance of fast reactor irradiated fueled emitters at goal burnup
NASA Astrophysics Data System (ADS)
Lawrence, Leo A.; Paxton, Dean M.; Begg, Les L.
1993-01-01
UO2-fueled W emitters were examined that had been irradiated to goal burnups of approximately 4 at.% at emitter surface temperatures to 1820 K in a fast reactor to establish their performance for use in thermionic reactors with power levels from tens of kilowatts to multimegawatts. The examinations provided first-time data on structural integrity, dimensional stability, component compatibility, and fuel and fission product behavior. The data are consistent with similar measurements at approximately 2 at.% burnup with the exception of one emitter which breached the W during irradiation.
Shamasundar, B.I.; Fehrenbach, M.E.
1981-05-01
The KUGEL computer code is designed to perform thermal/hydraulic analysis and coated-fuel particle performance calculations for axisymmetric pebble bed reactor (PBR) cores. This computer code was developed as part of a Department of Energy (DOE)-funded study designed to verify the published core performance data on PBRs. The KUGEL code is designed to interface directly with the 2DB code, a two-dimensional neutron diffusion code, to obtain distributions of thermal power, fission rate, fuel burnup, and fast neutron fluence, which are needed for thermal/hydraulic and fuel performance calculations. The code is variably dimensioned so that problem size can be easily varied. An interpolation routine allows variable mesh size to be used between the 2DB output and the two-dimensional thermal/hydraulic calculations.
LWR fuel reactivity depletion verification using 2D full core MOC and flux map data
Gunow, Geoffrey Alexander
2015-01-01
Experimental quantification of PWR fuel reactivity burnup decrement biases and uncertainties using in-core flux map data from operating power reactors has previously been conducted employing analytical methods to systematically ...
Armstrong, J.; Hamilton, H.; Hyland, B.
2013-07-01
A program is underway at Atomic Energy of Canada Limited (AECL) to develop a new fuel bundle concept to enable greater burnups for PT-HWR (pressure tube heavy water reactor) cores. One option that AECL is investigating is an internally cooled annular fuel (ICAF) element concept. ICAF contains annular cylindrical pellets with cladding on the inner and outer diameters. Coolant flows along the outside of the element and through the centre. With such a concept, the maximum fuel temperature as a function of linear element rating is significantly reduced compared to conventional, solid-rod type fuel. The preliminary ICAF bundle concept considered in this study contains 24 half-metre long internally cooled annular fuel elements and one non-fuelled centre pin. The introduction of the non-fuelled centre pin reduces the coolant void reactivity (CVR), which is the increase in reactivity that occurs on voiding the coolant in accident scenarios. Lattice cell and full core physics calculations of the preliminary ICAF fuel bundle concept have been performed for medium burnups of approximately 18 GWd/tU using WIMS-AECL and reactor fuel simulation program (RFSP). The results will be used to assist in concept configuration optimization. The effects of radial and axial core power distributions, linear element power ratings, refuelling rates and operational power ramps have been analyzed. The results suggest that burnups of greater than 18 GWd/tU can be achieved in current reactor designs. At approximately 18 GWd/tU, expected maximum linear element ratings in a PT-HWR with online-refuelling are approximately 90 kW/m. These conditions would be prohibitive for solid-rod fuel, but may be possible in ICAF fuel given the reduced maximum fuel temperature as a function of linear element rating. (authors)
Methodologies to assess potential lifetime limits for extended burnup nuclear fuel
De Vore, Curtis Vincent
1986-01-01
economic incentives associated with extended burnup fuel management schemes. The purpose of this study is to use developed methodologies to investigate the possible existence of nuclear fuel limitations with respect to extended burnup utilization... for high burnup and low burnup data . . . . . . . . . . . . 77 6 Fuel rod parameters used for modelling. . . . 164 7 Forcing functions for the Excess Load transient. . . . 167 8 Forcing functions for the Seized Pump Shaft without shutdown transient 167...
Mitenkova, E. F.; Novikov, N. V.; Blokhin, A. I.
2012-07-01
The different (U-Pu) fuel compositions are considered for next generation of sodium fast breeder reactors. The considerable discrepancies in axial and radial neutron spectra for hybrid reactor systems compared to the cores with UO{sub 2} fuel cause increasing uncertainty of generating the group nuclear constants in those reactor systems. The calculation results of BFS-62-3A critical assembly which is considered as full-scale model of BN-600 hybrid core with steel reflector specify quite different spectra in local areas. For those systems the MCNP 5 calculations demonstrate significant sensitivity of effective multiplication factor K{sub eff} and spectral indices to nuclear data libraries. For {sup 235}U, {sup 238}U, {sup 239}Pu the results of calculated radial fission rate distributions against the reconstructed ones are analyzed. Comparative analysis of spectral indices, neutron spectra and radial fission rate distributions are performed using the different versions of ENDF/B, JENDL-3.3, JENDL-4, JEFF-3.1.1 libraries and BROND-3 for Fe, Cr isotopes. For analyzing the fission rate sensitivity to the plutonium presence in the fuel {sup 239}Pu is substituted for {sup 235}U (enrichment 90%) in the FA areas containing the plutonium. For {sup 235}U, {sup 238}U, {sup 239}Pu radial fission rate distributions the explanation of pick values discrepancies is based on the group fission constants analyses and possible underestimation of some features at the experimental data recovery method (Westcott factors, temperature dependence). (authors)
NASA Astrophysics Data System (ADS)
Richards, Simon D.; Davies, Nigel; Armishaw, Malcolm J.; Dobson, Geoff P.; Wright, George A.
2014-06-01
Monte Carlo methods are increasingly being used for whole core reactor physics modelling. We describe a number of recent developments to the MONK nuclear criticality and reactor physics code to implement parallel processing, mesh-dependent burn-up and coupling to both thermal hydraulics and gamma transport codes. Results are presented which demonstrate the e_ects of gamma heating in a MONK calculation coupled to the MCBEND Monte Carlo shielding code. Experimental validation of the mesh-dependent tracking and gamma coupling methods is provided by comparison with the results of the NESSUS experiment. The gamma heating calculated by coupled MONK-MCBEND, and the neutron heating calculated by MONK, both compare well against measurement. Finally results are presented from a parallel MONK calculation of a highly detailed PWR benchmark model, which show encouraging speed-up factors on a small development cluster.
Radulescu, Georgeta [ORNL] [ORNL; Gauld, Ian C [ORNL] [ORNL; Ilas, Germina [ORNL] [ORNL; Wagner, John C [ORNL] [ORNL
2011-01-01
The expanded use of burnup credit in the United States (U.S.) for storage and transport casks, particularly in the acceptance of credit for fission products, has been constrained by the availability of experimental fission product data to support code validation. The U.S. Nuclear Regulatory Commission (NRC) staff has noted that the rationale for restricting the Interim Staff Guidance on burnup credit for storage and transportation casks (ISG-8) to actinide-only is based largely on the lack of clear, definitive experiments that can be used to estimate the bias and uncertainty for computational analyses associated with using burnup credit. To address the issues of burnup credit criticality validation, the NRC initiated a project with the Oak Ridge National Laboratory to (1) develop and establish a technically sound validation approach for commercial spent nuclear fuel (SNF) criticality safety evaluations based on best-available data and methods and (2) apply the approach for representative SNF storage and transport configurations/conditions to demonstrate its usage and applicability, as well as to provide reference bias results. The purpose of this paper is to describe the isotopic composition (depletion) validation approach and resulting observations and recommendations. Validation of the criticality calculations is addressed in a companion paper at this conference. For isotopic composition validation, the approach is to determine burnup-dependent bias and uncertainty in the effective neutron multiplication factor (keff) due to bias and uncertainty in isotopic predictions, via comparisons of isotopic composition predictions (calculated) and measured isotopic compositions from destructive radiochemical assay utilizing as much assay data as is available, and a best-estimate Monte Carlo based method. This paper (1) provides a detailed description of the burnup credit isotopic validation approach and its technical bases, (2) describes the application of the approach for representative pressurized water reactor and boiling water reactor safety analysis models to demonstrate its usage and applicability, (3) provides reference bias and uncertainty results based on a quality-assurance-controlled prerelease version of the Scale 6.1 code package and the ENDF/B-VII nuclear cross section data.
Thermal Behavior of Advanced UO Fuel at High Burnup
E. Muller; T. Lambert; K. Silberstein; N. LHullier; C. Delafoy; B. Therache
2007-01-01
To improve the fuel performance, advanced UO products are developed to reduce significantly Pellet-Cladding Interaction and Fission Gas Release to increase high burnup safety margins on Light Water Reactors. To achieve the expected improvements, doping elements are currently used, to produce large grain viscoplastic UO fuel microstructures. In that scope, AREVA NP is conducting the qualification of a new UO
Copyright Notice Triton Burnup Study in JT-60U
-T produced 3.5 MeV alphas because 1 MeV tritons and 3.5 MeV alphas have similar kinematic properties burnup measurements have been performed using a new type 14 MeV neutron detector based on scintillating fibers[1], as part of a US-Japan tokamak collaboration. Loss of alpha particles due to toroidal ripple
Spent fuel dissolution rates as a function of burnup and water chemistry
Gray, W.J.
1998-06-01
To help provide a source term for performance-assessment calculations, dissolution studies on light-water-reactor (LWR) spent fuel have been conducted over the past few years at Pacific Northwest National Laboratory in support of the Yucca Mountain Site Characterization Project. This report describes that work for fiscal years 1996 through mid-1998 and includes summaries of some results from previous years for completeness. The following conclusions were based on the results of various flowthrough dissolution rate tests and on tests designed to measure the inventories of {sup 129}I located within the fuel/cladding gap region of different spent fuels: (1) Spent fuels with burnups in the range 30 to 50 MWd/kgM all dissolved at about the same rate over the conditions tested. To help determine whether the lack of burnup dependence extends to higher and lower values, tests are in progress or planned for spent fuels with burnups of 13 and {approximately} 65 MWd/kgM. (2) Oxidation of spent fuel up to the U{sub 4}O{sub 9+x} stage does not have a large effect on intrinsic dissolution rates. However, this degree of oxidation could increase the dissolution rates of relatively intact fuel by opening the grain boundaries, thereby increasing the effective surface area that is available for contact by water. From a disposal viewpoint, this is a potentially more important consideration than the effect on intrinsic rates. (3) The gap inventories of {sup 129}I were found to be smaller than the fission gas release (FGR) for the same fuel rod with the exception of the rod with the highest FGR. Several additional fuels would have to be tested to determine whether a generalized relationship exists between FGR and {sup 129}I gap inventory for US LWR fuels.
Need for higher fuel burnup at the Hatch Plant
Beckhman, J.T. [Georgia Power Co., Birmingham, AL (United States)
1996-03-01
Hatch is a BWR 4 and has been in operation for some time. The first unit became commercial about 1975. Obtaining higher burnups, or higher average discharge exposures, is nothing new at Hatch. Since we have started, the discharge exposure of the plant has increased. Now, of course, we are not approaching the numbers currently being discussed but, the average discharge exposure has increased from around 20,000 MWD/MTU in the early to mid-1980s to 34,000 MWD/MTU in 1994, I am talking about batch average values. There are also peak bundle and peak rod values. You will have to make the conversions if you think in one way or the other because I am talking in batch averages. During Hatch`s operating history we have had some problems with fuel failure. Higher burnup fuel raises a concern about how much fuel failure you are going to have. Fuel failure is, of course, an economic issue with us. Back in the early 1980s, we had a problem with crud-induced localized corrosion, known as CILC. We have gotten over that, but we had some times when it was up around 27 fuel failures a year. That is not a pleasant time to live through because it is not what you want from an economic viewpoint or any other. We have gotten that down. We have had some fuel failures recently, but they have not been related to fuel burnup or to corrosion. In fact, the number of failures has decreased from the early 1980s to the 90s even though burnup increased during that time. The fuel failures are more debris-related-type failures. In addition to increasing burnups, utilities are actively evaluating or have already incorporated power uprate and longer fuel cycles (e.g., 2-year cycles). The goal is to balance out the higher power density, longer cycles, higher burnup, and to have no leakers. Why do we as an industry want to have higher burnup fuel? That is what I want to tell you a little bit about.
Pancake core high conversion light water reactor concept
Y. Ishiguro; K. Okumura
1989-01-01
A new concept is proposed for a high conversion light water reactor (HCLWR) that achieves both high conversion and high burnup while maintaining a negative void reactivity coefficient. This HCLWR has a flat pancake core with thick axial blankets. By using the flat core, a potential problem of HCLWRs, the positive void reactivity coefficient can be reduced by neutron leakage,
Design and analysis of a nuclear reactor core for innovative small light water reactors
NASA Astrophysics Data System (ADS)
Soldatov, Alexey I.
In order to address the energy needs of developing countries and remote communities, Oregon State University has proposed the Multi-Application Small Light Water Reactor (MASLWR) design. In order to achieve five years of operation without refueling, use of 8% enriched fuel is necessary. This dissertation is focused on core design issues related with increased fuel enrichment (8.0%) and specific MASLWR operational conditions (such as lower operational pressure and temperature, and increased leakage due to small core). Neutron physics calculations are performed with the commercial nuclear industry tools CASMO-4 and SIMULATE-3, developed by Studsvik Scandpower Inc. The first set of results are generated from infinite lattice level calculations with CASMO-4, and focus on evaluation of the principal differences between standard PWR fuel and MASLWR fuel. Chapter 4-1 covers aspects of fuel isotopic composition changes with burnup, evaluation of kinetic parameters and reactivity coefficients. Chapter 4-2 discusses gadolinium self-shielding and shadowing effects, and subsequent impacts on power generation peaking and Reactor Control System shadowing. The second aspect of the research is dedicated to core design issues, such as reflector design (chapter 4-3), burnable absorber distribution and programmed fuel burnup and fuel use strategy (chapter 4-4). This section also includes discussion of the parameters important for safety and evaluation of Reactor Control System options for the proposed core design. An evaluation of the sensitivity of the proposed design to uncertainty in calculated parameters is presented in chapter 4-5. The results presented in this dissertation cover a new area of reactor design and operational parameters, and may be applicable to other small and large pressurized water reactor designs.
Evaluation of fission gas release in high-burnup light water reactor fuel rods
Barner, J.O.; Cunningham, M.E.; Freshley, M.D.; Lanning, D.D. (Battelle Pacific Northwest Lab., Richland, WA (United States))
1993-05-01
Research to define the behavior of Zircaloy-clad light water reactor (LWR) UO[sub 2] fuel irradiated to high burnup levels was conducted as part of the High Burnup Effects Program (HBEP). The HBEP was a 12-yr program that ultimately acquired, characterized, irradiated, and examined after irradiation 82 LWR fuel rods ranging in rod-average fuel burnup from 22 to 69 MWd/kgM with a peak pellet burnup of 83 MWd/kg M. A principal emphasis of the HBEP was to evaluate the effect of high burnup on fission gas release. It was confirmed that fission gas release remained as dependent on design and irradiation history parameters at high burnup levels as at low to moderate burnup levels. One observed high-burnup effect was the development of a burnup-dependent microstructure at the fuel pellet surface when pellet-edge burnup exceeded 65 MWd/kgM. This low-temperature rim region' was characterized by a loss of optically definable grain structure, a high volume of porosity, and diffusion of fission gas from the UO[sub 2] matrix to the porosity. Although the rim region has the potential for enhanced fission gas release, it is concluded that no significant enhancement of rod-average fission gas release at high burnup levels was observed for the examined fuel rods.
Thermal Analysis for a Heterogeneous VHTR Transmutation Fuel Block as a Function of Burnup
Bays, Samuel [University of Florida, P.O Box 1625, Idaho Falls, ID 83415-3860 (United States); Sabharwall, Piyush [Idaho State University, P.O Box 1625, Idaho Falls, ID 83415-3860 (United States); Herring, Stephen; Weaver, Kevan [Idaho National Laboratory, P.O Box 1625, Idaho Falls, ID 83415-3860 (United States)
2006-07-01
The VHTR is one of the Generation IV Reactor concepts utilizing TRISO fuel, which enables it to have a higher burnup. Thus, giving higher fuel utilization which is capitalized upon in this work, where one third of the uranium carrying fuel compacts are replaced with transuranic burning fuel for light water reactor waste transmutation. The single fuel block burnup calculation of the transuranic fuel yielded disproportionate power sharing between transuranic and uranium fuel. Therefore, a finite difference code was prepared to assess the maximum temperature and temperature gradient that can be expected in a heterogeneous fuel block intended for waste transmutation. Power reduction by a factor of 1.4 was observed by coolant channels neighboring both Pu/MA TRISO and UO{sub 2} compacts from beginning of fuel life to the end of life. This power reduction allowed for a corresponding reduction in the end of life UO{sub 2} centerline temperature even though the UO{sub 2} power density was four times that of Pu/MA TRISO. (authors)
Study on Ultra-Long Life,Small U-Zr Metallic Fuelled Core With Burnable Poison
Kenji Tsuji; Hiromitsu Inagaki; Akira Nishikawa [Chubu Electric Power Company Inc., 1, Higashi-shincho Higashi-ku, Nagoya-shi, ACH 461-8680 (Japan); Hisato Matsumiya; Yoshiaki Sakashita; Yasuyuki Moriki; Mitsuaki Yamaoka; Norihiko Handa [Toshiba Corporation (Japan)
2002-07-01
A conceptual design for a 50 MWe sodium cooled, U-Pu-Zr metallic fuelled, fast reactor core, which aims at a core lifetime of 30 years, has been performed [1]. As for the compensation for a large burn-up reactivity through 30 years, an axially movable reflector, which is located around the core, carries the major part of it and a burnable poison does the rest. This concept has achieved not only a long core lifetime but also a high discharged burn-up. On this study, a conceptual design for a small fast reactor loading U-Zr metallic fuelled core instead of U-Pu-Zr fuelled core has been conducted, based on the original core arrangement of 4S reactor [2]. Within the range of this study including safety requirements, adopting the burnable poison would be effective to construct a core concept that achieves both a long lifetime and a high discharged burn-up. (authors)
Safronova, U.I.; Johnson, W.R.; Shlyaptseva, A.; Hamasha, S.
2003-05-01
Energies of (3s{sup 2}3p{sup 6}3d{sup 9}4l4l{sup '}), (3s{sup 2}3p{sup 5}3d{sup 10}4l4l{sup '}), and (3s3p{sup 6}3d{sup 10}4l4l{sup '}) states for Cu-like ions with Z=30-100 are evaluated to second order in relativistic many-body perturbation theory (RMBPT) starting from a Ni-like Dirac-Fock potential. Second-order Coulomb and Breit-Coulomb interactions are included. Correction for the frequency dependence of the Breit interaction is taken into account in lowest order. The Lamb shift correction to energies is also included in lowest order. Intrinsic particle-particle-hole contributions to energies are found to be 20-30 % of the sum of one- and two-body contributions. Transition rates and line strengths are calculated for the 3l-4l{sup '} electric-dipole (E1) transitions in Cu-like ions with nuclear charge Z=30-100. RMBPT including the Breit interaction is used to evaluate retarded E1 matrix elements in length and velocity forms. First-order RMBPT is used to obtain intermediate coupling coefficients, and second-order RMBPT is used to calculate transition matrix elements. A detailed discussion of the various contributions to the dipole- matrix elements and energy levels is given for copperlike tungsten (Z=74). The transition energies used in the calculation of oscillator strengths and transition rates are from second-order RMBPT. Trends of the transition rates as functions of Z are illustrated graphically for selected transitions. Comparisons are made with available experimental data. These atomic data are important in the modeling of M-shell radiation spectra of heavy ions generated in electron-beam ion trap experiments and in M-shell diagnostics of plasmas.
WIMS/PANTHER analysis of UO{sub 2}/MOX cores using embedded super-cells
Knight, M.; Bryce, P.; Hall, S.
2012-07-01
This paper describes a method of analysing PWR UO{sub 2}MOX cores with WIMS/PANTHER. Embedded super-cells, run within the reactor code, are used to correct the standard methodology of using 2-group smeared data from single assembly lattice calculations. In many other codes the weakness of this standard approach has been improved for MOX by imposing a more realistic environment in the lattice code, or by improving the sophistication of the reactor code. In this approach an intermediate set of calculations is introduced, leaving both lattice and reactor calculations broadly unchanged. The essence of the approach is that the whole core is broken down into a set of 'embedded' super-cells, each extending over just four quarter assemblies, with zero leakage imposed at the assembly mid-lines. Each supercell is solved twice, first with a detailed multi-group pin-by-pin solution, and then with the standard single assembly approach. Correction factors are defined by comparing the two solutions, and these can be applied in whole core calculations. The restriction that all such calculations are modelled with zero leakage means that they are independent of each other and of the core-wide flux shape. This allows parallel pre-calculation for the entire cycle once the loading pattern has been determined, in much the same way that single assembly lattice calculations can be pre-calculated once the range of fuel types is known. Comparisons against a whole core pin-by-pin reference demonstrates that the embedding process does not introduce a significant error, even after burnup and refuelling. Comparisons against a WIMS reference demonstrate that a pin-by-pin multi-group diffusion solution is capable of capturing the main interface effects. This therefore defines a practical approach for achieving results close to lattice code accuracy, but broadly at the cost of a standard reactor calculation. (authors)
B. Boer; A. M. Ougouag
2010-09-01
The Deep-Burn (DB) concept focuses on the destruction of transuranic nuclides from used light water reactor fuel. These transuranic nuclides are incorporated into TRISO coated fuel particles and used in gas-cooled reactors with the aim of a fractional fuel burnup of 60 to 70% in fissions per initial metal atom (FIMA). This high performance is expected through the use of multiple recirculation passes of the fuel in pebble form without any physical or chemical changes between passes. In particular, the concept does not call for reprocessing of the fuel between passes. In principle, the DB pebble bed concept employs the same reactor designs as the presently envisioned low-enriched uranium core designs, such as the 400 MWth Pebble Bed Modular Reactor (PBMR-400). Although it has been shown in the previous Fiscal Year (2009) that a PuO2 fueled pebble bed reactor concept is viable, achieving a high fuel burnup, while remaining within safety-imposed prescribed operational limits for fuel temperature, power peaking and temperature reactivity feedback coefficients for the entire temperature range, is challenging. The presence of the isotopes 239-Pu, 240-Pu and 241-Pu that have resonances in the thermal energy range significantly modifies the neutron thermal energy spectrum as compared to a ”standard,” UO2-fueled core. Therefore, the DB pebble bed core exhibits a relatively hard neutron energy spectrum. However, regions within the pebble bed that are near the graphite reflectors experience a locally softer spectrum. This can lead to power and temperature peaking in these regions. Furthermore, a shift of the thermal energy spectrum with increasing temperature can lead to increased absorption in the resonances of the fissile Pu isotopes. This can lead to a positive temperature reactivity coefficient for the graphite moderator under certain operating conditions. The effort of this task in FY 2010 has focused on the optimization of the core to maximize the pebble discharge burnup level, while retaining its inherent safety characteristics. Using generic pebble bed reactor cores, this task will perform physics calculations to evaluate the capabilities of the pebble bed reactor to perform utilization and destruction of LWR used-fuel transuranics. The task will use established benchmarked models, and will introduce modeling advancements appropriate to the nature of the fuel considered (high TRU content and high burn-up).
Reactivity loss validation of high burn-up PWR fuels with pile-oscillation experiments in MINERVE
Leconte, P.; Vaglio-Gaudard, C.; Eschbach, R.; Di-Salvo, J.; Antony, M.; Pepino, A. [CEA, DEN, DER, Cadarache, F-13108 Saint-Paul-Lez-Durance (France)
2012-07-01
The ALIX experimental program relies on the experimental validation of the spent fuel inventory, by chemical analysis of samples irradiated in a PWR between 5 and 7 cycles, and also on the experimental validation of the spent fuel reactivity loss with bum-up, obtained by pile-oscillation measurements in the MINERVE reactor. These latter experiments provide an overall validation of both the fuel inventory and of the nuclear data responsible for the reactivity loss. This program offers also unique experimental data for fuels with a burn-up reaching 85 GWd/t, as spent fuels in French PWRs never exceeds 70 GWd/t up to now. The analysis of these experiments is done in two steps with the APOLLO2/SHEM-MOC/CEA2005v4 package. In the first one, the fuel inventory of each sample is obtained by assembly calculations. The calculation route consists in the self-shielding of cross sections on the 281 energy group SHEM mesh, followed by the flux calculation by the Method Of Characteristics in a 2D-exact heterogeneous geometry of the assembly, and finally a depletion calculation by an iterative resolution of the Bateman equations. In the second step, the fuel inventory is used in the analysis of pile-oscillation experiments in which the reactivity of the ALIX spent fuel samples is compared to the reactivity of fresh fuel samples. The comparison between Experiment and Calculation shows satisfactory results with the JEFF3.1.1 library which predicts the reactivity loss within 2% for burn-up of {approx}75 GWd/t and within 4% for burn-up of {approx}85 GWd/t. (authors)
Mathematical modeling of the heat treatment and combustion of a coal particle. V. Burn-up stage
NASA Astrophysics Data System (ADS)
Enkhjargal, Kh.; Salomatov, V. V.
2011-07-01
The present material is a sequel of the previous publications of the authors in this journal under a common title in which by means of mathematical modeling the sequential stages of the process of combustion of coal fuels have been obtained: heating, drying, escape of volatiles, and ignition. Mathematical models of the final stage of combustion of an individual particle — the burn-up stage — have been formulated. On the basis of the solution methods for nonlinear boundary-value problems developed by us, approximate-analytic formulas for two characteristic regimes, burn-up simultaneously with the evaporation of the remaining moisture and burn-up of the completely dried coke residue, have been obtained. The previous history of the physical and chemical phenomena in the general burning pattern is taken into account. The influence of the ash shell on the duration of combustion has been extimated. Comparison of calculations by the obtained dependences with the results of other authors has been made. It showed an accuracy sufficient for engineering applications.
Feasibility Study of MOX Fuel Online Burnup Analysis
Dennis, M.L.; Usman, S. [University of Missouri-Rolla, 222 Fulton Hall, 1870 Miner Circle, Rolla, MO 65409-0170 (United States)
2006-07-01
This research is an extension of well established Non-Destructive Analysis of UO fuel using gamma spectroscopy of Cs-137 and other related isotopes. Given the performance similarities between UO fuel and MOX fuel, investigations are underway to develop similar correlation for MOX. MOX fuel burnup and decay simulations are being performed using ORIGEN-ARP (Oak Ridge Isotope Generation and Depletion Code - Automatic Rapid Processing). Simulation results are being analyzed and will be used to determine performance specifications of a detection system for field applications. Analysis of isotopic activity from irradiated fuel will be used to develop correlations to determine burn-up and Plutonium content of MOX fuel. These results will be particularly useful in view of the recent interest in MOX fuel. (authors)
Diaz, F.; Vilkas, M. J.; Ishikawa, Y. [Department of Chemistry and the Chemical Physics Program, University of Puerto Rico, P.O. Box 23346, San Juan, PR 00931-3346 (Puerto Rico); Beiersdorfer, P., E-mail: beiersdorfer1@llnl.gov [Physics Division, Lawrence Livermore National Laboratory, Livermore, CA 94550 (United States)
2013-07-01
Accurate theoretical energy level, lifetime, and transition probability calculations of core-excited Fe XVI were performed employing the relativistic Multireference Moller-Plesset perturbation theory. In these computations the term energies of the highly excited n {<=} 5 states arising from the configuration 1s {sup 2}2s{sup k} 2p{sup m} 3l {sup p} nl' {sup q}, where k + m + p + q = 9, l {<=} 3 and p + q {<=} 2 are considered, including those of the autoionizing levels with a hole-state in the L-shell. All even and odd parity states of sodium-like iron ion were included for a total of 1784 levels. Comparison of the calculated L-shell transition wavelengths with those from laboratory measurements shows excellent agreement. Therefore, our calculation may be used to predict the wavelengths of as of yet unobserved Fe XVI, such as the second strongest 2p-3d Fe XVI line, which has not been directly observed in the laboratory and which blends with one of the prominent Fe XVII lines.
Structure of high-burnup-fuel Zircaloy cladding. [PWR; BWR
Chung
1983-01-01
Zircaloy cladding from high-burnup (> 20 MWd\\/kg U) fuel rods in light-water reactors is characterized by a high density of irradiation-induced defects (RID), compositional changes (e.g., oxygen and hydrogen uptake) associated with in-service corrosion, and geometrical changes produced by creepdown, bowing, and irradiation-induced growth. During a reactor power transient, the cladding is subject to localized stress imposed by thermal expansion
Irradiation-induced recrystallization in high burnup UO2 fuel
NASA Astrophysics Data System (ADS)
Nogita, K.; Une, K.
1995-11-01
The formation mechanism of recrystallized grains and coarsened bubbles in the peripheral region of a high burnup UO2 pellet has been proposed on the basis of transmission electron microscopy (TEM) results. TEM observations were made on the pellet edge (local burnup: 100 GWd/t) of a fuel sample which had been irradiated to a pellet average burnup of 49 GWd/t in a BWR. A recrystallized grain region was found adjacent to an accumulation region having an extremely high density of dislocations. The dislocations were organized into subdivided grains with high angle boundaries which were regarded as the nuclei for recrystallization. Almost all the coarsened bubbles were surrounded by recrystallized grains. Small intergranular bubbles precipitated on the recrystallized grain boundaries and interconnected with each other, and then linked up with the coarsened bubbles. From these results, it was concluded that the coarsened bubbles were formed by sweeping out of small bubbles during grain growth on recrystallization. Finally an overall formation process for the unique microstructural change, the rim structure, was proposed.
Development of Erbia-bearing Super High Burnup Fuel
Akio, Yamamoto; Toshikazu, Takeda; Hironobu, Unesaki; Masaaki, Mori; Masatoshi Yamasaki
2006-07-01
In this paper, concept and development plan of the Erbia (Er{sub 2}O{sub 3})-bearing super high burnup (Er-SHB) fuel for LWRs are described. In order to reduce the number of spent fuel assemblies, utilization of high burnup fuels with higher uranium enrichment is effective. However, the upper limitation of enrichment for LWR fuels is 5 wt% and current advanced fuel assemblies for LWRs are already reaching this limit. Though various efforts to overcome the 5 wt% enrichment limit have been undergoing, it will require considerable cost that may offset the economic benefit of high burnup fuels. We are proposing another pathway. By adding low content ({>=}0.2 wt%) of Erbia in all UO{sub 2} powder, reactivity of high enrichment (>5 wt%) fuel is suppressed under that of current fuel assemblies, i.e. we leverage the negative reactivity credit of Erbia. Since Erbia is mixed into UO{sub 2} powder just after the re-conversion, we can avoid most of the criticality safety issues appearing in the front-end stream. Namely, major improvements and re-licensing for equipments in transportation, storage and fabrication process will not be necessary. Therefore, the Er-SHB fuel will significantly contribute to reduction of fuel cycle cost. (authors)
Gurevich, M. I.; Oleynik, D. S.; Russkov, A. A.; Voloschenko, A. M.
2006-07-01
The tracing algorithm that is implemented in the geometrical module of Monte-Carlo transport code MCU is applied to calculate the volume fractions of original materials by spatial cells of the mesh that overlays problem geometry. In this way the 3D combinatorial geometry presentation of the problem geometry, used by MCU code, is transformed to the user defined 2D or 3D bit-mapped ones. Next, these data are used in the volume fraction (VF) method to approximate problem geometry by introducing additional mixtures for spatial cells, where a few original materials are included. We have found that in solving realistic 2D and 3D core problems a sufficiently fast convergence of the VF method takes place if the spatial mesh is refined. Virtually, the proposed variant of implementation of the VF method seems as a suitable geometry interface between Monte-Carlo and S{sub n} transport codes. (authors)
NASA Astrophysics Data System (ADS)
Wang, Q. E.; Zhang, F. C.
2015-06-01
We study the electronic structure of vortex core states of FeSe superconductors based on a t2 g three-orbital model by solving the Bogoliubov-de Gennes (BdG) equation self-consistently. The orbital-resolved vortex core states of different pairing symmetries manifest themselves as distinguishable structures due to different quasiparticle wave functions. The obtained vortices are classified in terms of the invariant subgroups of the symmetry group of the mean-field Hamiltonian in the presence of a magnetic field. Isotropic s and anisotropic s -wave vortices have G5 symmetry for each orbital, whereas dx2-y2-wave vortices show G6* symmetry for dx z /y z orbitals and G5* symmetry for dx y orbital. In the case of dx2-y2-wave vortices, hybridized-pairing between dx z and dy z orbitals gives rise to a relative phase difference in terms of gauge transformed pairing order parameters between dx z /y z and dx y orbitals, which is essentially caused by a transformation of co-representation of G5* and G6* subgroups. The calculated local density of states (LDOS) of dx2-y2-wave vortices shows a qualitatively similar pattern with the experimental results. The phase difference of ?/4 between dx z /y z and dx y orbital-resolved dx2-y2-wave vortices can be verified by further experimental observation.
Spent LWR fuel dry storage in large transport and storage casks after extended burnup
NASA Astrophysics Data System (ADS)
Spilker, Harry; Peehs, Martin; Dyck, Hans-Peter; Kaspar, Guenter; Nissen, Klaus
1997-11-01
Dry spent LWR fuel storage is licensed for single fuel assemblies with rod burnup to 65 GWd/tHM. This allows dry spent fuel storage of reloads with a batch average up to 55 GWd/tHM. The leading defect mechanism for spent fuel rods in dry storage is hoop strain. Fuel rod degradation can be prevented by limiting creep. Post-pile creep of fuel rod cladding can be described conservatively by the creep of unirradiated cladding. In order to extend the database, internally pressurized creep samples were investigated for time intervals up to 10 000 h. Test temperatures were between 250 and 400°C, and the hoop stresses applied ranged from 80 to 150 N/mm 2. The resulting data were described mathematically by an interpolation formula. Based on the fuel assemblies end-of-life data the maximum CASTOR V cask storage temperature was calculated to be between 348°C and 358°C at the beginning.
Extended burnup fuel - a beneficial environmental step for the nuclear fuel cycle
Tulenko, J.S. (Univ. of Florida, Gainesville (United States))
1992-01-01
The beneficial environmental effects generated by extending the burnup of light water reactors (LWR) fuel covers the entire fuel cycle from cradle (mining) to grave (fuel disposal). The purpose of this paper is to outline the beneficial efforts of extending burnup and show how a proposed financial incentive waste charge would encourage the development and utilization of extended burnup. Environmental consciousness demands that economic incentives be implemented to encourage utilities to extend burnup of their LWR fuel. This paper builds on earlier work and focuses and extends their results. The cornerstone of the positive environmental effects of extended burnup results from the increased plutonium energy generation as burnup is increased and residence time is increased.
Kalantzis, Georgios; Tachibana, Hidenobu
2014-01-01
For microdosimetric calculations event-by-event Monte Carlo (MC) methods are considered the most accurate. The main shortcoming of those methods is the extensive requirement for computational time. In this work we present an event-by-event MC code of low projectile energy electron and proton tracks for accelerated microdosimetric MC simulations on a graphic processing unit (GPU). Additionally, a hybrid implementation scheme was realized by employing OpenMP and CUDA in such a way that both GPU and multi-core CPU were utilized simultaneously. The two implementation schemes have been tested and compared with the sequential single threaded MC code on the CPU. Performance comparison was established on the speed-up for a set of benchmarking cases of electron and proton tracks. A maximum speedup of 67.2 was achieved for the GPU-based MC code, while a further improvement of the speedup up to 20% was achieved for the hybrid approach. The results indicate the capability of our CPU-GPU implementation for accelerated MC microdosimetric calculations of both electron and proton tracks without loss of accuracy. PMID:24113420
Soewono, C. N.; Takaki, N. [Dept. of Applied Science Engineering, Faculty Tokai Univ., Kanagawa-ken, Hiratsuka-shi Kitakaname 4-1-1 (Japan)
2012-07-01
In this work genetic algorithm was proposed to solve fuel loading pattern optimization problem in thorium fueled heavy water reactor. The objective function of optimization was to maximize the conversion ratio and minimize power peaking factor. Those objectives were simultaneously optimized using non-dominated Pareto-based population ranking optimal method. Members of non-dominated population were assigned selection probabilities based on their rankings in a manner similar to Baker's single criterion ranking selection procedure. A selected non-dominated member was bred through simple mutation or one-point crossover process to produce a new member. The genetic algorithm program was developed in FORTRAN 90 while neutronic calculation and analysis was done by COREBN code, a module of core burn-up calculation for SRAC. (authors)
Global variance reduction for Monte Carlo reactor physics calculations
Zhang, Q.; Abdel-Khalik, H. S.
2013-07-01
Over the past few decades, hybrid Monte-Carlo-Deterministic (MC-DT) techniques have been mostly focusing on the development of techniques primarily with shielding applications in mind, i.e. problems featuring a limited number of responses. This paper focuses on the application of a new hybrid MC-DT technique: the SUBSPACE method, for reactor analysis calculation. The SUBSPACE method is designed to overcome the lack of efficiency that hampers the application of MC methods in routine analysis calculations on the assembly level where typically one needs to execute the flux solver in the order of 10{sup 3}-10{sup 5} times. It places high premium on attaining high computational efficiency for reactor analysis application by identifying and capitalizing on the existing correlations between responses of interest. This paper places particular emphasis on using the SUBSPACE method for preparing homogenized few-group cross section sets on the assembly level for subsequent use in full-core diffusion calculations. A BWR assembly model is employed to calculate homogenized few-group cross sections for different burn-up steps. It is found that using the SUBSPACE method significant speedup can be achieved over the state of the art FW-CADIS method. While the presented speed-up alone is not sufficient to render the MC method competitive with the DT method, we believe this work will become a major step on the way of leveraging the accuracy of MC calculations for assembly calculations. (authors)
Calculation of fuel pin failure timing under LOCA conditions
Jones, K.R.; Wade, N.L.; Siefken, L.J.; Straka, M.; Katsma, K.R.
1991-10-01
The objective of this research was to develop and demonstrate a methodology for calculation of the time interval between receipt of the containment isolation signals and the first fuel pin failure for loss-of-coolant accidents (LOCAs). Demonstration calculations were performed for a Babcock and Wilcox (B W) design (Oconee) and a Westinghouse (W) 4-loop design (Seabrook). Sensitivity studies were performed to assess the impacts of fuel pin burnup, axial peaking factor, break size, emergency core cooling system (ECCS) availability, and main coolant pump trip on these items. The analysis was performed using a four-code approach, comprised of FRAPCON-2, SCDAP/RELAP5/MOD3, TRAC-PF1/MOD1, and FRAP-T6. In addition to the calculation of timing results, this analysis provided a comparison of the capabilities of SCDAP/RELAP5/MOD3 with TRAC-PF1/MOD1 for large-break LOCA analysis. This paper discusses the methodology employed and the code development efforts required to implement the methodology. The shortest time intervals calculated between initiation of containment isolation and fuel pin failure were 11.4 s and 19.1 for the B W and W plants, respectively. The FRAP-T6 fuel pin failure times calculated using thermal-hydraulic data generated by SCDAP/RELAP5/MOD3 were more conservative than those calculated using data generated by TRAC-PF1/MOD1. 18 refs., 7 figs., 4 tabs.
Evaluation of fission gas release in high-burnup light water reactor fuel rods
J. O. Barner; M. E. Cunningham; M. D. Freshley; D. D. Lanning
1993-01-01
Research to define the behavior of Zircaloy-clad light water reactor (LWR) UO[sub 2] fuel irradiated to high burnup levels was conducted as part of the High Burnup Effects Program (HBEP). The HBEP was a 12-yr program that ultimately acquired, characterized, irradiated, and examined after irradiation 82 LWR fuel rods ranging in rod-average fuel burnup from 22 to 69 MWd\\/kgM with
Plutonium isotopic composition of high burnup spent fuel discharged from light water reactors
Yoshihiro Nakano; Tsutomu Okubo
2011-01-01
The isotopic composition and amount of plutonium (Pu) in spent fuel from a high burnup boiling water reactor (HB-BWR) and a high burnup pressurized water reactor (HB-PWR), each with an average discharge burnup of 70GWd\\/t, were estimated, in order to evaluate fast breeder reactor (FBR) fuel composition in the transition period from LWRs to FBRs.The HB-BWR employs spectral shift rods
Using SERPENT Monte Carlo and Burnup code to model Traveling Wave Reactors (TWR)
NASA Astrophysics Data System (ADS)
Gulik, Volodymyr; Pavlovych, Volodymyr; Tkaczyk, Alan Henry
2014-06-01
This paper is mainly devoted to the proof-of-principle implementation of the SERPENT code for the simulation of traveling wave reactors. The investigation of SERPENT 1.1.19 code for multiprocessor tasks with long burnup steps was performed. The investigation of SERPENT 2 code for multiprocessor tasks with long burnup steps was performed. Methods to remove the influence of the ignition zone were considered, and neutronics simulations with various fragmentations of the burnup zone were performed.
Navarro, Jorge; Ring, Terry A.; Nigg, David W.
2015-03-01
A deconvolution method for a LaBr? 1"x1" detector for nondestructive Advanced Test Reactor (ATR) fuel burnup applications was developed. The method consisted of obtaining the detector response function, applying a deconvolution algorithm to 1”x1” LaBr? simulated, data along with evaluating the effects that deconvolution have on nondestructively determining ATR fuel burnup. The simulated response function of the detector was obtained using MCNPX as well with experimental data. The Maximum-Likelihood Expectation Maximization (MLEM) deconvolution algorithm was selected to enhance one-isotope source-simulated and fuel- simulated spectra. The final evaluation of the study consisted of measuring the performance of the fuel burnup calibrationmore »curve for the convoluted and deconvoluted cases. The methodology was developed in order to help design a reliable, high resolution, rugged and robust detection system for the ATR fuel canal capable of collecting high performance data for model validation, along with a system that can calculate burnup and using experimental scintillator detector data.« less
NASA Astrophysics Data System (ADS)
Su'ud, Zaki; Sekimoto, H.
2014-09-01
Pb-Bi Cooled fast reactors with modified CANDLE burn-up scheme with 10 regions and 10 years cycle length has been investigated from neutronic aspects. In this study the safety aspect of such reactors have been investigated and discussed. Several condition of unprotected loss of flow (ULOF) and unprotected rod run-out transient over power (UTOP) have been simulated and the results show that the reactors excellent safety performance. At 80 seconds after unprotected loss of flow condition, the core flow rate drop to about 25% of its initial flow and slowly move toward its natural circulation level. The maximum fuel temperature can be managed below 1000°C and the maximum cladding temperature can be managed below 700°C. The dominant reactivity feedback is radial core expansion and Doppler effect, followed by coolant density effect and fuel axial expansion effect.
Yu. V. Petrov; A. I. Nazarov; M. S. Onegin; V. Yu. Petrov; E. G. Sakhnovsky
2005-09-15
Using modern methods of reactor physics we have performed full-scale calculations of the natural reactor Oklo. For reliability we have used recent version of two Monte Carlo codes: Russian code MCU REA and world wide known code MCNP (USA). Both codes produce similar results. We have constructed a computer model of the reactor Oklo zone RZ2 which takes into account all details of design and composition. The calculations were performed for three fresh cores with different uranium contents. Multiplication factors, reactivities and neutron fluxes were calculated. We have estimated also the temperature and void effects for the fresh core. As would be expected, we have found for the fresh core a significant difference between reactor and Maxwell spectra, which was used before for averaging cross sections in the Oklo reactor. The averaged cross section of Sm-149 and its dependence on the shift of resonance position (due to variation of fundamental constants) are significantly different from previous results. Contrary to results of some previous papers we find no evidence for the change of the fine structure constant in the past and obtain new, most accurate limits on its variation with time: -4 10^{-17}year^{-1} < d alpha/dt/alpha < 3 10^{-17} year^{-1} A further improvement in the accuracy of the limits can be achieved by taking account of the core burnup. These calculations are in progress.
Investigation of the Performance of D2O-Cooled High-Conversion Reactors for Fuel Cycle Calculations
Hikaru Hiruta; Gilles Youinou
2013-09-01
This report presents FY13 activities for the analysis of D2O cooled tight-pitch High-Conversion PWRs (HCPWRs) with U-Pu and Th-U fueled cores aiming at break-even or near breeder conditions while retaining the negative void reactivity. The analyses are carried out from several aspects which could not be covered in FY12 activities. SCALE 6.1 code system is utilized, and a series of simple 3D fuel pin-cell models are developed in order to perform Monte Carlo based criticality and burnup calculations. The performance of U-Pu fueled cores with axial and internal blankets is analyzed in terms of their impact on the relative fissile Pu mass balance, initial Pu enrichment, and void coefficient. In FY12, Pu conversion performances of D2O-cooled HCPWRs fueled with MOX were evaluated with small sized axial/internal DU blankets (approximately 4cm of axial length) in order to ensure the negative void reactivity, which evidently limits the conversion performance of HCPWRs. In this fiscal year report, the axial sizes of DU blankets are extended up to 30 cm in order to evaluate the amount of DU necessary to reach break-even and/or breeding conditions. Several attempts are made in order to attain the milestone of the HCPWR designs (i.e., break-even condition and negative void reactivity) by modeling of HCPWRs under different conditions such as boiling of D2O coolant, MOX with different 235U enrichment, and different target burnups. A similar set of analyses are performed for Th-U fueled cores. Several promising characteristics of 233U over other fissile like 239Pu and 235U, most notably its higher fission neutrons per absorption in thermal and epithermal ranges combined with lower ___ in the fast range than 239Pu allows Th-U cores to be taller than MOX ones. Such an advantage results in 4% higher relative fissile mass balance than that of U-Pu fueled cores while retaining the negative void reactivity until the target burnup of 51 GWd/t. Several other distinctions between U-Pu and Th-U fueled cores are identified by evaluating the sensitivity coefficients of keff, mass balance, and void coefficient. The effect of advanced iron alloy cladding (i.e., FeCrAl) on the performance of Pu conversion in MOX fueled cores is studied instead of using standard stainless-steel cladding. Variations in clad thickness and coolant-to-fuel volume ratio are also exercised. The use of FeCrAl instead of SS as a cladding alloy reduces the required Pu enrichment and improves the Pu conversion rate primarily due to the absence of nickel in the cladding alloy that results in the reduction of the neutron absorption. Also the difference in void coefficients between SS and FeCrAl alloys is nearly 500 pcm over the entire burnup range. The report also shows sensitivity and uncertainty analyses in order to characterize D2O cooled HCPWRs from different aspects. The uncertainties of integral parameters (keff and void coefficient) for selected reactor cores are evaluated at different burnup points in order to find similarities and trends respect to D2O-HCPWR.
Attainable Burnup in a LIFE Engine Loaded with Depleted Uranium
Fratoni, M; Kramer, K J; Latkowski, J F
2009-11-30
The Laser Inertial Fusion-based Energy (LIFE) system uses a laser-based fusion source for electricity production. The (D,T) reaction, beside a pure fusion system, allows the option to drive a sub-critical fission blanket in order to increase the total energy gain. In a typical fusion-fission LIFE engine the fission blanket is a spherical shell around the fusion source, preceded by a beryllium shell for neutron multiplications by means of (n,2n) reactions. The fuel is in the form of TRISO particles dispersed in carbon pebbles, cooled by flibe. The optimal design features 80 cm thick blanket, 16 cm multiplier, and 20% TRISO packing factor. A blanket loaded with depleted uranium and depleted in a single batch with continuous mixing can achieve burnup as high as {approx}85% FIMA while generating 2,000 MW of total thermal power and producing enough tritium to be used for fusion. A multi-segment blanket with a central promotion shuffling scheme enhances burnup to {approx}90% FIMA, whereas a blanket that is operated with continuous refueling achieves only 82% FIMA under the same constraints of thermal power and tritium self-sufficiency. Both, multi-segment and continuous refueling eliminate the need for a fissile breeding phase.
Microstructural Characterization of High Burn-up Mixed Oxide Fast Reactor Fuel
Melissa C. Teague; Brian P. Gorman; Steven L. Hayes; Douglas L. Porter; Jeffrey King
2013-10-01
High burn-up mixed oxide fuel with local burn-ups of 3.4–23.7% FIMA (fissions per initial metal atom) were destructively examined as part of a research project to understand the performance of oxide fuel at extreme burn-ups. Optical metallography of fuel cross-sections measured the fuel-to-cladding gap, clad thickness, and central void evolution in the samples. The fuel-to-cladding gap closed significantly in samples with burn-ups below 7–9% FIMA. Samples with burn-ups in excess of 7–9% FIMA had a reopening of the fuel-to-cladding gap and evidence of joint oxide-gain (JOG) formation. Signs of axial fuel migration to the top of the fuel column were observed in the fuel pin with a peak burn-up of 23.7% FIMA. Additionally, high burn-up structure (HBS) was observed in the two highest burn-up samples (23.7% and 21.3% FIMA). The HBS layers were found to be 3–5 times thicker than the layers found in typical LWR fuel. The results of the study indicate that formation of JOG and or HBS prevents any significant fuel-cladding mechanical interaction from occurring, thereby extending the potential life of the fuel elements.
BD Hanson; J Abrefah; SC Marschman; SG Prussin
2000-01-01
The main objective of this project is to develop an advanced fuel matrix capable of achieving extended burnup while improving safety margins and reliability for present operations. In the course of this project, the authors improve understanding of the mechanism for high burnup structure (HBS) formation and attempt to design a fuel to minimize its formation. The use of soluble
Extension of the TRANSURANUS burnup model to heavy water reactor conditions
K. Lassmann; C. T. Walker; J. van de Laar
1998-01-01
The extension of the light water reactor burnup equations of the TRANSURANUS code to heavy water reactor conditions is described. Existing models for the fission of 235U and the buildup of plutonium in a heavy water reactor are evaluated. In order to overcome the limitations of the frequently used RADAR model at high burnup, a new model is presented. After
Numerical analysis for microstructure change of a light water reactor fuel pellet at high burnup
Takanori Kameyama; Tetsuo Matsumura; Motoyasu Kinoshita
1994-01-01
The peripheral region of a high burnup light water reactor (LWR) fuel pellet shows a microstructure that is different from the as-fabricated microstructure. The region where the microstructure change occurs (the rim region) is highly porous, and the original grains in the rim region are divided into much smaller subgrains. The electron probe microanalysis data of high burnup fuels indicate
Assessment of the use of extended burnup fuel in light water power reactors
D. A. Baker; W. J. Bailey; C. E. Beyer; F. C. Bold; J. J. Tawil
1988-01-01
This study has been conducted by Pacific Northwest Laboratory for the US Nuclear Regulatory Commission to review the environmental and economic impacts associated with the use of extended burnup nuclear fuel in light water power reactors. It has been proposed that current batch average burnup levels of 33 GWd\\/t uranium be increased to above 50 GWd\\/t. The environmental effects of
Analyse de l'impact de l'environnement dans un schema de calcul a deux etapes avec DRAGON et DONJON
NASA Astrophysics Data System (ADS)
Bodin, Christophe
The calculation of the neutron flux is an important data that is used to determine the dynamic of the core of a Pressurized Water Reactor (PWR). However the transport equation which gives the neutron flux, cannot be solved in three dimensions over the whole core, in evolution because of the power of the current computers, which are too slow. So some simplifications are necessary to calculate this flux. Two-levels schemes are used, where, in a first step, some macroscopic cross sections libraries are generated by solving the transport equation using infinite lattice calculations on two dimensions assemblies. These sections are generally homogenized on the whole assembly and condensed to two energy groups. In a second step, the whole core calculation is carried out using the diffusion equation, with the cross sections of the libraries previously generated, interpolated at the values of the different parameters. However the core of a PWR is made up of many assemblies, that can contain two types of fuel : Uranium OXyde (UOX) or plutonium and uranium Mixed OXyde (MOX). Moreover all these assemblies have different burnup because each one can be used for three or four cycles depending on the PWR. So that imply some burnup gradients. Thus the hypothesis of the infinite lattice used to generate the cross sections libraries can be highly inaccurate. The first goal of this project is to generate cross sections libraries that take into account the environment and to evaluate the impact of this heterogeneous environment on the core calculation. The flux obtained with the diffusion equation at the end of the core calculation is not accurate enough, du to the homogenization by assembly, to determine and to locate the hotspot factor, which represents an important industrial problematic. The principle of the power reconstruction method (PRM) is to reconstruct the more accurately possible the flux in the pins, with a combination of the diffusion flux and some microscopic flux which take into account the heterogeneities in the assemblies. This method is currently used with the data calculated with the infinite lattice. The second goal of this project is to develop a theory to apply the PRM with environmented data and to establish the PRM at the end of a calculation of the core and observe if the results are improved with the environmented data.
The burnup dependence of light water reactor spent fuel oxidation
Hanson, B.D.
1998-07-01
Over the temperature range of interest for dry storage or for placement of spent fuel in a permanent repository under the conditions now being considered, UO{sub 2} is thermodynamically unstable with respect to oxidation to higher oxides. The multiple valence states of uranium allow for the accommodation of interstitial oxygen atoms in the fuel matrix. A variety of stoichiometric and nonstoichiometric phases is therefore possible as the fuel oxidizers from UO{sub 2} to higher oxides. The oxidation of UO{sub 2} has been studied extensively for over 40 years. It has been shown that spent fuel and unirradiated UO{sub 2} oxidize via different mechanisms and at different rates. The oxidation of LWR spent fuel from UO{sub 2} to UO{sub 2.4} was studied previously and is reasonably well understood. The study presented here was initiated to determine the mechanism and rate of oxidation from UO{sub 2.4} to higher oxides. During the early stages of this work, a large variability in the oxidation behavior of samples oxidized under nearly identical conditions was found. Based on previous work on the effect of dopants on UO{sub 2} oxidation and this initial variability, it was hypothesized that the substitution of fission product and actinide impurities for uranium atoms in the spent fuel matrix was the cause of the variable oxidation behavior. Since the impurity concentration is roughly proportional to the burnup of a specimen, the oxidation behavior of spent fuel was expected to be a function of both temperature and burnup. This report (1) summarizes the previous oxidation work for both unirradiated UO{sub 2} and spent fuel (Section 2.2) and presents the theoretical basis for the burnup (i.e., impurity concentration) dependence of the rate of oxidation (Sections 2.3, 2.4, and 2.5), (2) describes the experimental approach (Section 3) and results (Section 4) for the current oxidation tests on spent fuel, and (3) establishes a simple model to determine the activation energies associated with spent fuel oxidation (Section 5).
Ankerst, Donna P.; Till, Cathee; Boeck, Andreas; Goodman, Phyllis; Tangen, Catherine M.; Feng, Ziding; Partin, Alan W.; Chan, Daniel W.; Sokoll, Lori; Kagan, Jacob; Wei, John T.; Thompson, Ian M.
2013-01-01
Purpose We assessed the independent predictive value of prostate volume, number of biopsy cores and AUASS (American Urological Association symptom score) compared to risk factors included in the PCPTRC (Prostate Cancer Prevention Trial risk calculator for prostate cancer) and PCPTHG (Prostate Cancer Prevention Trial risk calculator for high grade cancer [Gleason grade 7 or greater]). Materials and Methods Of 5,519 PCPT (Prostate Cancer Prevention Trial) participants the 4,958 used to construct the PCPTRC with AUASS and prostate specific antigen 10 ng/ml or less were included on logistic regression analysis. Risk algorithms were evaluated in 571 EDRN (Early Detection Research Network) participants using the ROC AUC. Results A total of 1,094 participants (22.1%) had prostate cancer, of whom 232 (21.2%) had high grade disease. For prostate cancer prediction higher prostate specific antigen, abnormal digital rectal examination, family history of prostate cancer and number of cores were associated with increased risk, while volume was associated with decreased risk. Excluding prostate volume and number of cores, a history of negative biopsy and increased AUASS were also associated with lower risk. For high grade cancer higher prostate specific antigen, abnormal digital rectal examination, black race and number of cores were associated with increased risk and volume, while AUASS was associated with decreased risk. The AUC of the PCPTRC adjusted for volume and number of cores was 72.7% using EDRN data and 68.2% when adjusted for AUASS alone vs 67.6% for the PCPTRC. For high grade disease the AUC was 74.8% and 74.0%, respectively, vs 73.5% for the PCPTHG. Conclusions Adjusted PCPT risk calculators for volume, number of cores and AUASS improve cancer detection. PMID:23313212
Generation of lumped fission product cross sections for high burnup, highly enriched uranium fuel
Primm, R.T. III; Greene, N.M.
1988-01-01
The first set of reactor design calculations for the reactor design considered here was performed with a depletion methodology developed for converter reactor studies. These analyses showed that the ANS reactor would have a cycle length of 14 days when operated at a power level of 270 MW. Since both the cycle length and the discharge fuel burnup (209,000 MWD/MT) are very different from any of the reactors for which the depletion methodology was developed, a new study of the depletion process was initiated. Since the expected cycle length and fuel loading (18.1 kg /sup 235/U) were known, input for an ORIGEN calculation could be prepared. For the work described here, cross section updates for the actinides and major fission products were prepared with data from an ENDF/B-V-derived library. The NITAWL-S and XSDRNPM-S codes were used to perform this update. The XSDRNPM model was a one-dimensional, buckled, cylindrical representation of the reactor. Fission yield values were derived from ENDF/B-IV data as contained in the ORIGEN Pressurized Water Reactor Library. 9 refs., 2 figs.
Assessment of the use of extended burnup fuel in light water power reactors
Baker, D.A.; Bailey, W.J.; Beyer, C.E.; Bold, F.C.; Tawil, J.J.
1988-02-01
This study has been conducted by Pacific Northwest Laboratory for the US Nuclear Regulatory Commission to review the environmental and economic impacts associated with the use of extended burnup nuclear fuel in light water power reactors. It has been proposed that current batch average burnup levels of 33 GWd/t uranium be increased to above 50 GWd/t. The environmental effects of extending fuel burnup during normal operations and during accident events and the economic effects of cost changes on the fuel cycle are discussed in this report. The physical effects of extended burnup on the fuel and the fuel assembly are also presented as a basis for the environmental and economic assessments. Environmentally, this burnup increase would have no significant impact over that of normal burnup. Economically, the increased burnup would have favorable effects, consisting primarily of a reduction: (1) total fuel requirements; (2) reactor downtime for fuel replacement; (3) the number of fuel shipments to and from reactor sites; and (4) repository storage requirements. 61 refs., 4 figs., 27 tabs.
Temperature and burnup correlated fuel-cladding chemical interaction in U-10ZR metallic fuel
NASA Astrophysics Data System (ADS)
Carmack, William J.
Metallic fuels are proposed for use in advanced sodium cooled fast reactors and provide a number of advantages over other fuel types considering their fabricability, performance, recyclability, and safety. Resistance to cladding "breach" and subsequent release of fission products and fuel constituents to the nuclear power plant primary coolant system is a key performance parameter for a nuclear fuel system. In metallic fuel, FCCI weakens the cladding, especially at high power-high temperature operation, contributing to fuel pin breach. Empirical relationships for FCCI have been developed from a large body of data collected from in-pile (EBR-II) and out-of-pile experiments [1]. However, these relationships are unreliable in predicting FCCI outside the range of EBR-II experimental data. This dissertation examines new FCCI data extracted from the MFF-series of prototypic length metallic fuel irradiations performed in the Fast Flux Test Facility (FFTF). The fuel in these assemblies operated a temperature and burnup conditions similar to that in EBR-II but with axial fuel height three times longer than EBR-II experiments. Comparing FCCI formation data from FFTF and EBR-II provides new insight into FCCI formation kinetics. A model is developed combining both production and diffusion of lanthanides to the fuel-cladding interface and subsequent reaction with the cladding. The model allows these phenomena to be influenced by fuel burnup (lanthanide concentrations) and operating temperature. Parameters in the model are adjusted to reproduce measured FCCI layer thicknesses from EBR-II and FFTF. The model predicts that, under appropriate conditions, rate of FCCI formation can be controlled by either fission product transport or by the reaction rate of the interaction species at the fuel-cladding interface. This dissertation will help forward the design of metallic fuel systems for advanced sodium cooled fast reactors by allowing the prediction of FCCI layer formation in full length reactor designs. This should improve lifetime prediction of fuel performance capability for new advanced sodium cooled fast reactors with extended core designs.
Burnup and feasibility study of low power density PWR's
Molins-Bartra
1981-01-01
Operational and safety problems of current Pressurized Water Reactors are often associated with the high power density level of the cores. An alternate use of current-design cores is proposed by reducing the power density.The effects should be improved safety, improved ore utilization, and improved operational characteristics. A scoping study is performed in order to define core parameters suitable for optimization
Dement'ev, V. G.; Oleinik, D. S.
2011-12-15
The Monte Carlo method has been used to simulate the neutron transport in nuclear reactors for over fifty years. Fast progress in computer power and development of more and more robust and reliable algorithms, codes, and nuclear databases allow solving more challenging problems, including three-dimensional (3D) simulations of full-scale reactor cores. Short descriptions of a full-scale 3D model of the VVER-1000 core and algorithms and methods implemented in the MCU-PD and BIPR-7A codes and a comparison of the calculations by each program as well as a comparison with experimental data are given in this paper.
A Metal Fuel Core Concept for 1000 MWt Advanced Burner Reactor
Yang, W.S.; Kim, T.K.; Grandy, C.
2007-07-01
This paper describes the core design and performance characteristics of a metal fuel core concept for a 1000 MWt Advanced Burner Reactor. A ternary metal fuel form of U-TRU-Zr was assumed with weapons grade plutonium feed for the startup core and TRU recovered from LWR spent fuel for the recycled equilibrium core. A compact burner core was developed by trade-off between the burnup reactivity loss and TRU conversion ratio, with a fixed cycle length of one-year. In the startup core, the average TRU enrichment is 15.5%, the TRU conversion ratio is 0.81, and the burnup reactivity loss over a cycle is 3.6% {delta}k. The heavy metal and TRU inventories are 13.1 and 2.0 metric tons, respectively. The average discharge burnup is 93 MWd/kg, and the TRU consumption rate is 55.5 kg/year. For the recycled equilibrium core, the average TRU enrichment is 22.1 %, the TRU conversion ratio is 0.73, and the burnup reactivity loss is 2.2% {delta}k. The TRU inventory and consumption rate are 2.9 metric tons and 81.6 kg/year, respectively. The evaluated reactivity coefficients provide sufficient negative feedbacks. The control systems provide shutdown margins that are more than adequate. The integral reactivity parameters for quasi-static reactivity balance analysis indicate favorable passive safety features, although detailed safety analyses are required to verify passive safety behavior. (authors)
NASA Technical Reports Server (NTRS)
Triner, J. E.
1979-01-01
The basic magnetic properties under various operating conditions encountered in the state-of-the-art DC-AC/DC converters are examined. Using a novel core excitation circuit, the basic B-H and loss characteristics of various core materials may be observed as a function of circuit configuration, frequency of operation, input voltage, and pulse-width modulation conditions. From this empirical data, a mathematical loss characteristics equation is developed to analytically predict the specific core loss of several magnetic materials under various waveform excitation conditions.
Determination of actinide and fission-product isotopes in very-high-burnup spent nuclear fuel.
Sullivan, V. S.; Bowers, D. L.; Clark, M. A.; Graczyk, D. G.; Tsai, Y.; Streets, W. E.; Vander Pol, M. H.; Billone, M. C.
2008-07-01
A work plan was desired that would produce data for a wide array of actinide and fission-product isotopes with reasonably good accuracy and relatively low cost. An analysis scheme involving a fairly small number of separations, dilutions, and measurement methods was used to generate information on 74 isotopes in two spent-fuel samples of >70 GWd/MTU burnup. Some of the measured isotopes are of high interest for burnup-credit evaluations and had not been reported previously for high-burnup fuels.
A Two-Step Approach to Uncertainty Quantification of Core Simulators
Yankov, Artem; Collins, Benjamin; Klein, Markus; Jessee, Matthew A.; Zwermann, Winfried; Velkov, Kiril; Pautz, Andreas; Downar, Thomas
2012-01-01
For the multiple sources of error introduced into the standard computational regime for simulating reactor cores, rigorous uncertainty analysis methods are available primarily to quantify the effects of cross section uncertainties. Two methods for propagating cross section uncertainties through core simulators are the XSUSA statistical approach and the “two-step” method. The XSUSA approach, which is based on the SUSA code package, is fundamentally a stochastic sampling method. Alternatively, the two-step method utilizes generalized perturbation theory in the first step and stochastic sampling in the second step. The consistency of these two methods in quantifying uncertainties in the multiplication factor andmore »in the core power distribution was examined in the framework of phase I-3 of the OECD Uncertainty Analysis in Modeling benchmark. With the Three Mile Island Unit 1 core as a base model for analysis, the XSUSA and two-step methods were applied with certain limitations, and the results were compared to those produced by other stochastic sampling-based codes. Based on the uncertainty analysis results, conclusions were drawn as to the method that is currently more viable for computing uncertainties in burnup and transient calculations.« less
Comparison of XSUSA and 'two-step' approaches for full-core uncertainty quantification
Yankov, A. [Univ. of Michigan, 2355 Bonisteel Blvd, Ann Arbor, MI 48109 (United States); Klein, M. [Gesellschaft fuer Anlagen- und Reaktorsicherheit GRS MbH, Boltzmannstr. 14, D- 85748 Garching b. Muenchen (Germany); Jessee, M. A. [Oak Ridge National Laboratory (United States); Zwermann, W.; Velkov, K.; Pautz, A. [Gesellschaft fuer Anlagen- und Reaktorsicherheit GRS MbH, Boltzmannstr. 14, D- 85748 Garching b. Muenchen (Germany); Collins, B.; Downar, T. [Univ. of Michigan (United States)
2012-07-01
While there are multiple sources of error that are introduced into the standard computational regime for simulating reactor cores, rigorous uncertainty analysis methods are available primarily for quantifying the effects of cross section uncertainties. Two methods for propagating cross section uncertainties through core simulators are the XSUSA statistical approach and the 'Two-Step' method. The XSUSA approach, which is based on the SUSA code package, is fundamentally a stochastic sampling method. Alternatively, the Two-Step method utilizes generalized perturbation theory in the first step and stochastic sampling in the second step. The consistency of these two methods in quantifying uncertainties in the multiplication factor and in the core power distribution will be examined in the framework of phase 1-3 of the UAM Benchmark. Using the TMI core as a base model for analysis, the XSUSA and Two-Step methods are applied with certain limitations and the results are compared to those produced by other stochastic sampling-based codes. Based on the uncertainty analysis results, conclusions are made for which method is currently a more viable option for computing uncertainties in burnup and transient calculations. (authors)
Comparison of XSUSA and "Two-Step" Approaches for Full Core Uncertainty Quantification
Yankov, Artem [University of Michigan; Klein, M. [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS); Jessee, Matthew Anderson [ORNL; Zwermann, W. [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS); Velkov, Kiril [Gesellschaft fur Anlagen; Pautz, Andreas [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS); Collins, Benjamin [University of Michigan; Downar, Thomas [University of Michigan
2012-01-01
While there are multiple sources of error that are introduced into the standard computational regime for simulating reactor cores, rigorous uncertainty analysis methods are available primarily for quantifying the effects of cross section uncertainties. Two methods for propagating cross section uncertainties through core simulators are the XSUSA statistical approach and the 'Two-Step' method. The XSUSA approach, which is based on the SUSA code package, is fundamentally a stochastic sampling method. Alternatively, the Two-Step method utilizes generalized perturbation theory in the first step and stochastic sampling in the second step. The consistency of these two methods in quantifying uncertainties in the multiplication factor and in the core power distribution will be examined in the framework of phase I-3 of the UAM Benchmark. Using the TMI core as a base model for analysis, the XSUSA and Two-Step methods are applied with certain limitations and the results are compared to those produced by other stochastic sampling-based codes. Based on the uncertainty analysis results, conclusions are made for which method is currently a more viable option for computing uncertainties in burnup and transient calculations.
Ji-Young Lee; Ji-Won Kim; Seung-Ryul Moon; Jung-Hwan Chang; Shi-Uk Chung; Do-Hyun Kang; Jung-Pyo Hong
2009-01-01
This paper deals with a method for dynamic characteristic analysis considering core losses in transverse flux linear machines (TFLMs) with solid cores. This paper focuses on how to calculate the core losses of solid cores and how to apply the core losses to the dynamic simulation. The magnetic field characteristics, which are used for the core loss calculation and dynamic
Development and characteristics of the rim region in high burnup UO 2 fuel pellets
NASA Astrophysics Data System (ADS)
Cunningham, M. E.; Freshley, M. D.; Lanning, D. D.
1992-06-01
Research to define the behavior of Zircaloy-clad light-water reactor fuel irradiated to high burnup levels was conducted as part of the High Burnup Effects Program. One observed microstructural change related to irradiation to high burnup levels (up to 83 MWd/kgM pellet-average) is the development of a well-defined, unique microstructural region at the fuel pellet edge (rim). This rim region is characterized by the loss of optically-definable grain structure, increased porosity, and the depletion of matrix fission gas. The rim region holds the potential for significant localized increases in the athermal release of fission gases at high burnup levels, but the contribution of the rim release to the fractional release of the total fission gas produced in the rod is small.
Design strategies for optimizing high burnup fuel in pressurized water reactors
Xu, Zhiwen, 1975-
2003-01-01
This work is focused on the strategy for utilizing high-burnup fuel in pressurized water reactors (PWR) with special emphasis on the full array of neutronic considerations. The historical increase in batch-averaged discharge ...
Development of Technical Basis for Burnup Credit Regulatory Guidance in the United States
Parks, Cecil V [ORNL] [ORNL; Wagner, John C [ORNL] [ORNL; Mueller, Don [ORNL] [ORNL; Gauld, Ian C [ORNL] [ORNL
2011-01-01
In the United States (U.S.) there has been and continues to be considerable interest in the increased use of burnup credit as part of the safety basis for SNF systems and this interest has motivated numerous technical studies related to the application of burnup credit for maintaining subcriticality. Responding to industry requests and needs, the U.S. Nuclear Regulatory Commission initiated a burnup credit research program, with support from the Oak Ridge National Laboratory, to develop regulatory guidance and the supporting technical basis for allowing and expanding the use of burnup credit in pressurized-water reactor SNF storage and transport applications. The objective of this paper is to summarize the work and significant accomplishments, with references to the technical reports and publications for complete details.
Core-core and core-valence correlation
NASA Technical Reports Server (NTRS)
Bauschlicher, Charles W., Jr.; Langhoff, Stephen R.; Taylor, Peter R.
1988-01-01
The effect of (1s) core correlation on properties and energy separations was analyzed using full configuration-interaction (FCI) calculations. The Be 1 S - 1 P, the C 3 P - 5 S and CH+ 1 Sigma + or - 1 Pi separations, and CH+ spectroscopic constants, dipole moment and 1 Sigma + - 1 Pi transition dipole moment were studied. The results of the FCI calculations are compared to those obtained using approximate methods. In addition, the generation of atomic natural orbital (ANO) basis sets, as a method for contracting a primitive basis set for both valence and core correlation, is discussed. When both core-core and core-valence correlation are included in the calculation, no suitable truncated CI approach consistently reproduces the FCI, and contraction of the basis set is very difficult. If the (nearly constant) core-core correlation is eliminated, and only the core-valence correlation is included, CASSCF/MRCI approached reproduce the FCI results and basis set contraction is significantly easier.
Scaglione, John M [ORNL] [ORNL; Mueller, Don [ORNL] [ORNL; Wagner, John C [ORNL] [ORNL
2011-01-01
One of the most significant remaining challenges associated with expanded implementation of burnup credit in the United States is the validation of depletion and criticality calculations used in the safety evaluation - in particular, the availability and use of applicable measured data to support validation, especially for fission products. Applicants and regulatory reviewers have been constrained by both a scarcity of data and a lack of clear technical basis or approach for use of the data. U.S. Nuclear Regulatory Commission (NRC) staff have noted that the rationale for restricting their Interim Staff Guidance on burnup credit (ISG-8) to actinide-only is based largely on the lack of clear, definitive experiments that can be used to estimate the bias and uncertainty for computational analyses associated with using burnup credit. To address the issue of validation, the NRC initiated a project with the Oak Ridge National Laboratory to (1) develop and establish a technically sound validation approach (both depletion and criticality) for commercial spent nuclear fuel (SNF) criticality safety evaluations based on best-available data and methods and (2) apply the approach for representative SNF storage and transport configurations/conditions to demonstrate its usage and applicability, as well as to provide reference bias results. The purpose of this paper is to describe the criticality (k{sub eff}) validation approach, and resulting observations and recommendations. Validation of the isotopic composition (depletion) calculations is addressed in a companion paper at this conference. For criticality validation, the approach is to utilize (1) available laboratory critical experiment (LCE) data from the International Handbook of Evaluated Criticality Safety Benchmark Experiments and the French Haut Taux de Combustion (HTC) program to support validation of the principal actinides and (2) calculated sensitivities, nuclear data uncertainties, and the limited available fission product LCE data to predict and verify individual biases for relevant minor actinides and fission products. This paper (1) provides a detailed description of the approach and its technical bases, (2) describes the application of the approach for representative pressurized water reactor and boiling water reactor safety analysis models to demonstrate its usage and applicability, (3) provides reference bias results based on the prerelease SCALE 6.1 code package and ENDF/B-VII nuclear cross-section data, and (4) provides recommendations for application of the results and methods to other code and data packages.
Failure of high burnup fuels under reactivity-initiated accident conditions
Tomoyuki Sugiyama; Miki Umeda; Toyoshi Fuketa; Hideo Sasajima; Yutaka Udagawa; Fumihisa Nagase
2009-01-01
Pulse irradiation experiments of high burnup light-water-reactor fuels were performed to assess the fuel failure limit in a postulated reactivity-initiated accident (RIA). A BWR-UO2 rod at a burnup of 69GWd\\/t failed due to pellet-cladding mechanical interaction (PCMI) in the test LS-1. The fuel enthalpy at which fuel failure occurred was comparable to those for PWR-UO2 rods of 71 to 77GWd\\/t
Radiation-induced microstructural change in high burnup UO2 fuel pellets
K. Nogita; K. Une
1994-01-01
The formation mechanism of a unique microstructure, the rim structure, in high burnup UO2 fuels has been elucidated by transmission electron microscopy (TEM). Specimens were prepared from the fuel peripheral region, using pellets which had been irradiated to a wide range of burnups (6-83 GWd\\/t; 10 GWd\\/t = 2.5 × 1020 fissions\\/cm3) in light water reactors. Dislocation density and volume
Research Program to Elucidate Outside-in Failure of High Burnup Fuel Cladding
Hiroshi HAYASHI; Keizo OGATA; Toshikazu BABA; Katsuichiro KAMIMURA
2006-01-01
Among a series of power ramp tests on 25 Zr-lined segment rods of burnup ranging from 43 to 61 GWd\\/t, five segment rods failed during the power ramp tests. One segment rod irradiated for 3 cycles (43 GWd\\/t) failed with a pinhole due to PCI\\/SCC. The rest of higher burnups failed with an axial crack on the outer surface. The
Vetter, Frederick J.
calculators hp calculators HP 50g Calculations involving plots Plotting on the HP 50g The 2D/3D;hp calculators HP 50g Calculations involving plots hp calculators - 2 - HP 50g Calculations involving plots Plotting on the HP 50g The HP 50g calculator provides a host of plots to allow the user
Wenzel, Jan; Wormit, Michael; Dreuw, Andreas
2014-10-01
Core-level excitations are generated by absorption of high-energy radiation such as X-rays. To describe these energetically high-lying excited states theoretically, we have implemented a variant of the algebraic-diagrammatic construction scheme of second-order ADC(2) by applying the core-valence separation (CVS) approximation to the ADC(2) working equations. Besides excitation energies, the CVS-ADC(2) method also provides access to properties of core-excited states, thereby allowing for the calculation of X-ray absorption spectra. To demonstrate the potential of our implementation of CVS-ADC(2), we have chosen medium-sized molecules as examples that have either biological importance or find application in organic electronics. The calculated results of CVS-ADC(2) are compared with standard TD-DFT/B3LYP values and experimental data. In particular, the extended variant, CVS-ADC(2)-x, provides the most accurate results, and the agreement between the calculated values and experiment is remarkable. PMID:25130619
NASA Astrophysics Data System (ADS)
Martinez, J. S.; Zwermann, W.; Gallner, L.; Puente-Espel, F.; Cabellos, O.; Velkov, K.; Hannstein, V.
2014-04-01
Propagation of nuclear data uncertainties in reactor calculations is interesting for design purposes and libraries evaluation. Previous versions of the GRS XSUSA library propagated only neutron cross section uncertainties. We have extended XSUSA uncertainty assessment capabilities by including propagation of fission yields and decay data uncertainties due to the their relevance in depletion simulations. We apply this extended methodology to the UAM6 PWR Pin-Cell Burnup Benchmark, which involves uncertainty propagation through burnup.
Burnup simulations and spent fuel characteristics of ZrO 2 based inert matrix fuels
NASA Astrophysics Data System (ADS)
Schneider, E. A.; Deinert, M. R.; Herring, S. T.; Cady, K. B.
2007-03-01
Reducing the inventory of long lived isotopes that are contained in spent nuclear fuel is essential for maximizing repository capacity and extending the lifetime of related storage. Because of their non-fertile matrices, inert matrix fuels (IMF's) could be an ideal vehicle for using light-water reactors to help decrease the inventory of plutonium and other transuranics (neptunium, americium, curium) that are contained within spent uranium oxide fuel (UOX). Quantifying the characteristics of spent IMF is therefore of fundamental importance to determining its effect on repository design and capacity. We consider six ZrO 2 based IMF formulations with different transuranic loadings in a 1-8 IMF to UOX pin-cell arrangement. Burnup calculations are performed using a collision probability model where transport of neutrons through space is modeled using fuel to moderator transport and escape probabilities. The lethargy dependent neutron flux is treated with a high resolution multigroup thermalization method. The results of the reactor physics model are compared to a benchmark case performed with Montebruns and indicate that the approach yields reliable results applicable to high-level analyses of spent fuel isotopics. The data generated show that a fourfold reduction in the radiological and integrated thermal output is achievable in single recycle using IMF, as compared to direct disposal of an energy equivalent spent UOX.
Fuel Burnup and Fuel Pool Shielding Analysis for Bushehr Nuclear Reactor VVER-1000
NASA Astrophysics Data System (ADS)
Hadad, Kamal; Ayobian, Navid
Bushehr Nuclear power plant (BNPP) is currently under construction. The VVER-1000 reactor will be loaded with 126 tons of about 4% enriched fuel having 3-years life cycle. The spent fuel (SF) will be transferred into the spent fuel pool (SPF), where it stays for 8 years before being transferred to Russia. The SPF plays a crucial role during 8 years when the SP resides in there. This paper investigates the shielding of this structure as it is designed to shield the SF radiation. In this study, the SF isotope inventory, for different cycles and with different burnups, was calculated using WIMS/4D transport code. Using MCNP4C nuclear code, the intensity of ? rays was obtained in different layers of SFP shields. These layers include the water above fuel assemblies (FA) in pool, concrete wall of the pool and water laid above transferring fuels. Results show that ? rays leakage from the shield in the mentioned layers are in agreement with the plant's PSAR data. Finally we analyzed an accident were the water height above the FA in the pool drops to 47 cm. In this case it was observed that exposure dose above pool, 10 and 30 days from the accident, are still high and in the levels of 1000 and 758 R/hr.
NASA Technical Reports Server (NTRS)
Pao, J. L.; Mehrotra, S. C.; Lan, C. E.
1982-01-01
A computer code base on an improved vortex filament/vortex core method for predicting aerodynamic characteristics of slender wings with edge vortex separations is developed. The code is applicable to camber wings, straked wings or wings with leading edge vortex flaps at subsonic speeds. The prediction of lifting pressure distribution and the computer time are improved by using a pair of concentrated vortex cores above the wing surface. The main features of this computer program are: (1) arbitrary camber shape may be defined and an option for exactly defining leading edge flap geometry is also provided; (2) the side edge vortex system is incorporated.
Sensitivity study on Xe depletion in the high burn-up structure of UO2
NASA Astrophysics Data System (ADS)
Holt, L.; Schubert, A.; Van Uffelen, P.; Walker, C. T.; Fridman, E.; Sonoda, T.
2014-09-01
Experimental results for the Xe depletion in the matrix of high burn-up fuel are presented from the High Burnup Rim Project (HBRP). In this project a number of UO2 fuel discs with 235U enrichment of 25.8 wt.% were irradiated. The Xe content of the fuel discs was analysed by means of electron probe microanalysis (EPMA). The influence of the burn-up and irradiation temperature on the Xe concentration was investigated using a multi-physics approach involving various simulation tools. The temperature influence was modelled by means of the temperature dependent effective burn-up. Good agreement was found between the modelled temperature threshold of the effective burn-up and the experimental temperature threshold between un- and restructured fuel in the HBRP. However, a systematic difference is observed between the onset burn-up derived from the Xe measurements in highly enriched discs such as those of HBRP and the corresponding values derived from irradiated Light Water Reactor (LWR) fuel rods and reported in the open literature. A sensitivity study identified the neutron flux spectrum and the fission product yields as the main reasons for the observed differences.
Calculation of the Phenix end-of-life test 'Control Rod Withdrawal' with the ERANOS code
Tiberi, V.
2012-07-01
The Inst. of Radiological Protection and Nuclear Safety (IRSN) acts as technical support to French public authorities. As such, IRSN is in charge of safety assessment of operating and under construction reactors, as well as future projects. In this framework, one current objective of IRSN is to evaluate the ability and accuracy of numerical tools to foresee consequences of accidents. Neutronic studies step in the safety assessment from different points of view among which the core design and its protection system. They are necessary to evaluate the core behavior in case of accident in order to assess the integrity of the first barrier and the absence of a prompt criticality risk. To reach this objective one main physical quantity has to be evaluated accurately: the neutronic power distribution in core during whole reactor lifetime. Phenix end of life tests, carried out in 2009, aim at increasing the experience feedback on sodium cooled fast reactors. These experiments have been done in the framework of the development of the 4. generation of nuclear reactors. Ten tests have been carried out: 6 on neutronic and fuel aspects, 2 on thermal hydraulics and 2 for the emergency shutdown. Two of them have been chosen for an international exercise on thermal hydraulics and neutronics in the frame of an IAEA Coordinated Research Project. Concerning neutronics, the Control Rod Withdrawal test is relevant for safety because it allows evaluating the capability of calculation tools to compute the radial power distribution on fast reactors core configurations in which the flux field is very deformed. IRSN participated to this benchmark with the ERANOS code developed by CEA for fast reactors studies. This paper presents the results obtained in the framework of the benchmark activity. A relatively good agreement was found with available measures considering the approximations done in the modeling. The work underlines the importance of burn-up calculations in order to have a fine core concentrations mesh for the calculation of the power distribution. (authors)
Core Optimization of a Deep-Burn Pebble Bed Reactor
Brian Boer; Abderrafi M. Ougouag
2010-06-01
Achieving a high fuel burnup in the Deep-Burn (DB) pebble bed reactor design, while remaining within the limits for fuel temperature, power peaking and temperature reactivity feedback, is challenging. The high content of Pu and Minor Actinides in the Deep-Burn fuel significantly impacts the thermal neutron energy spectrum as compared to a ’standard’ UO2 fueled core. This can result in power and temperature peaking in the pebble bed core in locally thermalized regions near the graphite reflectors. Furthermore, the interplay of the Pu resonances of the neutron absorption cross sections at low-lying energies can lead to a positive temperature reactivity coefficient for the graphite moderator at certain operating conditions. The DB concept focuses on the destruction of spent fuel transuranics in TRISO coated particle fueled gas-cooled reactors with the aim of a fractional fuel burnup of 60-70% in fissions per initial metal atom (FIMA), using a single-pass, multi in-core fuel (re)cycling scheme. In principle, the DB pebble bed concept employs the same reactor designs as the present low enriched uranium core designs, i.e. the 400 MWth Pebble Bed Modular Reactor (PBMR-400). A Pu and Minor Actinide fueled PBMR-400 design serves as the starting point for a core optimization study. The fuel temperature, power peak, temperature reactivity coefficients, and burnup capabilities of the modified designs are analyzed with the PEBBED code. A code-to-code coupling with the PASTA code allows for the analysis of the TRISO fuel performance for both normal and Loss Of Forced Cooling conditions. An improved core design is sought, maximizing the fuel discharge burnup, while retaining negative temperature reactivity feedback coefficients for the entire temperature range and avoiding high fuel temperatures (fuel failure probabilities).
Hybrid Reduced Order Modeling Algorithms for Reactor Physics Calculations
NASA Astrophysics Data System (ADS)
Bang, Youngsuk
Reduced order modeling (ROM) has been recognized as an indispensable approach when the engineering analysis requires many executions of high fidelity simulation codes. Examples of such engineering analyses in nuclear reactor core calculations, representing the focus of this dissertation, include the functionalization of the homogenized few-group cross-sections in terms of the various core conditions, e.g. burn-up, fuel enrichment, temperature, etc. This is done via assembly calculations which are executed many times to generate the required functionalization for use in the downstream core calculations. Other examples are sensitivity analysis used to determine important core attribute variations due to input parameter variations, and uncertainty quantification employed to estimate core attribute uncertainties originating from input parameter uncertainties. ROM constructs a surrogate model with quantifiable accuracy which can replace the original code for subsequent engineering analysis calculations. This is achieved by reducing the effective dimensionality of the input parameter, the state variable, or the output response spaces, by projection onto the so-called active subspaces. Confining the variations to the active subspace allows one to construct an ROM model of reduced complexity which can be solved more efficiently. This dissertation introduces a new algorithm to render reduction with the reduction errors bounded based on a user-defined error tolerance which represents the main challenge of existing ROM techniques. Bounding the error is the key to ensuring that the constructed ROM models are robust for all possible applications. Providing such error bounds represents one of the algorithmic contributions of this dissertation to the ROM state-of-the-art. Recognizing that ROM techniques have been developed to render reduction at different levels, e.g. the input parameter space, the state space, and the response space, this dissertation offers a set of novel hybrid ROM algorithms which can be readily integrated into existing methods and offer higher computational efficiency and defendable accuracy of the reduced models. For example, the snapshots ROM algorithm is hybridized with the range finding algorithm to render reduction in the state space, e.g. the flux in reactor calculations. In another implementation, the perturbation theory used to calculate first order derivatives of responses with respect to parameters is hybridized with a forward sensitivity analysis approach to render reduction in the parameter space. Reduction at the state and parameter spaces can be combined to render further reduction at the interface between different physics codes in a multi-physics model with the accuracy quantified in a similar manner to the single physics case. Although the proposed algorithms are generic in nature, we focus here on radiation transport models used in support of the design and analysis of nuclear reactor cores. In particular, we focus on replacing the traditional assembly calculations by ROM models to facilitate the generation of homogenized cross-sections for downstream core calculations. The implication is that assembly calculations could be done instantaneously therefore precluding the need for the expensive evaluation of the few-group cross-sections for all possible core conditions. Given the generic natures of the algorithms, we make an effort to introduce the material in a general form to allow non-nuclear engineers to benefit from this work.
Zhou, Shiqi; Solana, J R
2013-06-28
The first four perturbation coefficients in the expansion of the Helmholtz free energy in power series of the inverse of the reduced temperature for a number of potential models with hard-sphere cores plus core-softened and discontinuous tails are obtained from Monte Carlo simulations. The potential models considered include square-well, double square-well, and square-shoulder plus square-well, with different potential parameters. These simulation data are used to evaluate the performance of a traditional macroscopic compressibility approximation (MCA) for the second order coefficient and a recent coupling parameter series expansion (CPSE) for the first four coefficients. Comprehensive comparison indicates the incapability of the MCA for the second order coefficient in most non-stringent situations, and significance of the CPSE in accurately calculating these four coefficients. PMID:23822235
Irradiation performance of FBR Monju-type fuel with modified type 316 stainless steel at high burnup
Ukai, Shigeharu; Yoshitake, Ssunemitsu; Akasaka, Naoaki; Donomae, Takako; Katsuyama, Kozo; Mitsugi, Takeshi; Asaga, Takeo
1998-12-31
In the last three decades, extensive effort has been devoted to the development of austenitic stainless steels for the liquid-metal fast breeder reactor in Japan. In the driver fuels of the Joyo MK-II experimental fast reactor core, which has been operated successfully at 100 MW(thermal) since 1982, 127 fuel pins are assembled with wrapping wire into the hexagonal duct. The material for cladding, duct, and wire is 20% cold-worked modified Type 316 stainless steel (PNC316) and mixed-oxide (MOX) fuel pellets, with 29 wt% Pu content and 94% theoretical density (TD). The peak burnup and neutron dose at the current Joyo MK-II core has reached 84 GWd/tonne U and 50 dpa, respectively. The postirradiation examination of those assemblies showed excellent performance of the MOX fuel and a negligible pin diameter increase, which demonstrates that the conservative design for the Joyo MK-II driver fuels is satisfactory. A life-limiting factor of the austenitic steel fuel assemblies was demonstrated to be dominated by the swelling-induced bundle distortion in a duct. Specifically, for characterizing swelling of PNC316 materials, microstructure changes of the irradiated cladding and duct were extensively analyzed by means of transmission electron microscopy. In PNC316 the improvement of swelling resistance had been conducted by adjusting 20% cold working and by adding minor alloying elements such as P, B, Ti, and Nb within the specification range of chemical compositions in the standard Type 316 stainless steel.
Church, Stanley E.; Rice, Cyndi A.; Marot, Marci E.
2008-01-01
The U.S. Departments of Agriculture and Interior Abandoned Mine Lands (AML) Initiative is focused on the evaluation of the effect of past mining practices on the water quality and the riparian and aquatic habitats of impacted stream reaches downstream from historical mining districts located primarily on Federal lands. This problem is manifest in the eleven western states (west of longitude 102 degrees) where the majority of hardrock mines that had past production are located on Federal lands. In areas of temperate climate and moderate to heavy precipitation, the effects of rapid chemical and physical weathering of sulfides exposed on mine-waste dumps and acidic drainage from mines have resulted in elevated metal concentrations in the stream water and stream-bed sediment. The result of these mineral weathering processes has an unquantified impact on the quality of the water and the aquatic and riparian habitats that may limit their recreational resource value. One of the confounding factors in these studies is the determination of the component of metals derived from hydrothermally altered but unmined portions of these drainage basins. Several watersheds have been studied to evaluate the effects of acid mine drainage and acid rock drainage on the near-surface environment. The Animas River watershed in southwestern Colorado contains a large number of past-producing metal mines that have affected the watershed. Beginning in October 1996, the U.S. Geological Survey (USGS) began a collaborative study of these effects under the USGS-AML Initiative. In this report, we present the radionuclide and geochemical analytical results of sediment coring during 1997-1999 from two cores from oxbow lakes 0.5 mi. upstream from the 32nd Street Bridge near Durango, Colo., and from three cores from beaver ponds within the Mineral Creek drainage basin near Silverton, Colo.
Hans D. Gougar
2009-08-01
The Idaho National Laboratory’s deterministic neutronics analysis codes and methods were applied to the computation of the core multiplication factor of the HTR-Proteus pebble bed reactor critical facility. A combination of unit cell calculations (COMBINE-PEBDAN), 1-D discrete ordinates transport (SCAMP), and nodal diffusion calculations (PEBBED) were employed to yield keff and flux profiles. Preliminary results indicate that these tools, as currently configured and used, do not yield satisfactory estimates of keff. If control rods are not modeled, these methods can deliver much better agreement with experimental core eigenvalues which suggests that development efforts should focus on modeling control rod and other absorber regions. Under some assumptions and in 1D subcore analyses, diffusion theory agrees well with transport. This suggests that developments in specific areas can produce a viable core simulation approach. Some corrections have been identified and can be further developed, specifically: treatment of the upper void region, treatment of inter-pebble streaming, and explicit (multiscale) transport modeling of TRISO fuel particles as a first step in cross section generation. Until corrections are made that yield better agreement with experiment, conclusions from core design and burnup analyses should be regarded as qualitative and not benchmark quality.
Vetter, Frederick J.
calculators hp calculators HP 50g Working with Parametric Plots Plotting on the HP 50g Parametric calculators HP 50g Working with Parametric Plots hp calculators - 2 - HP 50g Working with Parametric Plots Plotting on the HP 50g The HP 50g calculator provides a host of plots to allow the user to visualize data
A Simplified Approach for Evaluation of the Burnup Potential of Alternative Fuels
Oggianu, Stella Maris [Massachusetts Institute of Technology (United States); No, Hee Cheon [Korea Advanced Institute of Science and Technology (Korea, Republic of); Kazimi, Mujid S. [Massachusetts Institute of Technology (United States)
2004-06-15
To evaluate the burnup potential of a fuel pin, a simplified fuel rod analysis code called FUEL-SIMplified model (FUELSIM) was developed using the general-purpose software VENSIM. FUELSIM is based on FRAPCON-3 models and validated against it. A sensitivity analysis was done using FUELSIM to determine the fuel parameters that have high importance in limiting the burnup potential of a fuel material. Among 16 parameters, 10 were identified as having high importance. For six fuel materials (uranium metal, UC, UN, Th/U metal, UO{sub 2}/ThO{sub 2} fuels, and UO{sub 2}), a simplified model for the pressure rise and volumetric changes inside the fuel is developed to estimate the operational index of each fuel; these models include only the variables with high importance. It was found that the highest burnup potential is that of the nitride fuel, followed by the UO{sub 2}/ThO{sub 2} fuel.
Validation of the new code package APOLLO2.8 for accurate PWR neutronics calculations
Santamarina, A.; Bernard, D.; Blaise, P.; Leconte, P.; Palau, J. M.; Roque, B.; Vaglio, C.; Vidal, J. F.
2013-07-01
This paper summarizes the Qualification work performed to demonstrate the accuracy of the new APOLLO2.S/SHEM-MOC package based on JEFF3.1.1 nuclear data file for the prediction of PWR neutronics parameters. This experimental validation is based on PWR mock-up critical experiments performed in the EOLE/MINERVE zero-power reactors and on P.I. Es on spent fuel assemblies from the French PWRs. The Calculation-Experiment comparison for the main design parameters is presented: reactivity of UOX and MOX lattices, depletion calculation and fuel inventory, reactivity loss with burnup, pin-by-pin power maps, Doppler coefficient, Moderator Temperature Coefficient, Void coefficient, UO{sub 2}-Gd{sub 2}O{sub 3} poisoning worth, Efficiency of Ag-In-Cd and B4C control rods, Reflector Saving for both standard 2-cm baffle and GEN3 advanced thick SS reflector. From this qualification process, calculation biases and associated uncertainties are derived. This code package APOLLO2.8 is already implemented in the ARCADIA new AREVA calculation chain for core physics and is currently under implementation in the future neutronics package of the French utility Electricite de France. (authors)
Post Irradiation Examination for Advanced Materials at Burnups Exceeding the Current Limit
John H. Strumpell
2004-12-31
Permitting fuel to be irradiated to higher burnups limits can reduce the amount of spent nuclear fuel (SNF) requiring storage and/or disposal and enable plants to operate with longer more economical cycle lengths and/or at higher power levels. Therefore, Framatome ANP (FANP) and the B&W Owner's Group (BWOG) have introduced a new fuel rod design with an advanced M5 cladding material and have irradiated several test fuel rods through four cycles. The U.S. Department of Energy (DOE) joined FANP and the BWOG in supporting this project during its final phase of collecting and evaluating high burnup data through post irradiation examination (PIE).
NASA Astrophysics Data System (ADS)
Petrov, Yu. V.; Nazarov, A. I.; Onegin, M. S.; Petrov, V. Yu.; Sakhnovsky, E. G.
2006-12-01
Using modern methods of reactor physics, we performed full-scale calculations of the Oklo natural reactor. For reliability, we used recent versions of two Monte Carlo codes: the Russian code MCU-REA and the well-known international code MCNP. Both codes produced similar results. We constructed a computer model of the Oklo reactor zone RZ2 which takes into account all details of design and composition. The calculations were performed for three fresh cores with different uranium contents. Multiplication factors, reactivities, and neutron fluxes were calculated. We also estimated the temperature and void effects for the fresh core. As would be expected, we found for the fresh core a significant difference between reactor and Maxwell spectra, which had been used before for averaging cross sections in the Oklo reactor. The averaged cross section of 62149Sm and its dependence on the shift of a resonance position Er (due to variation of fundamental constants) are significantly different from previous results. Contrary to the results of previous papers, we found no evidence of a change of the samarium cross section: a possible shift of the resonance energy is given by the limits -73??Er?62 meV. Following tradition, we have used formulas of Damour and Dyson to estimate the rate of change of the fine structure constant ?. We obtain new, more accurate limits of -4×10-17??·/??3×10-17yr-1. Further improvement of the accuracy of the limits can be achieved by taking account of the core burn-up. These calculations are in progress.
Alberto Talamo; Waclaw Gudowski; Francesco Venneri
2004-01-01
We have investigated the waste actinide burnup capabilities of a Gas Turbine Modular Helium Reactor (GT-MHR, similar to the reactor being designed by General Atomics and Minatom for surplus weapons plutonium destruction) with the Monte Carlo Continuous Energy Burnup Code MCB, an extension of MCNP developed at the Royal Institute of Technology in Stockholm and University of Mining and Metallurgy
Federal Register 2010, 2011, 2012, 2013, 2014
2013-11-12
...DEPARTMENT OF ENERGY Invitation for Public...for the High Burnup Dry Storage Cask Research and Development...Technologies, Office of Nuclear Energy, Department of Energy...Representative, High Burnup Dry Storage Cask Research and Development...U.S. Department of Energy--Idaho...
The Encapsulated Nuclear Heat Source (ENHS) Reactor Core Design
Hong, Ser Gi; Greenspan, Ehud; Kim, Yeong Il
2005-01-15
A once-for-life, uniform composition, blanket-free and fuel-shuffling-free reference core has been designed for the Encapsulated Nuclear Heat Source (ENHS) to provide the design goals of a nearly zero burnup reactivity swing throughout {approx}20 yr of full-power operation up to the peak discharge burnup of more than 100 GWd/t HM. What limits the core life is radiation damage to the HT-9 structural material. The temperature coefficients of reactivity are all negative, except for that of the coolant expansion. However, the negative reactivity coefficient associated with the radial expansion of the core structure can compensate for the coolant thermal expansion. The void coefficient is positive but of no safety concern because the boiling temperature of lead or lead-bismuth is so high that there is no conceivable mechanism for the introduction of significant void fraction into the core. The core reactivity coefficients, reactivity worth, and power distributions are almost constant throughout the core life.It was found possible to design such once-for-life cores using different qualities of Pu and transuranics as long as U is used as the primary fertile material. It is also feasible to design ENHS cores using nitride rather than metallic fuel. Relative to the reference metallic fuel core, nitride fuel cores offer up to {approx}25% higher discharge burnup and longer life, up to {approx}38% more energy per core, a significantly more negative Doppler reactivity coefficient, and less positive coolant expansion and coolant void reactivity coefficient but a somewhat smaller negative fuel expansion reactivity coefficient. The pitch-to-diameter ratio (1.45 of the nitride fuel cores using enriched N) is larger than that (1.36) for the reference metallic fuel core, implying a reduction of the coolant friction loss, thus enabling an increase in the power level that can be removed from the core by natural circulation cooling.It is also possible to design Pu-U(10Zr) fueled ENHS-type cores using Na as the primary coolant with either Na or Pb-Bi secondary coolants. The Na-cooled cores feature a tighter lattice and are therefore more compact but have spikier power distribution, more positive coolant temperature reactivity coefficients, and smaller reactivity worth of the control elements.
Vetter, Frederick J.
calculators hp calculators HP 50g The basics of plotting functions Plotting on the HP 50g The 2D/3D (PLOT SETUP) Form The WIN Form Examples of plotting functions #12;hp calculators HP 50g The basics of plotting functions hp calculators - 2 - HP 50g The basics of plotting functions Plotting on the HP 50g
A search for minimum volume of Breed and Burn cores
Di Sanzo, C.; Greenspan, E. [Dept. of Nuclear Engineering, Univ. of California, Berkeley Etcheverry Hall, Berkeley, CA 94720 (United States)
2012-07-01
The objective of the present study is to quantify the minimum volume a Breed and Burn (B and B) core can be designed to have and the corresponding burnup required for sustaining the breed-and-burn mode of operation based on neutronics; radiation damage constraints are ignored. The minimum radius for an idealized spherical B and B reactor is 136 cm or 110 cm for, respectively, 40% or 28% coolant volume fraction. The peak required burnup is about 25%. The minimum volume of a more realistic cylindrical B and B core is estimated to be only {approx}15% larger than that of the idealized spherical core but is only 43% of the volume of the medium-size B and B core previously designed to fit within the S-Prism reactor vessel. Thus it appears that SMR s can, in principle, be designed to have a B and B core. It was also found that the minimum volume B and B core does not necessarily coincide with the maximum permissible leakage from a core that can sustain the B and B mode of operation. (authors)
Core-core and core-valence correlation
NASA Technical Reports Server (NTRS)
Bauschlicher, Charles W., Jr.; Langhoff, Stephen R.; Taylor, Peter R.
1988-01-01
The effect of 1s core correlation on properties and energy separations are analyzed using full configuration-interaction (FCI) calculations. The Be1S - 1P, the C 3P - 5S,m and CH(+) 1Sigma(+) - 1Pi separations, and CH(+) spectroscopic constants, dipole moment, and 1Sigma(+) - 1Pi transition dipole moment have been studied. The results of the FCI calculations are compared to those obtained using approximate methods.
Control cell nuclear reactor core
Brown, R.E.; Crowther, R.L.; Fennern, L.E.; Gitnick, B.J.; Sawyer, C.D.; Specker, S.R.; Walters, K.V.
1981-08-25
A method and arrangement for operating a boiling water nuclear reactor and fuel assembly designs for such reactor wherein the reactor is operated with uniquely located control rods that serve the primary functions of power shaping and reactivity control. A second and separate distinct group of control rods is withdrawn when the reactor is at power and the control rods of this group serve the primary function of reactor shutdown. The design of the fuel assemblies and the selected patterns of fuel assemblies and control rods make the separation of control functions feasible. The power shaping and reactivity control control rods are located in low power regions of the core designated control cells. The design of the fuel is such that the control rods of the control cells may remain in fixed positions in these cells during the operating cycle until withdrawal for burnup reactivity compensation.
The Evolution of the First Core in Rotating Molecular Core
Kazuya Saigo; Kohji Tomisaka
2005-10-16
We present the evolution path of the first core in rotating molecular cloud core and the evolution during the second collapse phase using axisymmetric numerical calculations. The structure and evolution of the rotating first core is characterized by the angular momentum $J_{\\rm core}$ and mass $M_{\\rm core}$, both of which increase with time by accretion from the infalling envelope. We find the evolution path of it can be considered as a sequence of equilibrium solutions with a constant $J_{\\rm core}/M_{\\rm core}^2$. Such evolution paths are in good agreement with the results of three-dimensional numerical simulations. The evolution paths is completely different from spherical case. Except for extremely small rotation rate of $J_{\\rm core}/M_{\\rm core}^2 core does not increase in the central density beyond a certain maximum value and can not begins the second collapse direct and grows in mass by accretion. Such a first core has a very long time-scale and seems to suffer from a non-axisymmetric evolution. In the usual case, we expect that the time-scale of a rotating first core becomes several thousand years and several percent of the prestellar cores contain a first core at least. Such a rotating first core has a small average luminosity of $L_{\\rm core} = 0.003-0.03$($\\dot{M}_{\\rm core}/\\dot{M}_{\\rm ref}$) L$_{\\odot}$.
Evaluation of Cross-Section Sensitivities in Computing Burnup Credit Fission Product Concentrations
Gauld, I.C.
2005-08-12
U.S. Nuclear Regulatory Commission Interim Staff Guidance 8 (ISG-8) for burnup credit covers actinides only, a position based primarily on the lack of definitive critical experiments and adequate radiochemical assay data that can be used to quantify the uncertainty associated with fission product credit. The accuracy of fission product neutron cross sections is paramount to the accuracy of criticality analyses that credit fission products in two respects: (1) the microscopic cross sections determine the reactivity worth of the fission products in spent fuel and (2) the cross sections determine the reaction rates during irradiation and thus influence the accuracy of predicted final concentrations of the fission products in the spent fuel. This report evaluates and quantifies the importance of the fission product cross sections in predicting concentrations of fission products proposed for use in burnup credit. The study includes an assessment of the major fission products in burnup credit and their production precursors. Finally, the cross-section importances, or sensitivities, are combined with the importance of each major fission product to the system eigenvalue (k{sub eff}) to determine the net importance of cross sections to k{sub eff}. The importances established the following fission products, listed in descending order of priority, that are most likely to benefit burnup credit when their cross-section uncertainties are reduced: {sup 151}Sm, {sup 103}Rh, {sup 155}Eu, {sup 150}Sm, {sup 152}Sm, {sup 153}Eu, {sup 154}Eu, and {sup 143}Nd.
Double-strata high burnup fuel performance in light water reactors
Vladimir Barchevtsev; Vladimir Artisyuk; Hisashi Ninokata
2002-01-01
This study is aimed at the development of a fuel cycle concept for host countries with a lack of nuclear infrastructure. Two interrelated criteria, proliferation resistance and high-burnup, form the general framework of the fuel management scenario with the highest priority given to light water reactor technology and plutonium-free fresh fuel. Logically it implies the use of uranium oxide with
Nuclide Composition Benchmark Data Set for Verifying Burnup Codes on Spent Light Water Reactor Fuels
Yoshinori Nakahara; Jun Inagawa; Ryuji Nagaishi; Setsumi Kurosawa; Nobuaki Kohno; Mamoru Onuki; Hiroki Mochizuki
2002-01-01
To establish a nuclide composition benchmark data set for the verification of burnup codes, destructive analyses of light water reactor spent-fuel samples, which were cut out from several heights of spent-fuel rods, were carried out at the analytical laboratory at the Japan Atomic Energy Research Institute. The 16 samples from three kinds of pressurized water reactor (PWR) fuel rods and
Modeling the performance of high burnup thoria and urania PWR fuel
Long, Yun, 1972-
2002-01-01
Fuel performance models have been developed to assess the performance of ThO?-UO? fuels that can be operated to a high burnup up to 80-100MWd/kgHM in current and future Light Water Reactors (LWRs). Among the various issues ...
Pusa, M.; Leppaenen, J.
2012-07-01
The Chebyshev Rational Approximation Method (CRAM) has been recently introduced by the authors for solving the burnup equations with excellent results. This method has been shown to be capable of simultaneously solving an entire burnup system with thousands of nuclides both accurately and efficiently. The method was prompted by an analysis of the spectral properties of burnup matrices and it can be characterized as the best rational approximation on the negative real axis. The coefficients of the rational approximation are fixed and have been reported for various approximation orders. In addition to these coefficients, implementing the method only requires a linear solver. This paper describes an efficient method for solving the linear systems associated with the CRAM approximation. The introduced direct method is based on sparse Gaussian elimination where the sparsity pattern of the resulting upper triangular matrix is determined before the numerical elimination phase. The stability of the proposed Gaussian elimination method is discussed based on considering the numerical properties of burnup matrices. Suitable algorithms are presented for computing the symbolic factorization and numerical elimination in order to facilitate the implementation of CRAM and its adoption into routine use. The accuracy and efficiency of the described technique are demonstrated by computing the CRAM approximations for a large test case with over 1600 nuclides. (authors)
Burnup verification measurements at a US nuclear utility using the FORK measurement system
Ewing, R.I.; Bosler, G.E.; Walden, G.
1993-08-01
The FORK measurement system, designed at Los Alamos National Laboratory (LANL) for the International Atomic Energy Agency (IAEA) safeguards program, has been used to examine spent reactor fuel assemblies at Duke Power Company`s Oconee Nuclear Station. The FORK system measures the passive neutron and gamma-ray emission from spent fuel assemblies while in the storage pool. These measurements can be correlated with burnup and cooling time, and can be used to verify the reactor site records. Verification measurements may be used to help ensure nuclear criticality safety when burnup credit is applied to spent fuel transport and storage systems. By taking into account the reduced reactivity of spent fuel due to its burnup in the reactor, burnup credit results in more efficient and economic transport and storage. The objectives of these tests are to demonstrate the applicability of the FORK system to verify reactor records and to develop optimal procedures compatible with utility operations. The test program is a cooperative effort supported by Sandia National Laboratories, the Electric Power Research Institute (EPRI), Los Alamos National Laboratory, and the Duke Power Company.
Uwaba, Tomoyuki; Ito, Masahiro; Mizuno, Tomoyasu; Katsuyama, Kozo; Makenas, Bruce J.; Wootan, David W.; Carmack, Jon
2011-06-16
The ACO-3 irradiation test, which attained extremely high burnups of about 232 GWd/t and resisted a high neutron fluence (E > 0.1 MeV) of about 39E26 n/m2 as one of the lead tests of the Core Demonstration Experiment in the Fast Flux Test Facility, demonstrated that the fuel pin cladding made of ferritic/martensitic HT-9 alloy had superior void swelling resistance. The measured diameter profiles of the irradiated ACO-3 fuel pins showed axially extensive incremental strain in the MOX fuel column region and localized incremental strain near the interfaces between the MOX fuel and upper blanket columns. These incremental strains were as low as 1.5% despite the extremely high level of the fast neutron fluence. Evaluation of the pin diametral strain indicated that the incremental strain in the MOX fuel column region was substantially due to cladding void swelling and irradiation creep caused by internal fission gas pressure, while the localized strain near the MOX fuel/upper blanket interface was likely the result of the pellet/cladding mechanical interaction (PCMI) caused by cesium/fuel reactions. The evaluation also suggested that the PCMI was effectively mitigated by a large gap size between the cladding and blanket column.
Weapons-Grade MOX Fuel Burnup Characteristics in Advanced Test Reactor Irradiation
G. S. Chang
2006-07-01
Mixed oxide (MOX) test capsules prepared with weapons-derived plutonium have been irradiated to a burnup of 50 GWd/t. The MOX fuel was fabricated at Los Alamos National Laboratory (LANL) by a master-mix process and has been irradiated in the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL). Previous withdrawals of the same fuel have occurred at 9, 21, 30, 40, and 50 GWd/t. Oak Ridge National Laboratory (ORNL) manages this test series for the Department of Energy’s Fissile Materials Disposition Program (FMDP). A UNIX BASH (Bourne Again SHell) script CMO has been written and validated at the Idaho National Laboratory (INL) to couple the Monte Carlo transport code MCNP with the depletion and buildup code ORIGEN-2 (CMO). The new Monte Carlo burnup analysis methodology in this paper consists of MCNP coupling through CMO with ORIGEN-2(MCWO). MCWO is a fully automated tool that links the Monte Carlo transport code MCNP with the radioactive decay and burnup code ORIGEN-2. The fuel burnup analyses presented in this study were performed using MCWO. MCWO analysis yields time-dependent and neutron-spectrum-dependent minor actinide and Pu concentrations for the ATR small I-irradiation test position. The purpose of this report is to validate both the Weapons-Grade Mixed Oxide (WG-MOX) test assembly model and the new fuel burnup analysis methodology by comparing the computed results against the neutron monitor measurements and the irradiated WG-MOX post irradiation examination (PIE) data.
Test plan for high-burnup fuel cladding behavior under loss-of- coolant accident conditions
Chung, H.M.; Neimark, L.A.; Kassner, T.F.
1996-10-01
Excessive oxidation, hydriding, and extensive irradiation damage occur in high-burnup fuel cladding, and as result, mechanical properties of high-burnup fuels are degraded significantly. This may influence the current fuel cladding failure limits for loss-of- coolant-accident (LOCA) situations, which are based on fuel cladding behavior for zero burnup. To avoid cladding fragmentation and fuel dispersal during a LOCA, 10 CFR 50.46 requires that peak cladding temperature shall not exceed 1204 degrees C (2200 degrees F) and that total oxidation of the fuel cladding nowhere exceeds 0.17 times total cladding thickness before oxidation. Because of the concern, a new experimental program to investigate high-burnup fuel cladding behavior under LOCA situations has been initiated under the sponsorship of the U.S. Nuclear Regulatory Commission. A hot-cell test plan to investigate single-rod behavior under simulated LOCA conditions is described in this paper. In the meantime, industry fuel design and operating conditions are expected to undergo further changes as more advanced cladding materials are developed. Under these circumstances, mechanical properties of high-burnup fuel cladding require further investigation so that results from studies on LOCA, reactivity- initiated-accident (RIA), operational transient, and power-ramping situations, can be extrapolated to modified or advanced cladding materials and altered irradiation conditions without repeating major integral experiments in test reactors. To provide the applicable data base and mechanistic understanding, tests will be conducted to determine dynamic and static fracture toughness and tensile properties. Background and rationale for selecting the specific mechanical properties tests are also described.
NASA Technical Reports Server (NTRS)
1990-01-01
Spang & Company's new configuration of converter transformer cores is a composite of gapped and ungapped cores assembled together in concentric relationship. The net effect of the composite design is to combine the protection from saturation offered by the gapped core with the lower magnetizing requirement of the ungapped core. The uncut core functions under normal operating conditions and the cut core takes over during abnormal operation to prevent power surges and their potentially destructive effect on transistors. Principal customers are aerospace and defense manufacturers. Cores also have applicability in commercial products where precise power regulation is required, as in the power supplies for large mainframe computers.
Identifying and bounding uncertainties in nuclear reactor thermal power calculations
Phillips, J.; Hauser, E.; Estrada, H. [Cameron, 1000 McClaren Woods Drive, Coraopolis, PA 15108 (United States)
2012-07-01
Determination of the thermal power generated in the reactor core of a nuclear power plant is a critical element in the safe and economic operation of the plant. Direct measurement of the reactor core thermal power is made using neutron flux instrumentation; however, this instrumentation requires frequent calibration due to changes in the measured flux caused by fuel burn-up, flux pattern changes, and instrumentation drift. To calibrate the nuclear instruments, steam plant calorimetry, a process of performing a heat balance around the nuclear steam supply system, is used. There are four basic elements involved in the calculation of thermal power based on steam plant calorimetry: The mass flow of the feedwater from the power conversion system, the specific enthalpy of that feedwater, the specific enthalpy of the steam delivered to the power conversion system, and other cycle gains and losses. Of these elements, the accuracy of the feedwater mass flow and the feedwater enthalpy, as determined from its temperature and pressure, are typically the largest contributors to the calorimetric calculation uncertainty. Historically, plants have been required to include a margin of 2% in the calculation of the reactor thermal power for the licensed maximum plant output to account for instrumentation uncertainty. The margin is intended to ensure a cushion between operating power and the power for which safety analyses are performed. Use of approved chordal ultrasonic transit-time technology to make the feedwater flow and temperature measurements (in place of traditional differential-pressure- based instruments and resistance temperature detectors [RTDs]) allows for nuclear plant thermal power calculations accurate to 0.3%-0.4% of plant rated power. This improvement in measurement accuracy has allowed many plant operators in the U.S. and around the world to increase plant power output through Measurement Uncertainty Recapture (MUR) up-rates of up to 1.7% of rated power, while also decreasing the probability of significant over-power events. This paper will examine the basic elements involved in calculation of thermal power using ultrasonic transit-time technology and will discuss the criteria for bounding uncertainties associated with each element in order to achieve reactor thermal power calculations to within 0.3% to 0.4%. (authors)
NASA Astrophysics Data System (ADS)
Prytkov, A. N.; Tereshchenko, A. B.; Kravchenko, Yu. N.; Boldyrev, N. V.; Pozychaniuk, I. V.; Lisitsyn, D. I.; Golubev, E. I.
2014-04-01
In the course of upgrading the unit no. 5 reactor core of the Novovoronezh nuclear power plant, operational limits by local parameters, which limit the admissible linear power density and the relative power of fuel elements, were established. Due to applying modern computer technologies in systems of the in-core monitoring, the calculation of power density for all fuel elements in the real-time mode is implemented. To monitor the power density of fuel elements, the algorithm for determining the limiting linear power density is developed depending on the reactor core height and on the average nuclear fuel burnup. The admissible relative power of fuel elements is determined. In the course of the performed work, the excessive conservative limitations on nonuniformity of the reactor power density are excluded. The monitoring of power density by local parameters instead of indirect K q (fuel-assembly relative power) and K v (relative power of the fuel assembly section) made it possible to increase the fuel efficiency and to improve the economic parameters of fuel cycles of the unit no. 5 reactor core of the Novovoronezh nuclear power plant.
Evaluation of driver fuel performance in the Joyo Mk-II core
NASA Astrophysics Data System (ADS)
Asaga, T.; Shikakura, S.; Iwanaga, S.; Nomura, S.; Shibahara, I.; Katsuragawa, M.
1993-09-01
Experimental fast reactor "Joyo" attained first criticality of its Mk-II core (irradiation core) in 1982. Peak burnup and fluence has reached 83 MWd/kg M and 10 × 10 26 n/m 2 ( E > 0.1 MeV), respectively. Evaluation of the performance of Joyo Mk-II driver fuels has been carried out based on the post-irradiation examination (PIE) results including the highest burnup assembly in the core. The results showed that the performance of the Joyo Mk-II driver fuels was excellent and they were used with enough margin for the design criteria, suggesting the actual lifetime could be extended beyond the design limit. A large body of the PIE data of the driver assemblies and fuel pins made it possible to conduct a reliable evaluation of the fuel performance.
Pre-conceptual design study of ASTRID core
Varaine, F.; Marsault, P.; Chenaud, M. S.; Bernardin, B.; Conti, A.; Sciora, P.; Venard, C.; Fontaine, B.; Devictor, N.; Martin, L. [Alternative Energies and Atomic Energy Commission, CEA, DEN DER, 13108 St Paul lez Durance (France); Scholer, A. C.; Verrier, D. [AREVA-NP, 10 rue J. Recamier, 69456 Lyon Cedex 06 (France)
2012-07-01
In the framework of the ASTRID project at CEA, core design studies are performed at CEA with the AREVA and EDF support. At the stage of the project, pre-conceptual design studies are conducted in accordance with GEN IV reactors criteria, in particularly for safety improvements. An improved safety for a sodium cooled reactor requires revisiting many aspects of the design and is a rather lengthy process in current design approach. Two types of cores are under evaluation, one classical derived from the SFR V2B and one more challenging called CFV (low void effect core) with a large gain on the sodium void effect. The SFR V2b core have the following specifications: a very low burn-up reactivity swing (due to a small cycle reactivity loss) and a reduced sodium void effect with regard to past designs such as the EFR (around 2$ minus). Its performances are an average burn-up of 100 GWd/t, and an internal conversion ratio equal to one given a very good behavior of this core during a control rod withdrawal transient). The CFV with its specific design offers a negative sodium void worth while maintaining core performances. In accordance of ASTRID needs for demonstration those cores are 1500 MWth power (600 MWe). This paper will focus on the CFV pre-conceptual design of the core and S/A, and the performances in terms of safety will be evaluated on different transient scenario like ULOF, in order to assess its intrinsic behavior compared to a more classical design like V2B core. The gap in term of margin to a severe accident due to a loss of flow initiator underlines the potential capability of this type of core to enhance prevention of severe accident in accordance to safety demonstration. (authors)
The Challenges Associated with High Burnup and High Temperature for UO2 TRISO-Coated Particle Fuel
David Petti; John Maki
2005-02-01
The fuel service conditions for the DOE Next Generation Nuclear Plant (NGNP) will be challenging. All major fuel related design parameters (burnup, temperature, fast neutron fluence, power density, particle packing fraction) exceed the values that were qualified in the successful German UO2 TRISO-coated particle fuel development program in the 1980s. While TRISO-coated particle fuel has been irradiated at NGNP relevant levels for two or three of the design parameters, no data exist for TRISO-coated particle fuel for all five parameters simultaneously. Of particular concern are the high burnup and high temperatures expected in the NGNP. In this paper, where possible, we evaluate the challenges associated with high burnup and high temperature quantitatively by examining the performance of the fuel in terms of different known failure mechanisms. Potential design solutions to ameliorate the negative effects of high burnup and high temperature are also discussed.
Frolov
2000-09-01
Low-temperature (Tburn-up in deuterium-tritium mixtures with various deuterium-tritium-helium-3 ratios is considered. The general dependence is studied for the critical burn-up parameter x(c)=rhor(c) upon the initial temperature T, density rho(0), and tritium molar concentration y for the [D]:y[T]:(1-y)[3He] mixture. In particular, it is shown that, if the tritium concentration y decreases, then the critical burn-up parameter x(c)(T,rho(0),y) grows very quickly (at fixed T and rho(0)). This means that tritium beta(-) decay significantly complicates thermonuclear burn-up in deuterium-tritium mixtures. PMID:11088936
Kurchenkov, A. Yu.
2011-12-15
A method for determination of linear energy release of a VVER fuel assembly near a rhodium self-powered neutron detector (SPND) is described. The dependence of SPND burnup on the charge passing through it is specified.
NASA Astrophysics Data System (ADS)
Hipp, J. R.; Encarnacao, A.; Ballard, S.; Young, C. J.; Phillips, W. S.; Begnaud, M. L.
2011-12-01
Recently our combined SNL-LANL research team has succeeded in developing a global, seamless 3D tomographic P-velocity model (SALSA3D) that provides superior first P travel time predictions at both regional and teleseismic distances. However, given the variable data quality and uneven data sampling associated with this type of model, it is essential that there be a means to calculate high-quality estimates of the path-dependent variance and covariance associated with the predicted travel times of ray paths through the model. In this paper, we show a methodology for accomplishing this by exploiting the full model covariance matrix. Our model has on the order of 1/2 million nodes, so the challenge in calculating the covariance matrix is formidable: 0.9 TB storage for 1/2 of a symmetric matrix, necessitating an Out-Of-Core (OOC) blocked matrix solution technique. With our approach the tomography matrix (G which includes Tikhonov regularization terms) is multiplied by its transpose (GTG) and written in a blocked sub-matrix fashion. We employ a distributed parallel solution paradigm that solves for (GTG)-1 by assigning blocks to individual processing nodes for matrix decomposition update and scaling operations. We first find the Cholesky decomposition of GTG which is subsequently inverted. Next, we employ OOC matrix multiply methods to calculate the model covariance matrix from (GTG)-1 and an assumed data covariance matrix. Given the model covariance matrix we solve for the travel-time covariance associated with arbitrary ray-paths by integrating the model covariance along both ray paths. Setting the paths equal gives variance for that path. Sandia National Laboratories is a multi-program laboratory managed and operated by Sandia Corporation, a wholly owned subsidiary of Lockheed Martin Corporation, for the U.S. Department of Energy's National Nuclear Security Administration under contract DE-AC04-94AL85000.
THE EFFECT OF BURNUP AND SEPARATION EFFICIENCY ON URANIUM UTILIZATION AND RADIOTOXICITY
Samuel Bays; Steven Piet
2001-11-01
This paper addresses two fundamental issues of fuel cycle sustainability. The two primary issues of interest are efficient use of the natural uranium resource (cradle), and management of nuclear waste radiotoxicity (grave). Both uranium utilization and radiotoxicity are directly influenced by the burnup achieved during irradiation (transmutation related) and where applicable the separation efficiency (partitioning related). Burnup influences the in-growth of transuranics by breeding them into the fuel cycle. Transuranic breeding is virtually essential to resource sustainability because it increases utilization of naturally abundant fertile U-238. However, the direct consequence of this build-up is the in-growth of transuranic isotopes which generally increase the source of future geologically committed radiotoxicity. For scenarios involving recycle, separation efficiency influences the degree to which this transuranic source term is removed from active service in the fuel stream and made a disposal legacy of human activity.
Lanning, D.D.; Beyer, C.E.; Painter, C.L.
1997-12-01
This volume describes the fuel rod material and performance models that were updated for the FRAPCON-3 steady-state fuel rod performance code. The property and performance models were changed to account for behavior at extended burnup levels up to 65 Gwd/MTU. The property and performance models updated were the fission gas release, fuel thermal conductivity, fuel swelling, fuel relocation, radial power distribution, solid-solid contact gap conductance, cladding corrosion and hydriding, cladding mechanical properties, and cladding axial growth. Each updated property and model was compared to well characterized data up to high burnup levels. The installation of these properties and models in the FRAPCON-3 code along with input instructions are provided in Volume 2 of this report and Volume 3 provides a code assessment based on comparison to integral performance data. The updated FRAPCON-3 code is intended to replace the earlier codes FRAPCON-2 and GAPCON-THERMAL-2. 94 refs., 61 figs., 9 tabs.
Bailey
1989-01-01
The results of a study on extended-burnup, light-water reactor (LWR) spent fuel are described in this report. The study was performed by Pacific Northwest Laboratory for the US Department of Energy (DOE). The purpose of the study was to collect and evaluate information on the status of in-reactor performance and integrity of extended-burnup LWR fuel and initiate the investigation of
Fission product behavior in high-burnup water reactor fuel subjected to slow power increases
P. Knudsen; C. Bagger; H. Carlsen; I. Misfeldt; M. Mogensen
1986-01-01
Data are presented on fission gas release for UOâ-Zr fuel pins that were subjected to slow power increases late in life. These tests were performed with fuel pins that had been previously irradiated to average burnups of 27,000 to 35,000 MWd\\/ton U (peak pellet 43,700 MWd\\/ton U). The subsequent power increases were to 301 to 444 W\\/cm (peak pellet), and
Lars Olof JERNKVIST
2006-01-01
Best-estimate computational methods are here used to analyse the thermo-mechanical behaviour of high-burnup UO2 fuel rods under postulated reactivity initiated accidents in light water reactors. The considered accident scenarios are the hot zero power rod ejection accident in pressurised water reactors and the cold zero power control rod drop accident in boiling water reactors. For these accidents, fuel enthalpy thresholds
Advanced Corrosion-Resistant Zr Alloys for High Burnup and Generation IV Applications
Arthur Motta; Yong Hwan Jeong; R.J. Comstock; G.S. Was; Y.S. Kim
2006-10-31
The objective of this collaboration between four institutions in the US and Korea is to demonstrate a technical basis for the improvement of the corrosion resistance of zirconium-based alloys in more extreme operating environments (such as those present in severe fuel duty,cycles (high burnup, boiling, aggressive chemistry) andto investigate the feasibility (from the point of view of corrosion rate) of using advanced zirconium-based alloys in a supercritical water environment.
Analysis of Burnup and Economic Potential of Alternative Fuel Materials in Thermal Reactors
Oggianu, Stella Maris; No, Hee Cheon; Kazimi, Mujid S. [Massachusetts Institute of Technology (United States)
2003-09-15
A strategy is proposed for the assessment of nuclear fuel material economic potential use in future light water reactors (LWRs). In this methodology, both the required enrichment and the fuel performance limits are considered. In order to select the best fuel candidate, the optimal burnup that produces the lowest annual fuel cost within the burnup potential for a given fuel material and smear density ratio is determined.Several nuclear materials are presented as examples of the application of the methodology proposed in this paper. The alternative fuels considered include uranium dioxide (UO{sub 2}), uranium carbide (UC), uranium nitride (UN), metallic uranium (U-Zr alloy), combined thorium and uranium oxides (ThO{sub 2}/UO{sub 2}), and combined thorium and uranium metals (U/Th). For these examples, a typical LWR lattice geometry in a zirconium-based cladding was assumed. The uncertainties in the results presented are large due to the scarcity of experimental data regarding the behavior of the considered materials at high burnups. Also, chemical compatibility issues are to be considered separately.The same methodology can be applied in the future to evaluate the economic potential of other nuclear fuel materials including different cladding designs, dispersions of ceramics into ceramics, dispersions of ceramics into metals, and also for geometries other than the traditional circular fuel pin.
Comparison of nuclear data uncertainty propagation methodologies for PWR burn-up simulations
Carlos Javier Diez; Oliver Buss; Axel Hoefer; Dieter Porsch; Oscar Cabellos
2014-11-04
Several methodologies using different levels of approximations have been developed for propagating nuclear data uncertainties in nuclear burn-up simulations. Most methods fall into the two broad classes of Monte Carlo approaches, which are exact apart from statistical uncertainties but require additional computation time, and first order perturbation theory approaches, which are efficient for not too large numbers of considered response functions but only applicable for sufficiently small nuclear data uncertainties. Some methods neglect isotopic composition uncertainties induced by the depletion steps of the simulations, others neglect neutron flux uncertainties, and the accuracy of a given approximation is often very hard to quantify. In order to get a better sense of the impact of different approximations, this work aims to compare results obtained based on different approximate methodologies with an exact method, namely the NUDUNA Monte Carlo based approach developed by AREVA GmbH. In addition, the impact of different covariance data is studied by comparing two of the presently most complete nuclear data covariance libraries (ENDF/B-VII.1 and SCALE 6.0), which reveals a high dependency of the uncertainty estimates on the source of covariance data. The burn-up benchmark Exercise I-1b proposed by the OECD expert group "Benchmarks for Uncertainty Analysis in Modeling (UAM) for the Design, Operation and Safety Analysis of LWRs" is studied as an example application. The burn-up simulations are performed with the SCALE 6.0 tool suite.
Radiation-induced microstructural change in high burnup UO 2 fuel pellets
NASA Astrophysics Data System (ADS)
Nogita, K.; Une, K.
1994-06-01
The formation mechanism of a unique microstructure, the rim structure, in high burnup UO2 fuels has been elucidated by transmission electron microscopy (TEM). Specimens were prepared from the fuel peripheral region, using pellets which had been irradiated to a wide range of burnups (6-83 GWd/t; 10 GWd/t = 2.5 × 1020 fissions/cm3) in light water reactors. Dislocation density and volume fraction of intragranular bubbles increase with burnup. Low angle boundaries begin to form above 30-40 GWd/t. The TEM images and selected area electron diffraction (SAD) analyses of the rim structure observed in the 83 GWd/t fuel show: (1) sub-divided grains, 20-30 nm in size, with high angle boundaries due to the accumulation of an extremely high density of sub-boundaries; (2) recrystallized grains, 50-200 nm in size, adjacent to the sub-divided grain region, which are induced by the stored energy of the matrix; and (3) coarsened intragranular bubbles generated by radiation-induced excess vacancies.
Rim structure formation and high burnup fuel behavior of large-grained UO 2 fuels
NASA Astrophysics Data System (ADS)
Une, K.; Hirai, M.; Nogita, K.; Hosokawa, T.; Suzawa, Y.; Shimizu, S.; Etoh, Y.
2000-01-01
Irradiation-induced fuel microstructural evolution of the sub-divided grain structure, or rim structure, of large-grained UO 2 pellets has been examined through detailed PIEs. Besides standard grain size pellets with a grain size range of 9-12 ?m, two types of undoped and alumino-silicate doped large-grained pellets with a range of 37-63 ?m were irradiated in the Halden heavy water reactor up to a cross-sectional pellet average burnup of 86 GWd/t. The effect of grain size on the rim structure formation was quantitatively evaluated in terms of the average Xe depression in the pellet outside region measured by EPMA, based on its lower sensitivity for Xe enclosed in the coarsened rim bubbles. The Xe depression in the high burnup pellets above 60 GWd/t was proportional to d-0.5- d-1.0 ( d: grain size), and the two types of large-grained pellets showed remarkable resistance to the rim structure formation. A high density of dislocations preferentially decorated the as-fabricated grain boundaries and the sub-divided grain structure was localized there. These observations were consistent with our proposed formation mechanism of rim structure, in which tangled dislocation networks are organized into the nuclei for recrystallized or sub-divided grains. In addition to higher resistance to the microstructure change, the large-grained pellets showed a smaller swelling rate at higher burnups and a lower fission gas release during base irradiation.
Ilas, Germina; Gauld, Ian C
2011-01-01
This report is one of the several recent NUREG/CR reports documenting benchmark-quality radiochemical assay data and the use of the data to validate computer code predictions of isotopic composition for spent nuclear fuel, to establish the uncertainty and bias associated with code predictions. The experimental data analyzed in the current report were acquired from a high-burnup fuel program coordinated by Spanish organizations. The measurements included extensive actinide and fission product data of importance to spent fuel safety applications, including burnup credit, decay heat, and radiation source terms. Six unique spent fuel samples from three uranium oxide fuel rods were analyzed. The fuel rods had a 4.5 wt % {sup 235}U initial enrichment and were irradiated in the Vandellos II pressurized water reactor operated in Spain. The burnups of the fuel samples range from 42 to 78 GWd/MTU. The measurements were used to validate the two-dimensional depletion sequence TRITON in the SCALE computer code system.
Irradiation performance of PFBR MOX fuel after 112 GWd/t burn-up
NASA Astrophysics Data System (ADS)
Venkiteswaran, C. N.; Jayaraj, V. V.; Ojha, B. K.; Anandaraj, V.; Padalakshmi, M.; Vinodkumar, S.; Karthik, V.; Vijaykumar, Ran; Vijayaraghavan, A.; Divakar, R.; Johny, T.; Joseph, Jojo; Thirunavakkarasu, S.; Saravanan, T.; Philip, John; Rao, B. P. C.; Kasiviswanathan, K. V.; Jayakumar, T.
2014-06-01
The 500 MWe Prototype Fast Breeder Reactor (PFBR) which is in advanced stage of construction at Kalpakkam, India, will use mixed oxide (MOX) fuel with a target burnup of 100 GWd/t. The fuel pellet is of annular design to enable operation at a peak linear power of 450 W/cm with the requirement of minimum duration of pre-conditioning. The performance of the MOX fuel and the D9 clad and wrapper material was assessed through Post Irradiation Examinations (PIE) after test irradiation of 37 fuel pin subassembly in Fast Breeder Test Reactor (FBTR) to a burn-up of 112 GWd/t. Fission product distribution, swelling and fuel-clad gap evolution, central hole diameter variation, restructuring, fission gas release and clad wastage due to fuel-clad chemical interaction were evaluated through non-destructive and destructive examinations. The examinations have indicated that the MOX fuel can safely attain the desired target burn-up in PFBR.
NASA Astrophysics Data System (ADS)
Ramamoorthy, Karthikeyan
The main aim of this research is the development and validation of computational schemes for advanced lattice codes. The advanced lattice code which forms the primary part of this research is "DRAGON Version4". The code has unique features like self shielding calculation with capabilities to represent distributed and mutual resonance shielding effects, leakage models with space-dependent isotropic or anisotropic streaming effect, availability of the method of characteristics (MOC), burnup calculation with reaction-detailed energy production etc. Qualified reactor physics codes are essential for the study of all existing and envisaged designs of nuclear reactors. Any new design would require a thorough analysis of all the safety parameters and burnup dependent behaviour. Any reactor physics calculation requires the estimation of neutron fluxes in various regions of the problem domain. The calculation goes through several levels before the desired solution is obtained. Each level of the lattice calculation has its own significance and any compromise at any step will lead to poor final result. The various levels include choice of nuclear data library and energy group boundaries into which the multigroup library is cast; self shielding of nuclear data depending on the heterogeneous geometry and composition; tracking of geometry, keeping error in volume and surface to an acceptable minimum; generation of regionwise and groupwise collision probabilities or MOC-related information and their subsequent normalization thereof, solution of transport equation using the previously generated groupwise information and obtaining the fluxes and reaction rates in various regions of the lattice; depletion of fuel and of other materials based on normalization with constant power or constant flux. Of the above mentioned levels, the present research will mainly focus on two aspects, namely self shielding and depletion. The behaviour of the system is determined by composition of resonant and predominantly scattering isotopes. When the concentration of resonant isotopes is small, its presence does not affect the flux shape which is smooth. But when the concentration becomes high, there will be dips in the flux where resonances of the isotopes occur. This will affect the reaction rate, which is a product of cross section and flux. The reaction rate will thus be lower than that when one does not consider the flux dip. This is the phenomenon of self shielding. Self shielding treatment is thus a very important aspect of reactor lattice analysis code. This needs to be correctly modelled to obtain a physically sound and acceptable solution. In this research we will be looking into behaviour of the advanced self shielding models that have been incorporated in the code DRAGON Version4. The self shielding models are primarily classified into two broad groups, which are based on "equivalence in dilution" and "subgroup approach". These self shielding models will be tested against a variety of lattices which include Canada Deuterium Uranium (CANDU-6), CANDU-New Generation (CANDU-NG), Light Water Reactor (LWR), and High Conversion Light Water Reactor (HCLWR). The fuel composition will vary from natural uranium oxide to enriched uranium oxide and plutonium-uranium mixed oxide (MOX). We will also consider the presence of strong neutron absorbers like gadolinium and dysprosium in the lattice. The coolant/moderator chosen for the analysis will be light water/heavy water or a combination. The lattice geometry will vary from square, hexagonal and annular. Thus a broad spectrum of lattices will be analysed to assess the behaviour of advanced self shielding models. The results obtained using DRAGON will be validated against that obtained using Monte Carlo code MCNP5. The reference solutions for all situations will be provided by MCNP5. The depletion behaviour of any lattice will depend on the power or flux normalization that is considered. In general the flux in various regions is estimated with reference to a single neutron absorbed a
Automated Design and Optimization of Pebble-bed Reactor Cores
Hans D. Gougar; Abderrafi M. Ougouag; William K. Terry
2010-07-01
We present a conceptual design approach for high-temperature gas-cooled reactors using recirculating pebble-bed cores. The design approach employs PEBBED, a reactor physics code specifically designed to solve for and analyze the asymptotic burnup state of pebble-bed reactors, in conjunction with a genetic algorithm to obtain a core that maximizes a fitness value that is a function of user-specified parameters. The uniqueness of the asymptotic core state and the small number of independent parameters that define it suggest that core geometry and fuel cycle can be efficiently optimized toward a specified objective. PEBBED exploits a novel representation of the distribution of pebbles that enables efficient coupling of the burnup and neutron diffusion solvers. With this method, even complex pebble recirculation schemes can be expressed in terms of a few parameters that are amenable to modern optimization techniques. With PEBBED, the user chooses the type and range of core physics parameters that represent the design space. A set of traits, each with acceptable and preferred values expressed by a simple fitness function, is used to evaluate the candidate reactor cores. The stochastic search algorithm automatically drives the generation of core parameters toward the optimal core as defined by the user. The optimized design can then be modeled and analyzed in greater detail using higher resolution and more computationally demanding tools to confirm the desired characteristics. For this study, the design of pebble-bed high temperature reactor concepts subjected to demanding physical constraints demonstrated the efficacy of the PEBBED algorithm.
McCARD for Neutronics Design and Analysis of Research Reactor Cores
NASA Astrophysics Data System (ADS)
Shim, Hyung Jin; Park, Ho Jin; Kwon, Soonwoo; Seo, Geon Ho; Hyo Kim, Chang
2014-06-01
McCARD is a Monte Carlo (MC) neutron-photon transport simulation code developed exclusively for the neutronics design and analysis of nuclear reactor cores. McCARD is equipped with the hierarchical modeling and scripting functions, the CAD-based geometry processing module, the adjoint-weighted kinetics parameter and source multiplication factor estimation modules as well as the burnup analysis capability for the neutronics design and analysis of both research and power reactor cores. This paper highlights applicability of McCARD for the research reactor core neutronics analysis, as demonstrated for Kyoto University Critical Assembly, HANARO, and YALINA.
RIA Limits Based On Commercial PWR Core Response To RIA
Beard, Charles L.; Mitchell, David B.; Slagle, William H.
2006-07-01
Reactivity insertion accident (RIA) limits have been under intense review by regulators since 1993 with respect to what should be the proper limit as a function of burnup. Some national regulators have imposed new lower limits while in the United States the limits are still under review. The data being evaluated with respect to RIA limits come from specialized test reactors. However, the use of test reactor data needs to be balanced against the response of a commercial PWR core in setting reasonable limits to insure the health and safety of the public without unnecessary restrictions on core design and operation. The energy deposition limits for a RIA were set in the 1970's based on testing in CDC (SPERT), TREAT, PBF and NSRR test reactors. The US limits given in radially averaged enthalpy are 170 cal/gm for fuel cladding failure and 280 cal/gm for coolability. Testing conducted in the 1990's in the CABRI, NSRR and IGR test reactors have demonstrated that the cladding failure threshold is reduced with burnup, with the primary impact due to hydrogen pickup for in-reactor corrosion. Based on a review of this data very low enthalpy limits have been proposed. In reviewing proposed limits from RIL-0401(1) it was observed that much of the data used to anchor the low allowable energy deposition levels was from recent NSRR tests which do not represent commercial PWR reactor conditions. The particular characteristics of the NSRR test compared to commercial PWR reactor characteristics are: - Short pulse width: 4.5 ms vs > 8 ms; - Low temperature conditions: < 100 deg. F vs 532 deg. F. - Low pressure environment: atmospheric vs {approx} 2200 psi. A review of the historical RIA database indicates that some of the key NSRR data used to support the RIL was atypical compared to the overall RIA database. Based on this detailed review of the RIA database and the response of commercial PWR core, the following view points are proposed. - The Failure limit should reflect local fuel enthalpy and the condition of the fuel, i.e. burnup and expected oxide levels as a function of enthalpy. Limits based on cladding oxide needs to take into account that in many core designs the highest oxide will generally be on high burnup rods on the core periphery which have low reactivity, and lower peaking factors. Otherwise excessively low limits based generally on oxide could restrict use of fuel from the spent fuel pool. (authors)
Application of perturbation theory to lattice calculations based on method of cyclic characteristics
NASA Astrophysics Data System (ADS)
Assawaroongruengchot, Monchai
Perturbation theory is a technique used for the estimation of changes in performance functionals, such as linear reaction rate ratio and eigenvalue affected by small variations in reactor core compositions. Here the algorithm of perturbation theory is developed for the multigroup integral neutron transport problems in 2D fuel assemblies with isotropic scattering. The integral transport equation is used in the perturbative formulation because it represents the interconnecting neutronic systems of the lattice assemblies via the tracking lines. When the integral neutron transport equation is used in the formulation, one needs to solve the resulting integral transport equations for the flux importance and generalized flux importance functions. The relationship between the generalized flux importance and generalized source importance functions is defined in order to transform the generalized flux importance transport equations into the integro-differential equations for the generalized adjoints. Next we develop the adjoint and generalized adjoint transport solution algorithms based on the method of cyclic characteristics (MOCC) in DRAGON code. In the MOCC method, the adjoint characteristics equations associated with a cyclic tracking line are formulated in such a way that a closed form for the adjoint angular function can be obtained. The MOCC method then requires only one cycle of scanning over the cyclic tracking lines in each spatial iteration. We also show that the source importance function by CP method is mathematically equivalent to the adjoint function by MOCC method. In order to speed up the MOCC solution algorithm, a group-reduction and group-splitting techniques based on the structure of the adjoint scattering matrix are implemented. A combined forward flux/adjoint function iteration scheme, based on the group-splitting technique and the common use of a large number of variables storing tracking-line data and exponential values, is proposed to reduce the computing time when both direct and adjoint solutions are required. A problem that arises for the generalized adjoint problem is that the direct use of the negative external generalized adjoint sources in the adjoint solution algorithm results in negative generalized adjoint functions. A coupled flux biasing/decontamination scheme is applied to make the generalized adjoint functions positive using the adjoint functions in such a way that it can be used for the multigroup rebalance technique. Next we consider the application of the perturbation theory to the reactor problems. Since the coolant void reactivity (CVR) is a important factor in reactor safety analysis, we have decided to select this parameter for optimization studies. We consider the optimization and adjoint sensitivity techniques for the adjustments of CVR at beginning of burnup cycle (BOC) and k eff at end of burnup cycle (EOC) for a 2D Advanced CANDU Reactor (ACR) lattice. The sensitivity coefficients are evaluated using the perturbation theory based on the integral transport equations. Three sets of parameters for CVR-BOC and keff-EOC adjustments are studied: (1) Dysprosium density in the central pin with Uranium enrichment in the outer fuel rings, (2) Dysprosium density and Uranium enrichment both in the central pin, and (3) the same parameters as in the first case but the objective is to obtain a negative checkerboard CVR at beginning of cycle (CBCVR-BOC). To approximate the sensitivity coefficient at EOC, we perform constant-power burnup/depletion calculations for 600 full power days (FPD) using a slightly perturbed nuclear library and the unperturbed neutron fluxes to estimate the variation of nuclide densities at EOC. Sensitivity analyses of CVR and eigenvalue are included in the study. In addition the optimization and adjoint sensitivity techniques are applied to the CBCVR-BOC and keff-EOC adjustment of the ACR lattices with Gadolinium in the central pin. Finally we apply these techniques to the CVR-BOC, CVR-EOC and keff-EOC adjustment of a CANDU lattice of which the burnup period is extended f
Criticality Analysis of Assembly Misload in a PWR Burnup Credit Cask
Wagner, J. C.
2008-01-31
The Interim Staff Guidance on bumup credit (ISG-8) for spent fuel in storage and transportation casks, issued by the Nuclear Regulatory Commission's Spent Fuel Project Office, recommends a bumup measurement for each assembly to confirm the reactor record and compliance with the assembly bumup value used for loading acceptance. This recommendation is intended to prevent unauthorized loading (misloading) of assemblies due to inaccuracies in reactor burnup records and/or improper assembly identification, thereby ensuring that the appropriate subcritical margin is maintained. This report presents a computational criticality safety analysis of the consequences of misloading fuel assemblies in a highcapacity cask that relies on burnup credit for criticality safety. The purpose of this report is to provide a quantitative understanding of the effects of fuel misloading events on safety margins. A wide variety of fuel-misloading configurations are investigated and results are provided for informational purposes. This report does not address the likelihood of occurrence for any of the misload configurations considered. For representative, qualified bumup-enrichment combinations, with and without fission products included, misloading two assemblies that are underburned by 75% results in an increase in k_{eff} of 0.025-0.045, while misloading four assemblies that are underburned by 50% also results in an increase in k_{eff} of 0.025-0.045. For the cask and conditions considered, a reduction in bumup of 20% in all assemblies results in an increase in kff of less than 0.035. Misloading a single fresh assembly with 3, 4, or 5 wt% ^{235}U enrichment results in an increase in k_{eff}of--0.02, 0.04, or 0.06, respectively. The report concludes with a summary of these and other important findings, as well as a discussion of relevant issues that should be considered when assessing the appropriate role of burnup measurements.
Determining Spent Nuclear Fuel's Plutonium Content, Initial Enrichment, Burnup, and Cooling Time
Cheatham, Jesse R; Francis, Matthew W
2011-01-01
The Next Generation of Safeguards Initiative is examining nondestructive assay techniques to determine the total plutonium content in spent nuclear fuel. The goal of this research was to develop new techniques that can independently verify the plutonium content in a spent fuel assembly without relying on an operator's declarations. Fundamentally this analysis sought to answer the following questions: (1) do spent fuel assemblies contain unique, identifiable isotopic characteristics as a function of their burnup, cooling time, and initial enrichment; (2) how much variation can be seen in spent fuel isotopics from similar and dissimilar reactor power operations; and (3) what isotopes (if any) could be used to determine burnup, cooling time, and initial enrichment? To answer these questions, 96,000 ORIGEN cases were run that simulated typical two-cycle operations with burnups ranging from 21,900 to 72,000 MWd/MTU, cooling times from 5 to 25 years, and initial enrichments between 3.5 and 5.0 weight percent. A relative error coefficient was determined to show how numerically close a reference solution has to be to another solution for the two results to be indistinguishable. By looking at the indistinguishable solutions, it can be shown how a precise measurement of spent fuel isotopics can be inconclusive when used in the absence of an operator's declarations. Using this Method of Indistinguishable Solutions (MIS), we evaluated a prominent method of nondestructive analysis - gamma spectroscopy. From this analysis, a new approach is proposed that demonstrates great independent forensic examination potential for spent nuclear fuel by examining both the neutron emissions of Cm-244 and the gamma emissions of Cs-134 and Eu-154.
Modeling of fuel rod steady state and transient behavior over the full range of burnup
Yagnik, S.K.; Yang, R.L. [Electric Power Research Institute, Palo Alto, NC (United States); Rashid, Y.R.; Montgomery, R.O. [Anatech Research Corp., San Diego, CA (United States)
1997-12-01
The heightened recent attention given to fuel rod behavior at high burnup has largely been the result of a few reactivity initiation experiments conducted in France and Japan. Regardless of the merits of these tests, their outcome has underscored the need for improved analytical methods for both steady-state performance evaluation and transient safety analysis. At issue is the ability to reliably predict fuel rod behavior over the full range of burnup under all conditions. The overall process can be divided into two essential components: the collection of high-burnup material properties data and the development of predictive computer codes with essential special-effects models. Lessons learned from the Electric Power Research Institute`s (EPRI`s) recent efforts in understanding and properly interpreting the reactivity insertion accident test results indicate that a first-principles approach to fuel rod behavioral modeling within a robust analytical capability is vital for describing the complex interaction between the various phenomena involved. Thus, the heavy reliance on analysis-by-correlation and the excessive use of adjustable parameters is not adequate, especially in extrapolating the analytical results beyond the correlation database. Guided by this philosophy, EPRI has initiated a major fuel rod modeling effort that builds on two existing EPRI codes: the ESCORE code for steady-state analysis and the FREY code for transient analysis. The new combined code, FALCON, is not a mere merging of the one-dimensional ESCORE and the two-dimensional FREY but rather an innovative construct of robust numerics and relevant material models.
Analysis of high burnup spent nuclear fuel by ICP-MS
S. F. Wolf; D. L. Bowers; J. C. Cunnane
2005-01-01
Summary We have used inductively coupled plasma mass spectrometry (ICP-MS) as the primary tool for determining concentrations of a suite of nuclides in samples excised from high-burnup spent nuclear fuel rods taken from light water nuclear reactors. The complete analysis included the determination of 95Mo, 99Tc, 101Ru, 103Rh, 109Ag, 137Cs, 143Nd, 145Nd,148Nd,147Sm, 149Sm, 150Sm, 151Sm, 152Sm, 151Eu, 153Eu, 155Eu, 155Gd, 237Np,
Joe, Kihsoo; Jeon, Young-Shin; Han, Sun-Ho; Lee, Chang-Heon; Ha, Yeong-Keong; Song, Kyuseok
2012-06-01
The content of plutonium isotopes in high burnup pressurized water reactor fuel samples was examined using both alpha spectrometry and mass spectrometry after anion exchange separation. The measured values were compared with results calculated by the ORIGEN-2 code. On average, the ratios (m/c) of the measured values (m) over the calculated values (c) were 1.22±0.16 for (238)Pu, 1.02±0.14 for (239)Pu, 1.08±0.06 for (240)Pu, 1.06±0.16 for (241)Pu, and 1.13±0.08 for (242)Pu. Using the Pu data obtained in this work, correlations were derived between the alpha activity ratios of (238)Pu/((239)Pu+(240)Pu), the alpha specific activities of Pu, and the atom % abundances of the Pu isotopes. Using these correlations, the atom % abundances of the plutonium isotopes in the target samples were calculated. These calculated results agreed within a range from 2 to 8% of the experimentally derived values according to the isotopes of plutonium. PMID:22476019
24. A CORE WORKER DISPLAYS THE CORE BOX AND CORES ...
24. A CORE WORKER DISPLAYS THE CORE BOX AND CORES FOR A BRASS GATE VALVE BODY MADE ON A CORE BOX, CA. 1950. - Stockham Pipe & Fittings Company, 4000 Tenth Avenue North, Birmingham, Jefferson County, AL
Accuracy of Monte Carlo Criticality Calculations During BR2 Operation
Kalcheva, Silva; Koonen, Edgar; Ponsard, Bernard [SCK-CEN (Belgium)
2005-08-15
The Belgian Material Test Reactor BR2 is a strongly heterogeneous high-flux engineering test reactor at SCK-CEN (Centre d'Etude de l'Energie Nucleaire) in Mol with a thermal power of 60 to 100 MW. It deploys highly enriched uranium, water-cooled concentric plate fuel elements, positioned inside a beryllium reflector with a complex hyperboloid arrangement of test holes. The objective of this paper is to validate the MCNP and ORIGEN-S three-dimensional (3-D) model for reactivity predictions of the entire BR2 core during reactor operation. We employ the Monte Carlo code MCNP-4C to evaluate the effective multiplication factor k{sub eff} and 3-D space-dependent specific power distribution. The one-dimensional code ORIGEN-S is used to calculate the isotopic fuel depletion versus burnup and to prepare a database with depleted fuel compositions. The approach taken is to evaluate the 3-D power distribution at each time step and along with the database to evaluate the 3-D isotopic fuel depletion at the next step and to deduce the corresponding shim rod positions of the reactor operation. The capabilities of both codes are fully exploited without constraints on the number of involved isotope depletion chains or an increase of the computational time. The reactor has a complex operation, with important shutdowns between cycles, and its reactivity is strongly influenced by poisons, mainly {sup 3}He and {sup 6}Li from the beryllium reflector, and the burnable absorbers {sup 149}Sm and {sup 10}B in the fresh UAl{sub x} fuel. The computational predictions for the shim rod positions at various restarts are within 0.5 $ ({beta}{sub eff} = 0.0072)
EBSD and TEM Characterization of High Burn-up Mixed Oxide Fuel
Teague, Melissa C.; Gorman, Brian P.; Miller, Brandon D.; King, Jeffrey
2014-01-01
Understanding and studying the irradiation behavior of high burn-up oxide fuel is critical to licensing of future fast breeder reactors. Advancements in experimental techniques and equipment are allowing for new insights into previously irradiated samples. In this work dual column focused ion beam (FIB)/scanning electron microscope (SEM) was utilized to prepared transmission electron microscope samples from mixed oxide fuel with a burn-up of 6.7% FIMA. Utilizing the FIB/SEM for preparation resulted in samples with a dose rate of <0.5 mRem/h compared to approximately 1.1 R/h for a traditionally prepared TEM sample. The TEM analysis showed that the sample taken from the cooler rim region of the fuel pellet had approximately 2.5x higher dislocation density than that of the sample taken from the mid-radius due to the lower irradiation temperature of the rim. The dual column FIB/SEM was additionally used to prepared and serially slice approximately 25 um cubes. High quality electron back scatter diffraction (EBSD) were collected from the face at each step, showing, for the first time, the ability to obtain EBSD data from high activity irradiated fuel.
R and D of Oxide Dispersion Strengthening Steels for High Burn-up Fuel Claddings
Kimura, A.; Cho, H.S.; Lee, J.S.; Kasada, R.; Ukai, S.; Fujiwara, M.
2004-07-01
Research and development of fuel clad materials for high burn-up operation of light water reactor and super critical water reactor (SCPWR) will be shown with focusing on the effort to overcome the requirements of material performance as the fuel clad. Oxide dispersion strengthening (ODS) steels are well known as a high temperature structural material. Recent irradiation experiments indicated that the steels were quite highly resistant to neutron irradiation embrittlement, showing hardening without accompanying loss of ductility. High Cr ODS steels whose chromium concentration was in the range from 15 to 19 wt% showed high resistance to corrosion in supercritical pressurized water (SCPW). As for the susceptibility to hydrogen embrittlement of ODS steels, the critical hydrogen concentration required to hydrogen embrittlement is ranging 10{approx}12 wppm that is approximately one order of magnitude higher value than that of 9Cr reduced activation ferritic (RAF) steel. In the ODS steels, the fraction of helium desorption by bubble migration mechanism was smaller than that in the RAF steel, indicating that the ODS steels are also resistant to helium He bubble-induced embrittlement. Finally, it is demonstrated that the ODS steels are very promising for the fuel clad material for high burn-up operation of water-cooling reactors. (authors)
Determination of high burn-up nuclear fuel elastic properties with acoustic microscopy
NASA Astrophysics Data System (ADS)
Laux, D.; Baron, D.; Despaux, G.; Kellerbauer, A. I.; Kinoshita, M.
2012-01-01
We report the measurement of elastic constants of non-irradiated UO 2, SIMFUEL (simulated spent fuel: UO 2 with several additives which aim to simulate the effect of burnup) and irradiated fuel by focused acoustic microscopy. To qualify the technique a parametric study was conducted by performing measurements on depleted uranium oxide (with various volume fraction of porosity, Oxygen-to-metal ratios, grain sizes) and SIMFUEL and by comparing them with previous works presented in the literature. Our approach was in line with existing literature for each parameter studied. It was shown that the main parameters influencing the elastic moduli are the amount of fission products in solution (related to burnup) and the pore density and shape, the influence of which has been evaluated. The other parameters (irradiation defects, oxygen-to-metal ratio and grain sizes) mainly increase the attenuation of the ultrasonic wave but do not change the wave velocity, which is used in the proposed method to evaluate Young's modulus. Measurements on irradiated fuel (HBRP and N118) were then performed. A global decrease of 25% of the elastic modulus between 0 and 100 GWd/tM was observed. This observation is compared to results obtained with measurements conducted at ITU by Knoop indentation techniques.
Thermal Diffusivity Of Homogeneous SBR MOX Fuel With A Burn-up of 35 MWd/kgHM
Staicu, D.; Pagliosa, G.; Papaioannou, D.; Rondinella, V.V.; Cozzo, C.; Konings, R.; Walker, C.T.; Barker, M.; Weston, R.
2007-07-01
New thermal diffusivity results obtained on SBR MOX fuel with a pellet burn-up of 35 MWd/kgHM are reported. The thermal diffusivity measurements were carried out at three radial positions using a shielded 'laser-flash' device and show that the thermal diffusivity increases from the pellet periphery to the centre. Comparison shows that the thermal conductivity is in the same range than of UO{sub 2} of similar burn-up. The annealing behaviour was characterized in order to identify the degradation due to the out-of pile auto-irradiation. (authors)
NASA Technical Reports Server (NTRS)
Davison, H. W.; Fiero, I. B.
1971-01-01
Fuel volume swelling and clad diametral creep strains were calculated for five fuel pins, clad with either T-111 (Ta-8W-2.4Hf) or PWC-11 (Nb-1Zr-0.1C). The fuel pins were irradiated to burnups between 2.7 and 4.6%. Clad temperatures were between 1750 and 2400 F (1228 and 1589 K). The maximum percentage difference between calculated and experimentally measured values of volumetric fuel swelling is 60%.
CALCULATION OF THE NEUTRON NOISE INDUCED BY SHELL-MODE
Demazière, Christophe
CALCULATION OF THE NEUTRON NOISE INDUCED BY SHELL-MODE FISSION REACTORS CORE-BARREL VIBRATIONS for Publication October 12, 2005 The subject of this paper is the calculation of the in-core neutron noise induced was calculated by representing the vibrations of the core barrel by a model developed earlier to describe control
Accident source terms for light-water nuclear power plants using high-burnup or MOX fuel
Michael Salay; Randall O. Gauntt; Richard Y. Lee; Dana Auburn Powers; Mark Thomas Leonard
2011-01-01
Representative accident source terms patterned after the NUREG-1465 Source Term have been developed for high burnup fuel in BWRs and PWRs and for MOX fuel in a PWR with an ice-condenser containment. These source terms have been derived using nonparametric order statistics to develop distributions for the timing of radionuclide release during four accident phases and for release fractions of
Extension of fuel burnup in light water reactors by using a strict in-out refueling scheme
Yen
1983-01-01
To utilize energy resources as efficiently as possible has become a necessity today. The purpose of this study is to see how this can be done by extending burnup in a light water reactor. Specifically, an in-out refueling scheme might extract the maximum energy from nuclear fuel during its redidence in the reactor. In principle, a reactor loading with minimum
Yudkevich, M. S.
2012-12-15
This paper described the procedure of implementation and the possibilities of the BURNUP program. The purpose of the program is to predict the change in the nuclear composition of the materials of which a reactor is made in the course of its run and compute the radiation characteristics of the materials after their irradiation.
Core restraint contributions to radial expansion reactivity
Moran, T.J.
1986-01-01
Bowing of core assemblies caused by thermal gradients, swelling gradients, and irradiation creep can cause significant changes in reactivity of an LMFBR during startup, overpower and loss-of-flow without scram transients. This paper summarizes calculations of bowing reactivity effects for both a small homogeneous and a small heterogeneous core design. It includes two core restraint concepts for each core design and concentrates on reactivity changes in the critical power-to-flow range of 1.0 to 2.0.
NASA Astrophysics Data System (ADS)
Rusov, V. D.; Pavlovich, V. N.; Vaschenko, V. N.; Tarasov, V. A.; Zelentsova, T. N.; Bolshakov, V. N.; Litvinov, D. A.; Kosenko, S. I.; Byegunova, O. A.
2007-09-01
We give an alternative description of the data produced in the KamLAND experiment. Assuming the existence of a natural nuclear reactor on the boundary of the liquid and solid phases of the Earth's core, a geoantineutrino spectrum is obtained. This assumption is based on the experimental results of V. Anisichkin and his collaborators on the interaction of uranium dioxide and uranium carbide with iron-nickel and silica-alumina melts at high pressure (5-10 GPa) and temperature (1600-2200°C), which led to the proposal of the existence of an actinide shell in the Earth's core. We describe the operating mechanism of this reactor as solitary waves of nuclear burning in 238U and/or 232Th medium, in particular, as neutron fission progressive waves of Feoktistov and/or Teller et al. type. Next, we propose a simplified model for the accumulation and burn-up kinetics in Feoktistov's U-Pu fuel cycle. We also apply this model for numerical simulations of neutron fission wave in a two-phase UO2/Fe medium on the surface of the Earth's solid core. The proposed georeactor model offers a mechanism for the generation of 3He. The 3He/4He distribution in the Earth's interior is calculated, which in turn can be used as a natural quantitative criterion of the georeactor thermal power. Finally, we give a tentative estimation of the geoantineutrino intensity and spectrum on the Earth's surface. For this purpose we use the O'Nions et al. geochemical model of mantle differentiation and crust growth complemented by a nuclear energy source (georeactor with power of 30 TW).
® Powder Cores s Molypermalloy s High Flux s Kool Mµ #12;Since 1949, MAGNETICS, a division of Spang products as industry standards in tape wound cores, powder cores, and ferrite cores. #12;w w w . m a g - i-4 General Powder Core Information CORE SELECTION 2-1 Core Selection Procedure 2-2 Core Selection Example 2
Unknown
2011-08-17
Two types of core/shell nanoparticles (CS-NPs) generation based on laser ablation are developed in this study, namely, double pulse laser ablation and laser ablation in colloidal solutions. In addition to the study of the generation mechanism of CS...
EXOTIC-7: Irradiation of ceramic breeder materials to high lithium burnup
NASA Astrophysics Data System (ADS)
van der Laan, J. G.; Kwast, H.; Stijkel, M.; Conrad, R.; May, R.; Casadio, S.; Roux, N.; Werle, H.; Verrall, R. A.
1996-10-01
The EXOTIC-7 irradiation experiment in the high flux reactor (HFR) has been completed. Its aim has been to investigate the effects of high lithium-burnup on the mechanical stability and tritium release characteristics of candidate ceramic breeder materials, originating from the fusion programmes of CEA, FZK, ENEA, AECL and ECN. The tested ceramic breeder materials were pellets of Li 2ZrO 3, LiAlO 2 and Li 8ZrO 6 and pebbles of Li 4SiO 4 and Li 2ZrO 3, with a variety of characteristics, like grain size and porosity. The test matrix provided the simultaneous irradiation of eight independent capsules with on-line tritium monitoring. Two capsules contained a mixture of Li 4SiO 4 and beryllium pebbles. The experimental design, sample loading and main irradiation parameters are described. Some PIE results and analysis of in-situ tritium release behaviour are presented.
Analyses of Greek Research Reactor with mixed HEU-LEU Be reflected core
Deen, J.R.; Snelgrove, J.L.; Papastergiou, K.
1993-12-31
The fuel-cycle analyses presented in this paper provide specific steps to be taken in the transition from a 36-element water-reflected HEU core to a 33-element LEU equilibrium core with a Be reflector on two faces. The first step will be to install the Be reflector and remove the highest burnup HEU fuel. The smaller Be-reflected core will be refueled with LEU fuel. All analyses were performed using a planar 5-group REBUS3 model benchmarked to VIM Monte Carlo. In addition to fuel cycle results, the control rod worth, reactivity response to increased fuel and water temperature and decreased water density were compared for the transition core and the reference HEU core.
Calculation of the radionuclides in PWR spent fuel samples for SFR experiment planning.
Naegeli, Robert Earl
2004-06-01
This report documents the calculation of radionuclide content in the pressurized water reactor (PWR) spent fuel samples planned for use in the Spent Fuel Ratio (SPR) Experiments at Sandia National Laboratories, Albuquerque, New Mexico (SNL) to aid in experiment planning. The calculation methods using the ORIGEN2 and ORIGEN-ARP computer codes and the input modeling of the planned PWR spent fuel from the H. B. Robinson and the Surry nuclear power plants are discussed. The safety hazards for the calculated nuclide inventories in the spent fuel samples are characterized by the potential airborne dose and by the portion of the nuclear facility hazard category 2 and 3 thresholds that the experiment samples would present. In addition, the gamma ray photon energy source for the nuclide inventories is tabulated to facilitate subsequent calculation of the direct and shielded dose rates expected from the samples. The relative hazards of the high burnup 72 gigawatt-day per metric ton of uranium (GWd/MTU) spent fuel from H. B. Robinson and the medium burnup 36 GWd/MTU spent fuel from Surry are compared against a parametric calculation of various fuel burnups to assess the potential for higher hazard PWR fuel samples.
Viscosity of the earth's core.
NASA Technical Reports Server (NTRS)
Gans, R. F.
1972-01-01
Calculation of the viscosity of the core at the boundary of the inner and outer core. It is assumed that this boundary is a melting transition and the viscosity limits of the Andrade (1934,1952) hypothesis (3.7 to 18.5 cp) are adopted. The corresponding kinematic viscosities are such that the precessional system explored by Malkus (1968) would be unstable. Whether it would be sufficiently unstable to overcome a severely subadiabatic temperature gradient cannot be determined.
REACTOR PHYSICS CALCULATIONS FOR THE MSRE
Nestor; C. W. Jr
1960-01-01
A compilation is presented of results obtained to date from a number of ; reactor physics calculations for the molten salt reactor experiment (MSRE). ; Included are one-dimensional multigroup and two-dimensional twogroup calculations ; of critical mass, flux, and power density distributions; gamma heating in the ; core can, reactor vessel, and core support grid; drain tank criticality; and an
NSDL National Science Digital Library
The Ethics CORE Digital Library, funded by the National Science Foundation, "brings together information on best practices in research, ethics instruction and responding to ethical problems that arise in research and professional life." It's a remarkable site where visitors can make their way through ethics resources for dozens of different professions and activities. The Resources by Discipline area is a great place to start. Here you will find materials related to the biological sciences, business, computer & information science, along with 14 additional disciplines. The Current News area is a great place to learn about the latest updates from the field. Of note, these pieces can easily be used in the classroom or shared with colleagues. The dynamism of the site can be found at the Interact with Ethics CORE area. Active learning exercises can be found here, along with instructional materials and visitors' own lessons learned.
NSDL National Science Digital Library
Birnbaum, Michael H.
This page, created by Michael H. Birnbaum of Fullerton University, uses Bayes' Theorem to calculate the probability of a hypothesis given a datum. An example about cancer is given to help users understand Bayes' Theorem and the calculator. This page is a great representation of conditional probability. Detailed instructions are provided on proper use of the calculator.
Cousot, Patrick
= # = let p = p 1 u p 2 u ? F in hp, piI Å¸ = . should be: The calculational design of the abstract equalityThe Calculational Design of a Generic Abstract Interpreter Corrigendum, February 7, 1999 Patrick 1 , - F (q 2 ), p) in (r 1 , - F (r 2 )) . Section 10.3, page 34 The calculational design
Core Description: The Macromolecule Core (for short)
Hammack, Richard
Core Description: The Macromolecule Core (for short) is a new core consolidating the services and implement the production of macromolecules for multiple end points, both within the Macromol- ecule Core://www.pubinfo.vcu.edu/ mlbiocre/default.htm Biological Macromolecule Core Laboratory Tuesday, March 1, 2011 #12;
Overview of core designs and requirements/criteria for core restraint systems
Sutherland, W.H.
1984-09-01
The requirements and lifetime criteria for the design of a Liquid Metal Fast Breeder Reactor (LMFBR) Core Restraint System are presented. A discussion of the three types of core restraint systems used in LMFBR core design is given. Details of the core restraint system selected for FFTF are presented and the reasons for this selection given. Structural analysis procedures being used to manage the FFTF assembly irradiations are discussed. Efforts that are ongoing to validate the calculational methods and lifetime criteria are presented.
Benchmark data for validating irradiated fuel compositions used in criticality calculations
Bierman, S.R.; Talbert, R.J.
1994-10-01
To establish criticality safety margins utilizing burnup credit in the storage and transport of spent reactor fuels requires a knowledge of the uncertainty in the calculated fuel composition used in making the reactivity assessment. To provide data for validating such calculated burnup fuel compositions, radiochemical assays have been obtained as part of the United States Department of Energy From-Reactor Cask Development Program. Assay results and associated operating histories on the initial three samples analyzed in this effort are presented. The three samples were taken from different axial regions of a Pressurized Water Reactor fuel rod and represent radiation exposures of about 37, 27, and 44 GWd/MTU. The data are presented in a benchmark type format to facilitate identification/referencing and computer code input.
NASA Astrophysics Data System (ADS)
Nur Asiah, A.; Su'ud, Zaki; Ferhat, A.; Sekimoto, H.
2010-06-01
In this paper, conceptual design study of Small Long-life Gas Cooled Fast Reactors with Natural Uranium as Fuel Cycle Input has been performed. In this study Gas Cooled Fast Reactor is slightly modified by employing modified CANDLE burn-up scheme so that it can use Natural Uranium as fuel cycle input. Due to their hard spectrum, GCFR in this study showed very good performance in converting U-238 to plutonium in order to maintain the operation condition requirement of long-life reactors. Due to the limitation of thermal hydraulic aspects, the average power density of the proposed design is selected about 70 W/cc. With such condition we got an optimal design of 325 MWt reactors which can be operated 10 years without refueling and fuel shuffling and just need natural uranium as fuel cycle input. The average discharge burn-up is about 290 GWd/ton HM.
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.
Results of AVR fuel pebble irradiation at increased temperature and burn-up in the HFR Petten
Michael A. Fütterer; Gerard Berg; Alain Marmier; Enrique Toscano; Daniel Freis; Klaas Bakker; Sander de Groot
2008-01-01
The irradiation experiment HFR-EU1bis was performed by the European Commission's Joint Research Centre-Institute for Energy (JRC-IE) in the HFR Petten to test five spherical High Temperature Reactor (HTR) fuel pebbles of former German production with TRISO coated particles for their potential for very high temperature performance and high burn-up. The irradiation started on 9 September 2004 and was terminated on
Masaaki MORI; Masayuki HIJIYA; Seiji SHIROYA; Keiji KANDA
1997-01-01
A conceptual design study was carried out on a super high-burnup mixed-oxide (MOX) fuel assembly (SHB FA) for pressurized water reactors (PWRs) using transuranium (TRU). This study aims to avoid the surplus plutonium (Pu) accumulation and to reduce the accumulation of long-lived radioactive minor actinides (MAS) by utilizing the currently existing PWRs under the condition that the Japanese program to
Einziger, Robert E.; Beyer, Carl E.
2007-08-01
Current risk assessments of spent fuel in storage and transportation casks use the properties of LWR fuel below 45 GWd/MTU. Fuel is being driven to higher burnups that may influence the source term in cask accidents. To achieve these burnups the manufacturers are introducing new assembly designs and cladding alloys. As a result, at the higher burnups (? 50 GWd/MTU) some of the characteristics of the fuel pellets, cladding, and assembly design used in the safety analysis have changed. The fuel pellet has developed a fine grained, Pu rich rim zone on its exterior surface. The source term may increase by 1 – 3 orders of magnitude depending on the fracture characteristics of the rim. The cladding may acquire hydrogen contents up to 700 wppm during the increased exposure. Embrittlement with subsequent lose of ductility may occur, especially if there is hydride reorientation. As a result, there may be a greater propensity for fracture of the rods upon impact with subsequent release of fuel particulate and gas. Significantly improved source terms can be developed if additional data on fuel rim fracture as a function of impact energy, the dependence of cladding ductility for Zircaloy and the newer cladding alloys as a function of hydride reorientation, and release characteristics for fractured rods were obtained. CRUD spallation characteristics only make a significant contribution to the source term if the rods do not fracture in the accident or if a fire only accident occurs.
SEM Characterization of the High Burn-up Microstructure of U-7Mo Alloy
Dennis D. Keiser, Jr.; Jan-Fong Jue; Jian Gan; Brandon Miller; Adam Robinson; Pavel Medvedev; James Madden; Dan Wachs; M. Teague
2014-04-01
During irradiation, the microstructure of U-7Mo evolves until at a fission density near 5x1021 f/cm3 a high-burnup microstructure exists that is very different than what was observed at lower fission densities. This microstructure is dominated by randomly distributed, relatively large, homogeneous fission gas bubbles. The bubble superlattice has collapsed in many microstructural regions, and the fuel grain sizes, in many areas, become sub-micron in diameter with both amorphous fuel and crystalline fuel present. Solid fission product precipitates can be found inside the fission gas bubbles. To generate more information about the characteristics of the high-fission density microstructure, three samples irradiated in the RERTR-7 experiment have been characterized using a scanning electron microscope equipped with a focused ion beam. The FIB was used to generate samples for SEM imaging and to perform 3D reconstruction of the microstructure, which can be used to look for evidence of possible fission gas bubble interlinkage.
In-Situ Safeguards Verification of Low Burn-up Pressurized Water Reactor Spent Fuel Assemblies
Ham, Y S; Sitaraman, S; Park, I; Kim, J; Ahn, G
2008-04-16
A novel in-situ gross defect verification method for light water reactor spent fuel assemblies was developed and investigated by a Monte Carlo study. This particular method is particularly effective for old pressurized water reactor spent fuel assemblies that have natural uranium in their upper fuel zones. Currently there is no method or instrument that does verification of this type of spent fuel assemblies without moving the spent fuel assemblies from their storage positions. The proposed method uses a tiny neutron detector and a detector guiding system to collect neutron signals inside PWR spent fuel assemblies through guide tubes present in PWR assemblies. The data obtained in such a manner are used for gross defect verification of spent fuel assemblies. The method uses 'calibration curves' which show the expected neutron counts inside one of the guide tubes of spent fuel assemblies as a function of fuel burn-up. By examining the measured data in the 'calibration curves', the consistency of the operator's declaration is verified.
Hertzian indentation of advanced lmfbr fuels with simulated burn-up
NASA Astrophysics Data System (ADS)
Matzke, Hj.; Inoue, T.
1982-10-01
Hertzian indentation fracture of advanced fast breeder reactor fuel materials [mixed carbonitrides, (U 0.8, Pu 0.2)C 0.8N 0.2, and nitrides (U 0.8Pu 0.2)N was evaluated to yield the fracture surface energy, ?, and the fracture toughness, K Ic. Both technological grade fuels and fuels with added fission products to chemically simulate burn-up values of 3 and 10 at.% were used. As in previous self-diffusion studies on the same materials, identical behavior (identical critical loads, P c for crack formation) was observed for 3 and 10% b.u. Simulated M(C, N), whereas the 10% b.u. Simulated MN showed a cracking behavior identical with that of the undoped MN. In contrast, the 3 at.% b.u. Simulated MN showed lower P c values. This is compatible with differences in fission product solubilities in these materials. The effect of fission products on ? was < 20% whereas ? increased from (U, Pu)(C, N) to (U, Pu)N by up to 80 to 90%, depending on content in fission products.
High Burn-Up Spent Nuclear Fuel Vibration Integrity Study 15134
Wang, Jy-An John; Wang, Hong; Jiang, Hao; Bevard, Bruce Balkcom; Howard, Rob L; Scaglione, John M
2015-01-01
The Oak Ridge National Laboratory (ORNL) has developed the cyclic integrated reversible-bending fatigue tester (CIRFT) approach to successfully demonstrate the controllable fatigue fracture on high burnup (HBU) spent nuclear fuel (SNF) in a normal vibration mode. CIRFT enables examination of the underlying mechanisms of SNF system dynamic performance. Due to the inhomogeneous composite structure of the SNF system, the detailed mechanisms of the pellet-pellet and pellet-clad interactions and the stress concentration effects at the pellet-pellet interface cannot be readily obtained from a CIRFT system measurement. Therefore, finite element analyses (FEAs) are used to translate the global moment-curvature measurement into local stress-strain profiles for further investigation. The major findings of CIRFT on the HBU SNF are as follows: SNF system interface bonding plays an important role in SNF vibration performance. Fuel structure contributes to SNF system stiffness. There are significant variations in stress and curvature of SNF systems during vibration cycles resulting from segment pellets and clad interactions. SNF failure initiates at the pellet-pellet interface region and appears to be spontaneous.
Calculating Flows In Turbomachine Channels
NASA Technical Reports Server (NTRS)
Schumann, Lawrence F.
1989-01-01
Noniterative integral-entrainment method yields good approximations. Method of approximate calculation of flow in channel of turbomachine based on interaction of viscous flow in boundary layers with inviscid flow in core layer. Faster and more robust than other approximate methods of same type. Suitable for use in preliminary calculations for design and for off-design operation of turbomachinery. Flows in conical diffuser channels represented by two-dimensional boundary-layer and one-dimensional core flows described by equations of new method.
Gas core reactors for actinide transmutation and breeder applications
NASA Technical Reports Server (NTRS)
Clement, J. D.; Rust, J. H.
1978-01-01
This work consists of design power plant studies for four types of reactor systems: uranium plasma core breeder, uranium plasma core actinide transmuter, UF6 breeder and UF6 actinide transmuter. The plasma core systems can be coupled to MHD generators to obtain high efficiency electrical power generation. A 1074 MWt UF6 breeder reactor was designed with a breeding ratio of 1.002 to guard against diversion of fuel. Using molten salt technology and a superheated steam cycle, an efficiency of 39.2% was obtained for the plant and the U233 inventory in the core and heat exchangers was limited to 105 Kg. It was found that the UF6 reactor can produce high fluxes (10 to the 14th power n/sq cm-sec) necessary for efficient burnup of actinide. However, the buildup of fissile isotopes posed severe heat transfer problems. Therefore, the flux in the actinide region must be decreased with time. Consequently, only beginning-of-life conditions were considered for the power plant design. A 577 MWt UF6 actinide transmutation reactor power plant was designed to operate with 39.3% efficiency and 102 Kg of U233 in the core and heat exchanger for beginning-of-life conditions.
Li, Zhian; Rahimi, Meraj; Tang, David; Aissa, Mourad; Flaganan, Michelle [U.S. Nuclear Regulatory Commission - NRC, Washington, DC 20555-0001 (United States)] [U.S. Nuclear Regulatory Commission - NRC, Washington, DC 20555-0001 (United States); Wagner, John C. [Oak Ridge National Laboratory (United States)] [Oak Ridge National Laboratory (United States)
2013-07-01
The recent experiments conducted by Argonne National Laboratory on high burnup fuel cladding material property show that the ductile to brittle transition temperature of high burnup fuel cladding is dependent on: (1) cladding material, (2) irradiation conditions, and (3) drying-storage histories (stress at maximum temperature) [1]. The experiment results also show that the ductile to brittle temperature increases as the fuel burnup increases. These results indicate that the current knowledge in cladding material property is insufficient to determine the structural performance of the cladding of high burnup fuel after it has been stored in a dry cask storage system for some time. The uncertainties in material property and the elevated ductile to brittle transition temperature impose a challenge to the storage cask and transportation packaging designs because the cask designs may not be able to rely on the structural integrity of the fuel assembly for control of fissile material, radiation source, and decay heat source distributions. The fuel may reconfigure during further storage and/or the subsequent transportation conditions. In addition, the fraction of radioactive materials available for release from spent fuel under normal condition of storage and transport may also change. The spent fuel storage and/or transportation packaging vendors, spent fuel shippers, and the regulator may need to consider this possible fuel reconfiguration and its impact on the packages' ability to meet the safety requirements of Part 72 and Part 71 of Title 10 of the Code of Federal Regulations. The United States Nuclear Regulatory Commission (NRC) is working with the scientists at Oak Ridge National Laboratory (ORNL) to assess the impact of fuel reconfiguration on the safety of the dry storage systems and transportation packages. The NRC Division of Spent Fuel Storage and Transportation has formed a task force to work on the safety and regulatory concerns in relevance to high burnup fuel storage and transportation. This paper discusses the staff's preliminary considerations on the safety implication of fuel reconfiguration with respect to nuclear safety (subcriticality control), radiation shielding, containment, the performance of the thermal functions of the packages, and the retrievability of the contents from regulatory perspective. (authors)
Thermal property change of MOX and UO2 irradiated up to high burnup of 74 GWd/t
NASA Astrophysics Data System (ADS)
Nakae, Nobuo; Akiyama, Hidetoshi; Miura, Hiromichi; Baba, Toshikazu; Kamimura, Katsuichiro; Kurematsu, Shigeru; Kosaka, Yuji; Yoshino, Aya; Kitagawa, Takaaki
2013-09-01
Thermal property is important because it controls fuel behavior under irradiation. The thermal property change at high burnup of more than 70 GWd/t is examined. Two kinds of MOX fuel rods, which were fabricated by MIMAS and SBR methods, and one referenced UO2 fuel rod were used in the experiment. These rods were taken from the pre-irradiated rods (IFA 609/626, of which irradiation test were carried out by Japanese PWR group) and re-fabricated and re-irradiated in HBWR as IFA 702 by JNES. The specification of fuel corresponds to that of 17 × 17 PWR type fuel and the axially averaged linear heat rates (LHR) of MOX rods are 25 kW/m (BOL of IFA 702) and 20 kW/m (EOL of IFA 702). The axial peak burnups achieved are about 74 GWd/t for both of MOX and UO2. Centerline temperature and plenum gas pressure were measured in situ during irradiation. The measured centerline temperature is plotted against LHR at the position where thermocouples are fixed. The slopes of MOX are corresponded to each other, but that of UO2 is higher than those of MOX. This implies that the thermal conductivity of MOX is higher than that of UO2 at high burnup under the condition that the pellet-cladding gap is closed during irradiation. Gap closure is confirmed by the metallography of the postirradiation examinations. It is understood that thermal conductivity of MOX is lower than that of UO2 before irradiation since phonon scattering with plutonium in MOX becomes remarkable. A phonon scattering with plutonium decreases in MOX when burnup proceeds. Thus, thermal conductivity of MOX becomes close to that of UO2. A reverse phenomenon is observed at high burnup region. The phonon scattering with fission products such as Nd and Zr causes a degradation of thermal conductivity of burnt fuel. It might be speculated that this scattering effect causes the phenomenon and the mechanism is discussed here.
NSDL National Science Digital Library
This website created by Erez Kaplan "deals mainly with the mechanical calculating machines from a collector's point of view." Included here is an historical review of calculating machines, along with Kaplan's attempt to classify the machines, a collection of old advertisements for the machines, and a brief history of calculating. The latest feature is a Java applet that lets you operate an 1885 Felt adding machine to give you a sense of the way it was used. The photos and descriptions provide insight on other gadgets such as the Pocket Cash Registers used by "the sophisticated man or woman of 1900 who had everything." The Reference section provides some resources for further reading, including numerous other personal calculator collectors sites and museums.
NASA Technical Reports Server (NTRS)
Voorhies, C. V.
1999-01-01
The idea that geomagnetic westward drift indicates convective leveling of the planetary momentum gradient within Earth's core is pursued in search of a differentially rotating mean state, upon which various oscillations and secular effects might be superimposed. The desired state conforms to roughly spherical boundary conditions, minimizes dissipative interference with convective cooling in the bulk of the core, yet may aide core cooling by depositing heat in the uppermost core and lower mantle. The variational calculus of stationary dissipation applied to a spherical vortex within the core yields an interesting differential rotation profile akin to spherical Couette flow bounded by thin Hartmann layers. Four boundary conditions are required. To concentrate shear induced dissipation near the core-mantle boundary, these are taken to be: (i) no-slip at the core-mantle interface; (ii) geomagnetically estimated bulk westward flow at the base of the core-mantle boundary layer; (iii) no-slip at the inner-outer core interface; and, to describe magnetic locking of the inner core to the deep outer core, (iv) hydrodynamically stress-free at the inner-outer core boundary. By boldly assuming the axial core angular momentum anomaly to be zero, the super-rotation of the inner core is calculated to be at most 1.5 degrees per year.
NASA Technical Reports Server (NTRS)
Voorhies, Coerte V.
1998-01-01
The idea that geomagnetic westward drift indicates convective leveling of the planetary momentum gradient within Earth's core is pursued in search of a differentially rotating mean state, upon which various oscillations and secular effects might be superimposed. The desired state conforms to roughly spherical boundary conditions, minimizes dissipative interference with convective cooling in the bulk of the core, yet may aid core cooling by depositing heat in the uppermost core and lower mantle. The variational calculus of stationary dissipation applied to a spherical vortex within the core yields an interesting differential rotation profile, akin to spherical Couette flow bounded by thin Hartmann layers. Four boundary conditions are required. To concentrate shear induced dissipation near the core-mantle boundary, these are taken to be: (i) no-slip at the core-mantle interface; (ii) geomagnetically estimated bulk westward flow at the base of the core-mantle boundary layer; (iii) no-slip at the inner-outer core interface; and, to describe magnetic locking of the inner core to the deep outer core; (iv) hydrodynamically stress-free at the inner-outer core boundary. By boldly assuming the axial core angular momentum anomaly to be zero, the super-rotation of the inner core relative to the mantle is calculated to be at most 1.5 deg./yr.
Start-up fuel and power flattening of sodium-cooled candle core
Takaki, Naoyuki; Sagawa, Yu; Umino, Akitake [Department of Nuclear Safety Engineering, Tokyo City University 1-28-1 Tamazutsumi, Setagaya, Tokyo 158-8557 (Japan); Sekimoto, Hiroshi [University of California, Berkeley, CA 94720 (United States)
2013-07-01
The hard neutron spectrum and unique power shape of CANDLE enable its distinctive performances such as achieving high burnup more than 30% and exempting necessity of both enrichment and reprocessing. On the other hand, they also cause several challenging problems. One is how the initial fuel can be prepared to start up the first CANDLE reactor because the equilibrium fuel composition that enables stable CANDLE burning is complex both in axial and radial directions. Another prominent problem is high radial power peaking factor that worsens averaged burnup, namely resource utilization factor in once-through mode and shorten the life time of structure materials. The purposes of this study are to solve these two problems. Several ideas for core configurations and startup fuel using single enrichment uranium and iron as a substitute of fission products are studied. As a result, it is found that low enriched uranium is applicable to ignite the core but all concepts examined here exceeded heat limits. Adjustment in enrichment and height of active and burnt zone is opened for future work. Sodium duct assemblies and thorium fuel assemblies loaded in the center region are studied as measures to reduce radial power peaking factor. Replacing 37 fuels by thorium fuel assemblies in the zeroth to third row provides well-balanced performance with flattened radial power distribution. The CANDLE core loaded with natural uranium in the outer and thorium in the center region achieved 35.6% of averaged burnup and 7.0 years of cladding life time owing to mitigated local fast neutron irradiation at the center. Using thorium with natural or depleted uranium in CANDLE reactor is also beneficial to diversifying fission resource and extending available term of fission energy without expansion of needs for enrichment and reprocessing.
NSDL National Science Digital Library
Friedman, S. Morgan.
This simple inflation calculator uses the Consumer Price Index to adjust any given amount of money, from 1800 to 1998. Creator S. Morgan Friedman uses data from the Historical Statistics of the United States for statistics predating 1975 and the annual Statistics Abstracts of the United States for data from 1975 to 1998. Links to other online inflation information are also included.
Lee, C. H.; Zhong, Z.; Taiwo, T.A.; Yang, W.S.; Khalil, H.S.; Smith, M.A.; Nuclear Engineering Division
2006-10-13
Enhancements have been made to the REBUS-3/DIF3D code suite to facilitate its use for the design and analysis of prismatic Very High Temperature Reactors (VHTRs). A new cross section structure, using table-lookup, has been incorporated to account for cross section changes with burnup and fuel and moderator temperatures. For representing these cross section dependencies, three new modules have been developed using FORTRAN 90/95 object-oriented data structures and implemented within the REBUS-3 code system. These modules provide a cross section storage procedure, construct microscopic cross section data for all isotopes, and contain a single block of banded scattering data for efficient data management. Fission products other than I, Xe, Pm, and Sm, can be merged into a single lumped fission product to save storage space, memory, and computing time without sacrificing the REBUS-3 solution accuracy. A simple thermal-hydraulic (thermal-fluid) feedback model has been developed for prismatic VHTR cores and implemented in REBUS-3 for temperature feedback calculations. Axial conduction was neglected in the formulation because of its small magnitude compared to radial (planar) conduction. With the simple model, the average fuel and graphite temperatures are accurately estimated compared to reference STAR-CD results. The feedback module is currently operational for the non-equilibrium fuel cycle analysis option of REBUS-3. Future work should include the extension of this capability to the equilibrium cycle option of the code and additional verification of the feedback module. For the simulation of control rods in VHTR cores, macroscopic cross section deviations (deltas) have been defined to account for the effect of control rod insertion. The REBUS-3 code has been modified to use the appropriately revised cross sections when control rods are inserted in a calculation node. In order to represent asymmetric core blocks (e.g., fuel blocks or reflector blocks containing asymmetric absorber rods), surface-dependent discontinuity factors based on nodal equivalence theory have been introduced into the nodal diffusion theory option of the DIF3D code (DIF3D-nodal) to improve modeling accuracy. Additionally, the discontinuity factors based on the Simplified Equivalence Theory (SET) have been incorporated as an alternative and may be employed for both the DIF3D-nodal and DIF3D-VARIANT (nodal transport) solution options. Two- and three-dimensional core calculations have been performed using the routines developed and modified in this work, along with cross sections generated from single fuel block and one-dimensional or two-dimensional fuel-reflector model. Generally, REBUS-3/DIF3D results for the core multiplication factor and power distribution are found to be in good agreement with reference results (generated with MCNP continuous energy calculations) particularly when discontinuity factors are applied. The DIF3D-VARIANT option was found to provide a more accurate solution in its diffusion approximation than the DIF3D-nodal option. Control rod worths can be estimated with acceptably small errors compared to MCNP results. However, estimation of the core power tilt needs to be improved by introducing the surface-dependent discontinuity factor capability in DIF3D-VARIANT.
Aspects of zircaloy core materials for advanced LWRs
Adamson, R.B.; Cheng, B.C.; Tucker, R.P.
1987-01-01
Zircaloy has been successfully used as the primary light water reactor (LWR) core structural material since its introduction in the early days of the US naval nuclear program. Its unique combination of low neutron absorption cross section, fabricability, mechanical strength, and corrosion resistance in high-temperature water has resulted in remarkable reliability of operation of pressurized and boiling water reactor (PWR, BWR) fuel through the years. Currently, BWRs use Zircaloy-2 for fuel cladding and Zircaloy-4 for channels and spacers, while PWRs use Zircaloy-4 for fuel cladding, spacer grids, and control rod guide tubes. As required fluences continue to increase and as a more statistically significant number of components reach ultra-high burnups, however, Zircaloy as a material is liable to be pushed to its operating limits. This paper discusses those areas to which fuel vendors are giving attention; e.g., microstructure, corrosion, irradiation growth and creep, and mechanical properties.
G. A. Berna; M. P. Bohn; W. N. Rausch; R. E. Williford; D. D. Lanning
1981-01-01
FRAPCON-2 is a FORTRAN IV computer code that calculates the steady state response of light water reactor fuel rods during long-term burnup. The code calculates the temperature, pressure, deformation, and failure histories of a fuel rod as functions of time-dependent fuel rod power and coolant boundary conditions. The phenomena modeled by the code include: (a) heat conduction through the fuel
NSDL National Science Digital Library
Goodman, Jonathan
This site contains many chemistry applets created by Jonathan Goodman and his group at Cambridge University. An example of an applet available is the Molecular Weight Calculation; whereby entering in a molecular formula, users are able to discover the HRMS weight, the molecular weight, the element percents, and the Molecular Ion Isotope Pattern. Interactive graphs are also available to assist chemistry students with concepts such as boiling points, pressure, and Consecutive First Step Reversible Reactions. Educators and students will also find many three dimensional depictions of the molecules including fused rings, aromatic rings, and Fullerenes.
Benker, D.E.; Bigelow, J.E.; Bond, W.D.; Chattin, F.R.; King, L.J.; Kitts, F.G.; Ross, R.G.; Stacy, R.G.
1986-10-01
A batch of high-burnup fuel from the Fast Flux Test Facility (FFTF) was processed in the Solvent Extraction Test Facility (SETF) during Campaign 9. The fuel had a burnup of {similar_to}0 MWd/kg and a cooling time of {similar_to} year. Two runs were made with this fuel; in the first, the solvent contained 30% tri-n-butyl phosphate (TBP) and partitioning of the uranium and plutonium was effected by reducing the plutonium with hydroxylamine nitrate (HAN); in the second, the solvent contained 10% TBP and a low operating temperature was used in an attempt to partition without reducing the plutonium valence. The plutonium reoxidation problem, which was present in previous runs that used HAN, may have been solved by lowering the temperature and acidity in the partition contactor. An automatic control system was used to maintain high loadings of heavy metals in the coextraction-coscrub contactor in order to increase its efficiency while maintaining low losses of uranium and plutonium to the aqueous raffinate. An in-line photometer system was used to measure the plutonium concentration in an intermediate extraction stage; and based on this data, a computer algorithm determined the appropriate adjustments in the addition rate of the extractant. The control system was successfully demonstrated in a preliminary run with purified uranium. However, a variety of equipment and start up problems prevented an extended demonstration from being accomplished during the runs with the FFTF fuel.
Kramer, K J; Latkowski, J F; Abbott, R P; Boyd, J K; Powers, J J; Seifried, J E
2008-10-24
Lawrence Livermore National Laboratory is currently developing a hybrid fusion-fission nuclear energy system, called LIFE, to generate power and burn nuclear waste. We utilize inertial confinement fusion to drive a subcritical fission blanket surrounding the fusion chamber. It is composed of TRISO-based fuel cooled by the molten salt flibe. Low-yield (37.5 MJ) targets and a repetition rate of 13.3 Hz produce a 500 MW fusion source that is coupled to the subcritical blanket, which provides an additional gain of 4-8, depending on the fuel. In the present work, we describe the neutron transport and nuclear burnup analysis. We utilize standard analysis tools including, the Monte Carlo N-Particle (MCNP) transport code, ORIGEN2 and Monteburns to perform the nuclear design. These analyses focus primarily on a fuel composed of depleted uranium not requiring chemical reprocessing or enrichment. However, other fuels such as weapons grade plutonium and highly-enriched uranium are also under consideration. In addition, we have developed a methodology using {sup 6}Li as a burnable poison to replace the tritium burned in the fusion targets and to maintain constant power over the lifetime of the engine. The results from depleted uranium analyses suggest up to 99% burnup of actinides is attainable while maintaining full power at 2GW for more than five decades.
Core excitation effects in the breakup of halo nuclei
Moro, A. M.; Diego, R. de; Lay, J. A.; Crespo, R.; Johnson, R. C.; Arias, J. M.; Gomez-Camacho, J. [Departamento de FAMN, Universidad de Sevilla, Apartado 1065, E-41080 Sevilla (Spain); Centro de Fisica Nuclear, Universidade de Lisboa, Av. Prof. Gama Pinto 2, 1649-003 Lisboa (Portugal) and Departamento de Fisica, Instituto Superior Tecnico, Universidade Tecnica de Lisboa, Av. Prof. Cavaco Silva, Taguspark (Portugal); Physics Department, University of Surrey, Guildford Surrey, GU2 7XH (United Kingdom); Departamento de FAMN, Universidad de Sevilla, Apartado 1065, E-41080 Sevilla (Spain); Departamento de FAMN, Universidad de Sevilla, Apartado 1065, E-41080 Sevilla (Spain) and Centro Nacional de Aceleradores, Universidad de Sevilla/Junta de Andalucia, E-41092 Sevilla (Spain)
2012-10-20
The role of core excitation in the structure and dynamics of two-body halo nuclei is investigated. We present calculations for the resonant breakup of {sup 11}Be on protons at an incident energy of 63.7 MeV/nucleon, where core excitation effects were shown to be important. To describe the reaction, we use a recently developed extension of the DWBA formalism which incorporates these core excitation effects within the no-recoil approximation. The validity of the no-recoil approximation is also examined by comparing with DWBA calculations which take into account core recoil. In addition, calculations with two different continuum representations are presented and compared.
NSDL National Science Digital Library
Stark, Philip B.
This tool lets you calculate the probability that a random variable X is in a specified range, for a variety of probability distributions for X: the normal distribution, the binomial distribution with parameters n and p, the chi-square distribution, the exponential distribution, the geometric distribution, the hypergeometric distribution, the negative binomial distribution, the Poisson distribution, and Student's t-distribution. The first choice box lets you select a probability distribution. Depending on the distribution you select, text areas will appear for you to enter the values of the parameters of the distribution. Parameters that are probabilities (e.g., the chance of success in each trial for a binomial distribution) can be entered either as decimal numbers between 0 and 1, or as percentages. If you enter a probability as a percentage, be sure to include the percent sign (%) after the number.
Bansal, Abhishek; Hiremath, Anand; Aluckal, Eby
2015-01-01
Start with end in mind' is a popular cliché in orthodontics. This aptly applies to the therapeutic occlusion the orthodontist strives to achieve. Predicting the post treatment occlusion is an essential component of treatment planning. When no extractions or symmetric extractions are done predicting the final occlusion is somewhat easy. Prediction is challenging when we do unconventional and/or asymmetric extractions. To aid this decision Kesling proposed the 'Kesling Setup'. Though it serves the purpose acceptably; it is time, energy and money consuming. We have developed a model which can help us visualize the final occlusion in matter of seconds. Although this model is primarily intended for orthodontic postgraduate teaching, it can be of considerable use even to a seasoned orthodontist. The regular use of "Orthodontic Calculator" in our department is a testimony to its usefulness. PMID:25738101
Rabacus: Analytic Cosmological Radiative Transfer Calculations
NASA Astrophysics Data System (ADS)
Altay, Gabriel
2015-02-01
Rabacus performs analytic radiative transfer calculations in simple geometries relevant to cosmology and astrophysics; it also contains tools to calculate cosmological quantities such as the power spectrum and mass function. With core routines written in Fortran 90 and then wrapped in Python, the execution speed is thousands of times faster than equivalent routines written in pure Python.
Academic Rigor: The Core of the Core
ERIC Educational Resources Information Center
Brunner, Judy
2013-01-01
Some educators see the Common Core State Standards as reason for stress, most recognize the positive possibilities associated with them and are willing to make the professional commitment to implementing them so that academic rigor for all students will increase. But business leaders, parents, and the authors of the Common Core are not the only…
NASA Astrophysics Data System (ADS)
Kato, Chinami; Delfan Azari, Milad; Yamada, Shoichi; Takahashi, Koh; Umeda, Hideyuki; Yoshida, Takashi; Ishidoshiro, Koji
2015-08-01
Aiming to distinguish two types of progenitors of core-collapse supernovae, i.e., one with a core composed mainly of oxygen and neon (abbreviated as ONe core) and the other with an iron core (or Fe core), we calculated the luminosities and spectra of neutrinos emitted from these cores prior to gravitational collapse, taking neutrino oscillation into account. We found that the total energies emitted as {\\bar{? }}{{e}} from the ONe core are ? {10}46 {erg}, which is much smaller than ? {10}47 {erg} for Fe cores. The average energy, on the other hand, is twice as large for the ONe core as those for the Fe cores. The neutrinos produced by the plasmon decays in the ONe core are more numerous than those from the electron–positron annihilation in both cores, but they have much lower average energies ? 1 {MeV}. Although it is difficult to detect the pre-supernova neutrinos from the ONe core even if it is located within 200 pc from Earth, we expect ?9–43 and ?7–61 events for Fe cores at KamLAND and Super-Kamiokande, respectively, depending on the progenitor mass and neutrino-mass hierarchy. These numbers might be increased by an order of magnitude if we envisage next-generation detectors such as JUNO. We will hence be able to distinguish the two types of progenitors by the detection or nondetection of the pre-supernova neutrinos if they are close enough (? 1 {kpc}).
Pressure vessel calculations for VVER-440 reactors.
Hordósy, G; Hegyi, Gy; Keresztúri, A; Maráczy, Cs; Temesvári, E; Vértes, P; Zsolnay, E
2005-01-01
For the determination of the fast neutron load of the reactor pressure vessel a mixed calculational procedure was developed. The procedure was applied to the Unit II of Paks NPP, Hungary. The neutron source on the outer surfaces of the reactor was determined by a core design code, and the neutron transport calculations outside the core were performed by the Monte Carlo code MCNP. The reaction rate in the activation detectors at surveillance positions and at the cavity were calculated and compared with measurements. In most cases, fairly good agreement was found. PMID:16381691
Trends in adsorbate induced core level shifts
NASA Astrophysics Data System (ADS)
Nilsson, Viktor; Van den Bossche, Maxime; Hellman, Anders; Grönbeck, Henrik
2015-10-01
Photoelectron core level spectroscopy is commonly used to monitor atomic and molecular adsorption on metal surfaces. As changes in the electron binding energies are convoluted measures with different origins, calculations are often used to facilitate the decoding of experimental signatures. The interpretation could in this sense benefit from knowledge on trends in surface core level shifts for different metals and adsorbates. Here, density functional theory calculations have been used to systematically evaluate core level shifts for (111) and (100) surfaces of 3d, 4d, and 5d transition metals upon CO, H, O and S adsorption. The results reveal trends and several non-intuitive cases. Moreover, the difficulties correlating core level shifts with charging and d-band shifts are underlined.
Efficient Core Maintenance in Large Dynamic Graphs
Li, Rong-Hua
2012-01-01
The $k$-core decomposition in a graph is a fundamental problem for social network analysis. The problem of $k$-core decomposition is to calculate the core number for every node in a graph. Previous studies mainly focus on $k$-core decomposition in a static graph. There exists a linear time algorithm for $k$-core decomposition in a static graph. However, in many real-world applications such as online social networks and the Internet, the graph typically evolves over time. Under such applications, a key issue is to maintain the core number of nodes given the graph changes over time. A simple implementation is to perform the linear time algorithm to recompute the core number for every node after the graph is updated. Such simple implementation is expensive when the graph is very large. In this paper, we propose a new efficient algorithm to maintain the core number for every node in a dynamic graph. Our main result is that only certain nodes need to update their core number given the graph is changed by inserting...
hp calculators HP 50g Date calculations
Vetter, Frederick J.
hp calculators HP 50g Date calculations The TIME menu Adding days to a date Days between dates Practice solving date problems #12;hp calculators HP 50g Date calculations hp calculators - 2 - HP 50g Date on the first level of the stack, prior to the execution of the DATE+ function. #12;hp calculators HP 50g Date
Tsai, Hanchung; Liu, Yung Y.; Wang, Da-Yung; Kramer, J.M.
1991-07-01
A series of ex-reactor heating tests on low burnup U-26wt.%Pu-10wt.%Zr metallic fuel for the PRISM reactor was conducted to evaluate fuel/cladding metallurgical interaction and its effect on cladding integrity at elevated temperatures. The reaction between the fuel and cladding caused liquid-phase formation and dissolution of the inner surface of the cladding. The rate of cladding penetration was below the existing design correlation, which provides a conservative margin to cladding failure. In a test which enveloped a wide range of postulated reactor transient events, a substantial temporal cladding integrity margin was demonstrated for an intact, whole fuel pin. The cause of the eventual pin breach was reaction-induced cladding thinning combined with fission-gas pressure loading. The behavior of the breached pin was benign. 7 refs., 7 figs., 1 tab.
MOX capsule post-irradiation examination. Volume 2: Test plan for 30-GWd/MT burnup fuel
Morris, R.N.
1997-12-01
This test plan is a Level-2 document as defined in the Fissile Materials Disposition Program Light-Water Reactor Mixed-Oxide Fuel Irradiation Test Project Plan. The planned post-irradiation examination (PIE) work to be performed on the mixed uranium and plutonium oxide fuel capsules that have received burnups of approximately 30 GWd/MT is described. The major emphasis of this PIE task will be material interactions, particularly indications of gallium transport and interactions. This PIE will include gamma scanning, ceramography, metallography, pellet radial gallium analysis, and clad gallium analysis. A preliminary PIE report will be generated before all the work is completed so that the progress of the fuel irradiation may be known in a timely manner.
Core Structure of Intracluster Gas: Effects of Radiative Cooling on Core Sizes
Takuya Akahori; Kuniaki Masai
2006-04-10
We investigate the core structure of radiatively cooling intracluster gas, using a hydrodynamics code. We calculate evolution of model clusters of the initial core radii 160--300 kpc until the initial central cooling time, and analyze the resultant clusters using the double beta-model as done by observational studies. It is found that the core-size distribution thus obtained shows two peaks \\sim 30--100 kpc and \\sim 100--300 kpc and marginally can reproduce the observed distribution which exhibits two distinct peaks around \\sim 50 kpc and \\sim 200 kpc. This result may suggest radiative-cooling origin for small cores, while cooling is yet insignificant in the clusters of large cores. It should be noted that the small core peak is reproduced by clusters that are still keeping quasi-hydrostatic balance before the initial central cooling time has elapsed.
Evolution of the core physics concept for the Canadian supercritical water reactor
Pencer, J.; Colton, A.; Wang, X.; Gaudet, M.; Hamilton, H.; Yetisir, M.
2013-07-01
The supercritical water cooled reactor (SCWR) is one of the advanced reactor concepts chosen by the GEN-IV International Forum (GIF) for research and development efforts. Canada's contribution is the Canadian SCWR, a heavy water moderated, pressure tube supercritical light water cooled reactor. Recent developments in the SCWR lattice and core concepts, primarily the introduction of a large central flow tube filled with coolant combined with a two-ring fuel assembly, have enabled significant improvements compared to earlier concepts. These improvements include a reduction in coolant void reactivity (CVR) by more than 10 mk, and an almost 40% increase in fuel exit burnup, which is achieved via balanced power distribution between the fuel pins in the fuel assembly. In this paper the evolution of the physics concept is reviewed, and the present lattice and core physics concepts are presented.
Merk, B. [Helmholtz-Zentrum Dresden-Rossendorf, Institut fuer Sicherheitsforschung, Postfach 51 01 19, 01314 Dresden (Germany); Weiss, F. P. [Gesellschaft fuer Anlagen- und Reaktorsicherheit GRS MbH Forschungszentrum, Boltzmannstr. 14, 85748 Garching (Germany)
2012-07-01
The use of fine distributed moderating material to enhance the feedback effects and to reduce the sodium void effecting sodium cooled fast reactor cores is described. The influence of the moderating material on the neutron spectrum, the power distribution, and the burnup distribution is shown. The consequences of the use of fine distributed moderating material into fuel assemblies with fuel configurations foreseen for minor actinide transmutation is analyzed and the transmutation efficiency is compared. The degradation of the feedback effects due to the insertion of minor actinides and the compensation by the use of moderating materials is discussed. (authors)
Price, Daniel
.2) Collapse of a molecular cloud core to stellar densities: stellar core and outflow formation in radiation of rotating, magnetised molecular cloud cores to form protostars. The calculations follow the formation and evolution of the first hydrostatic core, the collapse to form a stellar core, the launching of outflows from
Core Competence and Education.
ERIC Educational Resources Information Center
Holmes, Gary; Hooper, Nick
2000-01-01
Outlines the concept of core competence and applies it to postcompulsory education in the United Kingdom. Adopts an educational perspective that suggests accreditation as the core competence of universities. This economic approach suggests that the market trend toward lifetime learning might best be met by institutions developing a core competence…
Coring Sample Acquisition Tool
NASA Technical Reports Server (NTRS)
Haddad, Nicolas E.; Murray, Saben D.; Walkemeyer, Phillip E.; Badescu, Mircea; Sherrit, Stewart; Bao, Xiaoqi; Kriechbaum, Kristopher L.; Richardson, Megan; Klein, Kerry J.
2012-01-01
A sample acquisition tool (SAT) has been developed that can be used autonomously to sample drill and capture rock cores. The tool is designed to accommodate core transfer using a sample tube to the IMSAH (integrated Mars sample acquisition and handling) SHEC (sample handling, encapsulation, and containerization) without ever touching the pristine core sample in the transfer process.
Enhancement of the inherent self-protection of the fast sodium reactor cores with oxide fuel
Eliseev, V.A.; Malisheva, I.V.; Matveev, V.I.; Egorov, A.V.; Maslov, P.A. [SSC RF - IPPE, Obninsk, Kaluga region (Russian Federation)
2013-07-01
With the development and research into the generation IV fast sodium reactors, great attention is paid to the enhancement of the core inherent self-protection characteristics. One of the problems dealt here is connected with the reduction of the reactivity margin so that the control rods running should not result in the core overheating and melting. In this paper we consider the possibilities of improving the core of BN-1200 with oxide fuel by a known method of introducing an axial fertile layer into the core. But unlike earlier studies this paper looks at the possibility of using such a layer not only for improving breeding, but also for reducing sodium void reactivity effect (SVRE). This proposed improvement of the BN-1200 core does not solve the problem of strong interference in control and protection system (CPS) rods of BN-1200, but they reduce significantly the reactivity margin for burn-up compensation. This helps compensate all the reactivity balances in the improved core configurations without violating constraints on SVRE value.
Low Voltage High Current PM Traction Motor Design Using Recent Core Loss Results
Marubini J. Manyage; Pragasen Pillay
2007-01-01
A number of core loss models, requiring different amounts of measured data have been investigated for a permanent magnet synchronous motor (PMSM) design for traction applications. This paper presents the effects of two of these core loss models in core loss calculations. A new core loss model with a physical basis for the variation of loss coefficients with frequency and
How Do Calculators Calculate? Helmut Knaust
Knaust, Helmut
How Do Calculators Calculate? Helmut Knaust Department of Mathematical Sciences University of Texas at El Paso Helmut Knaust How Do Calculators Calculate? April 25, 1997 1 / 18 #12;History We give an introduction to the CORDIC method used my most handheld calculators (such as the ones by Texas Instruments
NSDL National Science Digital Library
Sandra Kolb
In this activity, students will explore the characteristics of ice and explain the influencing factors by using Internet connections to polar field experiences, making their own ice cores and taking a field trip for obtaining a local ice core. The students will practice scientific journaling to document their observations. They will assemble their findings, develop a poster of their ice core and explain their observations. The 'ice is ice' misconception will be dispelled. Students will explain what scientists learn from ice cores and define basic vocabulary associated with ice cores.
Core Formation in Giant Gaseous Protoplanets
Ravit Helled; Gerald Schubert
2008-08-20
Sedimentation rates of silicate grains in gas giant protoplanets formed by disk instability are calculated for protoplanetary masses between 1 M_Saturn to 10 M_Jupiter. Giant protoplanets with masses of 5 M_Jupiter or larger are found to be too hot for grain sedimentation to form a silicate core. Smaller protoplanets are cold enough to allow grain settling and core formation. Grain sedimentation and core formation occur in the low mass protoplanets because of their slow contraction rate and low internal temperature. It is predicted that massive giant planets will not have cores, while smaller planets will have small rocky cores whose masses depend on the planetary mass, the amount of solids within the body, and the disk environment. The protoplanets are found to be too hot to allow the existence of icy grains, and therefore the cores are predicted not to contain any ices. It is suggested that the atmospheres of low mass giant planets are depleted in refractory elements compared with the atmospheres of more massive planets. These predictions provide a test of the disk instability model of gas giant planet formation. The core masses of Jupiter and Saturn were found to be ~0.25 M_Earth and ~0.5 M_Earth, respectively. The core masses of Jupiter and Saturn can be substantially larger if planetesimal accretion is included. The final core mass will depend on planetesimal size, the time at which planetesimals are formed, and the size distribution of the material added to the protoplanet. Jupiter's core mass can vary from 2 to 12 M_Earth. Saturn's core mass is found to be ~8 M_Earth.
23. CORE WORKER OPERATING A COREBLOWER THAT PNEUMATICALLY FILLED CORE ...
23. CORE WORKER OPERATING A CORE-BLOWER THAT PNEUMATICALLY FILLED CORE BOXES WITH RESIGN IMPREGNATED SAND AND CREATED A CORE THAT THEN REQUIRED BAKING, CA. 1950. - Stockham Pipe & Fittings Company, 4000 Tenth Avenue North, Birmingham, Jefferson County, AL
John H. Cohen; Thomas E. Williams; Ali G. Kadaster; Bill V. Liddell
2002-11-01
The ''Methane Hydrate Production from Alaskan Permafrost'' project is a three-year endeavor being conducted by Maurer Technology Inc. (MTI), Noble, and Anadarko Petroleum, in partnership with the U.S. DOE National Energy Technology Laboratory (NETL). The project's goal is to build on previous and ongoing R&D in the area of onshore hydrate deposition. The project team plans to design and implement a program to safely and economically drill, core and produce gas from arctic hydrates. The current work scope includes drilling and coring one well on Anadarko leases in FY 2003 during the winter drilling season. A specially built on-site core analysis laboratory will be used to determine some of the physical characteristics of the hydrates and surrounding rock. Prior to going to the field, the project team designed and conducted a controlled series of coring tests for simulating coring of hydrate formations. A variety of equipment and procedures were tested and modified to develop a practical solution for this special application. This Topical Report summarizes these coring tests. A special facility was designed and installed at MTI's Drilling Research Center (DRC) in Houston and used to conduct coring tests. Equipment and procedures were tested by cutting cores from frozen mixtures of sand and water supported by casing and designed to simulate hydrate formations. Tests were conducted with chilled drilling fluids. Tests showed that frozen core can be washed out and reduced in size by the action of the drilling fluid. Washing of the core by the drilling fluid caused a reduction in core diameter, making core recovery very difficult (if not impossible). One successful solution was to drill the last 6 inches of core dry (without fluid circulation). These tests demonstrated that it will be difficult to capture core when drilling in permafrost or hydrates without implementing certain safeguards. Among the coring tests was a simulated hydrate formation comprised of coarse, large-grain sand in ice. Results with this core showed that the viscosity of the drilling fluid must also be carefully controlled. When coarse sand was being cored, the core barrel became stuck because the drilling fluid was not viscous enough to completely remove the large grains of sand. These tests were very valuable to the project by showing the difficulties in coring permafrost or hydrates in a laboratory environment (as opposed to a field environment where drilling costs are much higher and the potential loss of equipment greater). Among the conclusions reached from these simulated hydrate coring tests are the following: Frozen hydrate core samples can be recovered successfully; A spring-finger core catcher works best for catching hydrate cores; Drilling fluid can erode the core and reduces its diameter, making it more difficult to capture the core; Mud must be designed with proper viscosity to lift larger cuttings; and The bottom 6 inches of core may need to be drilled dry to capture the core successfully.
Core materials development for the fuel cycle R&D program
NASA Astrophysics Data System (ADS)
Maloy, S. A.; Toloczko, M.; Cole, J.; Byun, T. S.
2011-08-01
The Fuel Cycle Research and Development program is investigating methods of burning minor actinides in a transmutation fuel. One of the challenges of achieving this goal is to develop fuels capable of reaching extreme burnup levels (e.g. 40%). To achieve such high burnup levels' fast reactor core materials (cladding and duct) must be able to withstand very high doses (>300 dpa design goal) while in contact with the coolant and the fuel. Thus, these materials must withstand radiation effects that promote low temperature embrittlement, radiation induced segregation, high temperature helium embrittlement, swelling, accelerated creep, corrosion with the coolant, and chemical interaction with the fuel (FCCI). To develop and qualify materials to a total fluence greater than 200 dpa requires development of advanced alloys and irradiations in fast reactors to test these alloys. Test specimens of ferritic/martensitic alloys (T91/HT-9) previously irradiated in the FFTF reactor up to 210 dpa at a temperature range of 350-750 °C are presently being tested. This includes analysis of a duct made of HT-9 after irradiation to a total dose of 155 dpa at temperatures from 370 to 510 °C. Compact tension, charpy and tensile specimens have been machined from this duct and mechanical testing as well as SANS and Mossbauer spectroscopy are currently being performed. Initial results from compression testing and Charpy testing reveal a strong increase in yield stress (˜400 MPa) and a large increase in DBTT (up to 230 °C) for specimens irradiated at 383 °C to a dose of 28 dpa. Less hardening and a smaller increase in DBTT was observed for specimens irradiated at higher temperatures up to 500 °C. Advanced radiation tolerant materials are also being developed to enable the desired extreme fuel burnup levels. Specifically, coatings are being developed to minimize FCCI, and research is underway to fabricate large heats of radiation tolerant oxide dispersion steels with homogeneous oxide dispersions.
Core materials development for the fuel cycle R&D program
Toloczko, M; Maloy, S; Cole, James I.; Byun, Thak Sang
2011-01-01
The Fuel Cycle Research and Development program is investigating methods of burning minor actinides in a transmutation fuel. One of the challenges of achieving this goal is to develop fuels capable of reaching extreme burnup levels (e.g. 40%). To achieve such high burnup levels fast reactor core materials (cladding and duct) must be able to withstand very high doses (>300 dpa design goal) while in contact with the coolant and the fuel. Thus, these materials must withstand radiation effects that promote low temperature embrittlement, radiation induced segregation, high temperature helium embrittlement, swelling, accelerated creep, corrosion with the coolant, and chemical interaction with the fuel (FCCI). To develop and qualify materials to a total fluence greater than 200 dpa requires development of advanced alloys and irradiations in fast reactors to test these alloys. Test specimens of ferritic/martensitic alloys (T91/HT-9) previously irradiated in the FFTF reactor up to 210 dpa at a temperature range of 350 750 C are presently being tested. This includes analysis of a duct made of HT-9 after irradiation to a total dose of 155 dpa at temperatures from 370 to 510 C. Compact tension, charpy and tensile specimens have been machined from this duct and mechanical testing as well as SANS and Mossbauer spectroscopy are currently being performed. Initial results from compression testing and Charpy testing reveal a strong increase in yield stress (400 MPa) and a large increase in DBTT (up to 230 C) for specimens irradiated at 383 C to a dose of 28 dpa. Less hardening and a smaller increase in DBTT was observed for specimens irradiated at higher temperatures up to 500 C. Advanced radiation tolerant materials are also being developed to enable the desired extreme fuel burnup levels. Specifically, coatings are being developed to minimize FCCI, and research is underway to fabricate large heats of radiation tolerant oxide dispersion steels with homogeneous oxide dispersions.
Core Materials Development for the Fuel Cycle R&D Program
S. A. Maloy; M. Toloczko; J. Cole; T. S. Byun
2011-08-01
The Fuel Cycle Research and Development program is investigating methods of burning minor actinides in a transmutation fuel. One of the challenges of achieving this goal is to develop fuels capable of reaching extreme burnup levels (e.g. 40%). To achieve such high burnup levels fast reactor core materials (cladding and duct) must be able to withstand very high doses (greater than 300 dpa design goal) while in contact with the coolant and the fuel. Thus, these materials must withstand radiation effects that promote low temperature embrittlement, radiation induced segregation, high temperature helium embrittlement, swelling, accelerated creep, corrosion with the coolant, and chemical interaction with the fuel (FCCI). To develop and qualify materials to a total fluence greater than 200 dpa requires development of advanced alloys and irradiations in fast reactors to test these alloys. Test specimens of ferritic/martensitic alloys (T91/HT-9) previously irradiated in the FFTF reactor up to 210 dpa at a temperature range of 350-750 C are presently being tested. This includes analysis of a duct made of HT-9 after irradiation to a total dose of 155 dpa at temperatures from 370 to 510 C. Compact tension, charpy and tensile specimens have been machined from this duct and mechanical testing as well as SANS and Mossbauer spectroscopy are currently being performed. Initial results from compression testing and Charpy testing reveal a strong increase in yield stress ({approx}400 MPa) and a large increase in DBTT (up to 230 C) for specimens irradiated at 383 C to a dose of 28 dpa. Less hardening and a smaller increase in DBTT was observed for specimens irradiated at higher temperatures up to 500 C. Advanced radiation tolerant materials are also being developed to enable the desired extreme fuel burnup levels. Specifically, coatings are being developed to minimize FCCI, and research is underway to fabricate large heats of radiation tolerant oxide dispersion steels with homogeneous oxide dispersions.
NASA Technical Reports Server (NTRS)
Gheen, Darrell
2007-01-01
A tool makes a cut perpendicular to the cylindrical axis of a core hole at a predetermined depth to free the core at that depth. The tool does not damage the surrounding material from which the core was cut, and it operates within the core-hole kerf. Coring usually begins with use of a hole saw or a hollow cylindrical abrasive cutting tool to make an annular hole that leaves the core (sometimes called the plug ) in place. In this approach to coring as practiced heretofore, the core is removed forcibly in a manner chosen to shear the core, preferably at or near the greatest depth of the core hole. Unfortunately, such forcible removal often damages both the core and the surrounding material (see Figure 1). In an alternative prior approach, especially applicable to toxic or fragile material, a core is formed and freed by means of milling operations that generate much material waste. In contrast, the present tool eliminates the damage associated with the hole-saw approach and reduces the extent of milling operations (and, hence, reduces the waste) associated with the milling approach. The present tool (see Figure 2) includes an inner sleeve and an outer sleeve and resembles the hollow cylindrical tool used to cut the core hole. The sleeves are thin enough that this tool fits within the kerf of the core hole. The inner sleeve is attached to a shaft that, in turn, can be attached to a drill motor or handle for turning the tool. This tool also includes a cutting wire attached to the distal ends of both sleeves. The cutting wire is long enough that with sufficient relative rotation of the inner and outer sleeves, the wire can cut all the way to the center of the core. The tool is inserted in the kerf until its distal end is seated at the full depth. The inner sleeve is then turned. During turning, frictional drag on the outer core pulls the cutting wire into contact with the core. The cutting force of the wire against the core increases with the tension in the wire and, hence, with the frictional drag acting on the outer sleeve. As the wire cuts toward the center of the core, the inner sleeve rotates farther with respect to the outer sleeve. Once the wire has cut to the center of the core, the tool and the core can be removed from the hole. The proper choice of cutting wire depends on the properties of the core material. For a sufficiently soft core material, a nonmetallic monofilament can be used. For a rubber-like core material, a metal wire can be used. For a harder core material, it is necessary to use an abrasive wire, and the efficiency of the tool can be increased greatly by vacuuming away the particles generated during cutting. For a core material that can readily be melted or otherwise cut by use of heat, it could be preferable to use an electrically heated cutting wire. In such a case, electric current can be supplied to the cutting wire, from an electrically isolated source, via rotating contact rings mounted on the sleeves.
Core Forensics: Earth's Accretion and Differentiation
NASA Astrophysics Data System (ADS)
Badro, J.; Brodholt, J. P.; Siebert, J.; Piet, H.; Ryerson, F. J.
2013-12-01
Earth's accretion and its primitive differentiation are intimately interlinked processes. One way to constrain accretionary processes is by looking at the major differentiation event that took place during accretion: core formation. Understanding core formation and core composition can certainly shed a new light on early and late accretionary processes. On the other hand, testing certain accretionary models and hypothesis (fluxes, chemistries, timing) allows -short of validating them- at the very least to unambiguously refute them, through the 'filter'' of core formation and composition. Earth's core formed during accretion as a result of melting, phase-separation, and segregation of accretionary building blocks (from meteorites to planetesimals). The bulk composition of the core and mantle depends on the evolution (pressure, temperature, composition) of core extraction during accretion. The entire process left a compositional imprint on both reservoirs: (1) in the silicate Earth, in terms of siderophile trace-element (Ni, Co, V, Cr, among others) concentrations and isotopic fractionation (Si, Cu, among others), a record that is observed in present-day mantle rocks; and (2) on the core, in terms of major element composition and light elements dissolved in the metal, a record that is observed by seismology through the core density-deficit. This imprint constitutes actually a fairly impressive set of evidence (siderophile element concentration and fractionation, volatile and siderophile element isotopic fractionation), can be used today to trace back the primordial processes that occurred 4.5 billion years ago. We are seeking to provide an overhaul of the standard core formation/composition models, by using a new rationale that bridges geophysics and geochemistry. The new ingredients are (1) new laser-heated diamond anvil cell partitioning data, dramatically extending the previous P-T conditions for experimental work, (2) ab initio molecular dynamics calculations to estimate outer-core density and bulk sound velocity, and combine it with seismology to define a range of possible compositions of the core that satisfies the observations, (3) a refined core formation model bringing together the continuousness of the overall process with the discreetness of the final impacts, and equilibrium thermodynamics with the non-equilibrium nature of certain processes (giant impacts, deep magma ocean). We propose a few strong constraints that come out from our models: (1) the Earth accreted in a rather oxidizing environment, (2) yielding an oxygen-rich core, in a (3) deep magma ocean (~1500 km) that could have (4) never been fully molten or fully equilibrated, at least during core extraction, despite the giant impacts.
A Fission Gas Release Model for High-Burnup LWR ThOâ-UOâ Fuel
Yun Long; Yi Yuan; Mujid S. Kazimi; Ronald G. Ballinger; Edward E. Pilat
2002-01-01
Fission gas release in thoria-urania fuel has been investigated by creating a specially modified FRAPCON-3 code. Because of the reduced buildup of Â²Â³â¹Pu and a flatter distribution of Â²Â³Â³U, the new model THUPS (Thoria-Urania Power Shape) was developed to calculate the radial power distribution, including the effects of both plutonium and Â²Â³Â³U. Additionally, a new porosity model for the rim
The compactness of presupernova stellar cores
Sukhbold, Tuguldur; Woosley, S. E.
2014-03-01
The success or failure of the neutrino-transport mechanism for producing a supernova in an evolved massive star is known to be sensitive not only to the mass of the iron core that collapses, but also to the density gradient in the silicon and oxygen shells surrounding that core. Here we study the systematics of a presupernova core's 'compactness' as a function of the mass of the star and the physics used in its calculation. Fine-meshed surveys of presupernova evolution are calculated for stars from 15 to 65 M {sub ?}. The metallicity and the efficiency of semiconvection and overshoot mixing are both varied and bare carbon-oxygen cores are explored as well as full hydrogenic stars. Two different codes, KEPLER and MESA, are used for the study. A complex interplay of carbon and oxygen burning, especially in shells, can cause rapid variations in the compactness for stars of very nearly the same mass. On larger scales, the distribution of compactness with main sequence mass is found to be robustly non-monotonic, implying islands of 'explodabilty,' particularly around 8-20 M {sub ?} and 25-30 M {sub ?}. The carbon-oxygen (CO) core mass of a presupernova star is a better, (though still ambiguous) discriminant of its core structure than the main sequence mass.
PRIZMA predictions of in-core detection indications in the VVER-1000 reactor
NASA Astrophysics Data System (ADS)
Kandiev, Yadgar Z.; Kashayeva, Elena A.; Malyshin, Gennady N.; Modestov, Dmitry G.; Khatuntsev, Kirill E.
2014-06-01
The paper describes calculations which were done by the PRIZMA code(1) to predict indications of in-core rhodium detectors in the VVER-1000 reactor for some core fragments with allowance for fuel and rhodium burnout.
Ermler, Walter V.; Tilson, Jeffrey L.
2012-12-15
A procedure for structuring generally contracted valence-core/valence basis sets of Gaussian-type functions for use with relativistic effective core potentials (gcv-c/v-RECP basis sets) is presented. Large valence basis sets are enhanced using a compact basis set derived for outer core electrons in the presence of small-core RECPs. When core electrons are represented by relativistic effective core potentials (RECPs), and appropriate levels of theory, these basis sets are shown to provide accurate representations of atomic and molecular valence and outer-core electrons. Core/valence polarization and correlation effects can be calculated using these basis sets through standard methods for treating electron correlation. Calculations of energies and spectra for Ru, Os, Ir, In and Cs are reported. Spectroscopic constants for RuO2+, OsO2+, Cs2 and InH are calculated and compared with experiment.
NASA Technical Reports Server (NTRS)
Akins, James; Cobb, Billy; Hart, Steve; Leaptrotte, Jeff; Milhollin, James; Pernik, Mark
1989-01-01
The problem of retrieving and storing core samples from a hole drilled on the lunar surface is addressed. The total depth of the hole in question is 50 meters with a maximum diameter of 100 millimeters. The core sample itself has a diameter of 60 millimeters and will be two meters in length. It is therefore necessary to retrieve and store 25 core samples per hole. The design utilizes a control system that will stop the mechanism at a certain depth, a cam-linkage system that will fracture the core, and a storage system that will save and catalogue the cores to be extracted. The Rod Changer and Storage Design Group will provide the necessary tooling to get into the hole as well as to the core. The mechanical design for the cam-linkage system as well as the conceptual design of the storage device are described.
G. S. Chang; M. A. Lillo; R. G. Ambrosek
2008-06-01
The Advanced Test Reactor (ATR) is a high power density and high neutron flux research reactor operating in the United States. Powered with highly enriched uranium (HEU), the ATR has a maximum thermal power rating of 250 MWth. Because of the large test volumes located in high flux areas, the ATR is an ideal candidate for assessing the feasibility of converting an HEU driven reactor to a low-enriched core. The present work investigates the necessary modifications and evaluates the subsequent operating effects of this conversion. A detailed plate-by-plate MCNP ATR 1/8th core model was developed and validated for a fuel cycle burnup comparison analysis. Using the current HEU U 235 enrichment of 93.0 % as a baseline, an analysis was performed to determine the low-enriched uranium (LEU) density and U-235 enrichment required in the fuel meat to yield an equivalent K-eff versus effective full power days (EFPDs) between the HEU and the LEU cores. The MCNP ATR 1/8th core model was used to optimize the U 235 loading in the LEU core, such that the differences in K-eff and heat flux profiles between the HEU and LEU cores were minimized. The depletion methodology MCWO was used to calculate K-eff versus EFPDs in this paper. The MCWO-calculated results for the LEU demonstrated adequate excess reactivity such that the K-eff versus EFPDs plot is similar to the ATR reference HEU case study. Each HEU fuel element contains 19 fuel plates with a fuel meat thickness of 0.508 mm (20 mil). In this work, the proposed LEU (U-10Mo) core conversion case with nominal fuel meat thickness of 0.330 mm (13 mil) and U-235 enrichment of 19.7 wt% is used to optimize the radial heat flux profile by varying the fuel meat thickness from 0.191 mm (7.0 mil) to 0.330 mm (13.0 mil) at the inner 4 fuel plates (1-4) and outer 4 fuel plates (16-19). A 0.8g of Boron-10, a burnable absorber, was added in the inner and outer plates to reduce the initial excess reactivity, and the peak to average ratio of the inner/outer heat flux more effectively. Because the B-10 (n,a) reaction will produce Helium-4 (He-4), which might degrade the LEU foil type fuel performance, an alternative absorber option is proposed. The proposed LEU case study will have 6.918 g of Cadmium (Cd) mixed with the LEU at the inner 4 fuel plates (1-4) and outer 4 fuel plates (16-19) as a burnable absorber to achieve peak to average ratios similar to those for the ATR reference HEU case study.
KSI's Cross Insulated Core Transformer Technology
Uhmeyer, Uwe [Kaiser Systems, Inc, 126 Sohier Road, Beverly, MA 01915 (United States)
2009-08-04
Cross Insulated Core Transformer (CCT) technology improves on Insulated Core Transformer (ICT) implementations. ICT systems are widely used in very high voltage, high power, power supply systems. In an ICT transformer ferrite core sections are insulated from their neighboring ferrite cores. Flux leakage is present at each of these insulated gaps. The flux loss is raised to the power of stages in the ICT design causing output voltage efficiency to taper off with increasing stages. KSI's CCT technology utilizes a patented technique to compensate the flux loss at each stage of an ICT system. Design equations to calculate the flux compensation capacitor value are presented. CCT provides corona free operation of the HV stack. KSI's CCT based High Voltage power supply systems offer high efficiency operation, high frequency switching, low stored energy and smaller size over comparable ICT systems.
Calculations of thermal-reactor spent-fuel nuclide inventories and comparisons with measurements
Wilson, W.B.; LaBauve, R.J.; England, T.R.
1982-01-01
Comparisons with integral measurements have demonstrated the accuracy of CINDER codes and libraries in calculating aggregate fission-product properties, including neutron absorption, decay power, and decay spectra. CINDER calculations have, alternatively, been used to supplement measured integral data describing fission-product decay power and decay spectra. Because of the incorporation of the extensive actinide library and the use of ENDF/B-V data, it is desirable to compare the inventory of individual nuclides obtained from tandem EPRI-CELL/CINDER-2 calculations with those determined in documented benchmark inventory measurements of spent reactor fuel. The development of the popular /sup 148/Nd burnup measurement procedure is outlined, and areas of uncertainty in it and lack of clarity in its interpretation are indicated. Six inventory samples of varying quality and completeness are examined. The power histories used in the calculations have been listed for other users.
Dublin Core Metadata Initiative
NSDL National Science Digital Library
Dublin Core metadata has been implemented in several ways, including as HTML metatags and as database elements, as it is used in the Scout Archives (discussed in the June 20, 1997 issue of the Scout Report). The DC elements are title, author, subject, description, publisher, other contributor, date, resource type, format, resource identifier, source, language, relation, coverage, and rights management. Information about the Dublin Core Workshop Series, DC semantics and syntax, working papers, and projects that have implemented Dublin Core metadata can be found at the Dublin Core Metadata homepage.
Hicks, Joshua; Adrian, Betty
2009-01-01
The Core Research Center (CRC) of the U.S. Geological Survey (USGS), located at the Denver Federal Center in Lakewood, Colo., currently houses rock core from more than 8,500 boreholes representing about 1.7 million feet of rock core from 35 States and cuttings from 54,000 boreholes representing 238 million feet of drilling in 28 States. Although most of the boreholes are located in the Rocky Mountain region, the geologic and geographic diversity of samples have helped the CRC become one of the largest and most heavily used public core repositories in the United States. Many of the boreholes represented in the collection were drilled for energy and mineral exploration, and many of the cores and cuttings were donated to the CRC by private companies in these industries. Some cores and cuttings were collected by the USGS along with other government agencies. Approximately one-half of the cores are slabbed and photographed. More than 18,000 thin sections and a large volume of analytical data from the cores and cuttings are also accessible. A growing collection of digital images of the cores are also becoming available on the CRC Web site Internet http://geology.cr.usgs.gov/crc/.
VLBI Observations of the Free Core Nutations
NASA Astrophysics Data System (ADS)
Smylie, D. E.
2012-12-01
At core scale lengths with periods from a few hours to days, the Coriolis acceleration dominates the Lorentz force density and core modes can be considered as purely mechanical. One of the most interesting core modes is the spin-over mode, which reflects the ability of the outer core to rotate about an axis different from that of either the inner core or the shell. It has a nearly diurnal period. In the Earth frame of reference, this mode produces the nearly diurnal retrograde wobble. In the space frame of reference it is accompanied by the free core nutations. When the flattening of the boundaries of the fluid outer core and the figure-figure gravitational coupling are taken into account, as well as the deformability of the boundaries, both a retrograde free core nutation and a prograde free core nutation are found. The retrograde free core nutation was first predicted by Poincare (1910) for a completly fluid, incompressible core bounded by a rigid shell. In a variational calculation of wobble-nutation modes in realistic Earth models, Jiang (1993) found the classical retrograde free core nutation (RFCN) but discovered a prograde free core nutation (PFCN) as well. VLBI residuals in longitude and obliquity compared to the 1980 IAU nutation series, and their standard errors, were downloaded from the Goddard Space Flight Center website, for the period August 3, 1979 to March 6, 2003, giving 3343 points over a span of 8617 days. In an overlapping segment analysis, the discrete Fourier transform (DFT) for each segment was found for the corresponding series of unequally spaced nutation residuals by singular value decomposition (SVD), with the number of singular values eliminated determined by the satisfaction of Parseval's theorem. Both the RFCN and the PFCN resonances were found in the resulting power spectrum. The nutation resonances were found to be in free decay, the half-life of the PFCN at 2620 days and that of the RFCN at 2229 days, with Ekman boundary layer theory leading to viscosities at the top of the core of 3124 Pa s and 3611 Pa s, respectively.
Hydrologic characterization of four cores from the Geysers Coring Project
Persoff, Peter; Hulen, Jeffrey B.
1996-01-24
Results of hydrologic tests conducted on four representative core plugs from Geysers Coring Project drill hole SB-15-D have been related to detailed mineralogic and textural characterization of the plugs to yield new information about permeability, porosity, and capillary-pressure characteristics of the uppermost Geysers steam reservoir and its immediately overlying caprock. The core plugs are all fine- to medium-grained, Franciscan-assemblage (late Mesozoic) metagraywacke with sparse Franciscan metamorphic quartz-calcite veins and late Cenozoic, hydrothermal quartz-calcite-pyrite veins. The matrices of three plugs from the caprock are rich in metamorphic mixed-layer illite/smectite and disseminated hydrothermal pyrite; the reservoir plug instead contains abundant illite and only minor pyrite. The reservoir plug and one caprock plug are sparsely disrupted by latest-stage, unmineralized microfractures which both follow and crosscut veinlets but which could be artifacts. Porosities of the plugs, measured by Boyles-law gas expansion, range between 1.9 and 2.5%. Gas permeability and Klinkenberg slip factor were calculated from gas-pressure-pulse-decay measurements using a specially designed permeameter with small (2 mL) reservoirs. Matrix permeabilities in the range 10^{-21} m² ( = 1 nanodarcy) were measured for two plugs that included mineral-filled veins but no unfilled microfractures. Greater permeabilities were measured on plugs that contained microfractures; at 500 psi net confining pressure, an effective aperture of 1.6 µm was estimated for one plug. Capillary pressure curves were determined for three cores by measuring saturation as weight gain of plugs equilibrated with atmospheres in which the relative humidity was controlled by saturated brines.
Complete Depletion in prestellar cores
C. M. Walmsley; D. R. Flower; G. Pineau des Forets
2004-02-20
We have carried out calculations of ionization equilibrium and deuterium fractionation for conditions appropriate to a completely depleted, low mass pre--protostellar core, where heavy elements such as C, N, and O have vanished from the gas phase and are incorporated in ice mantles frozen on dust grain surfaces. We put particular emphasis on the interpretation of recent observations of H2D+ towards the centre of the prestellar core L1544 (Caselli et al. 2003) and also compute the ambipolar diffusion timescale. We consider explicitly the ortho and para forms of H2,H3+, and H2D+. Our results show that the ionization degree under such conditions depends sensitively on the grain size distribution or, more precisely, on the mean grain surface area per hydrogen nucleus. Depending upon this parameter and upon density, the major ion may be H+, H3+, or D3+. We show that the abundance of ortho-H2D+ observed towards L1544 can be explained satisfactorily in terms of a complete depletion model and that this species is, as a consequence, an important tracer of the kinematics of prestellar cores.
Effect of burn-up on the thermal conductivity of uranium-gadolinium dioxide up to 100 GWd/tHM
NASA Astrophysics Data System (ADS)
Staicu, D.; Rondinella, V. V.; Walker, C. T.; Papaioannou, D.; Konings, R. J. M.; Ronchi, C.; Sheindlin, M.; Sasahara, A.; Sonoda, T.; Kinoshita, M.
2014-10-01
The thermal diffusivity of reactor irradiated (U,Gd)O2 fuels has been measured, for burn-ups from 33 to 97 GWd tHM-1 and for irradiation temperatures from 670 to 1580 K. Measurements under thermal annealing cycles were performed in order to investigate the recovery of the thermal conductivity as a function of temperature. The analysis of the results showed a lower thermal conductivity for (U,Gd)O2 when compared to UO2, with similar effects of the burn-up and irradiation temperature. A correlation for the thermal conductivity could be proposed on the basis of that for UO2 presented in an earlier work, which describes the separate effects of soluble fission products, of fission gas frozen in dynamical solution and of radiation damage.
A. V. Der Linde; J. H. N. Verheugen
1982-01-01
Two irradiation tests were conducted in the pressurized water reactor loop of the High Flux Reactor at Petten to compare the low burnup (0.4% fima), high power (35 to 60 kW x mâ»Â¹) performance of MOâ sphere-pac fuel with the plutonium incorporated only in the large (1000-..mu..m) diameter spheres and that of homogeneously enriched UOâ sphere-pac and pellet fuel. The
Burn-up cross sections of /sup 51/Cr, /sup 59/Fe, /sup 65/Zn, /sup 86/Rb, /sup 103/Ru
Katcoff, S.
1985-01-01
Targets of Cr, Fe, Zn, Rb, and Ru were irradiated in the hydraulic tube of the Oak Ridge HFIR reactor at a neutron flux of 2.6 x 10/sup 15/ n/cm/sup 2/sec for 1 day and 20 days. The reactor burn-up cross sections (in barns) of the radioactive product nuclides are: /sup 51/Cr, <10; /sup 59/Fe, <10; /sup 65/Zn, 60 +- 30; /sup 86/Rb, <20; /sup 103/Ru, <20.
Integrated calculator programs for pharmacokinetic calculations.
Robb, R A; Bauer, L A; Koup, J R
1982-05-01
A package of integrated programs for calculating pharmacokinetic variables and drug-dosing regimens using a hand-held programmable calculator is described. Twelve pharmacokinetic programs, which were based on previously published pharmacokinetic equations, were developed for use in a HP-41C hand-held calculator (Hewlett-Packard). The programs perform, pharmacokinetic calculations for many drugs, including digoxin, theophylline, phenytoin, nd the aminoglycosides. Also programs for ideal body weight, body surface area, and creatinine clearance calculations are included. Eleven of the 12 programs can be stored in the calculator at any time. Values generated in one program are stored in memory registers and can be recalled directly for use in other programs. The calculator has a continuous memory; therefore, all stored data, programs, and functions are maintained when the calculator is turned off. The integrated calculator programs provide a quick and reliable means of applying pharmacokinetic principles to everyday hospital pharmacy practice. PMID:7081256
On-Site was developed to provide modelers and model reviewers with prepackaged tools ("calculators") for performing site assessment calculations. The philosophy behind OnSite is that the convenience of the prepackaged calculators helps provide consistency for simple calculations,...
Modeling of Nomex® Honeycomb Cores, Linear and Nonlinear Behaviors
L. Gornet; S. Marguet; G. Marckmann
2007-01-01
The purpose of this study is to develop tools dedicated to the design of sandwich panels involving Nomex® honeycomb cores. Special attention is paid to the ability to perform full three dimensional calculations up to failure of such structures. In the first part, the determination of effective elastic properties of Nomex® honeycomb cores is carried out thanks to strain based
Improved formulations for rotational core losses in rotating electrical machines
Jian Guo Zhu; Victor Stuart Ramsden
1998-01-01
This paper presents the measurement and modeling of rotational core losses in electrical steel sheets and rotating electrical machines. Novel formulations of rotational hysteresis, eddy current, and excess losses in electrical sheet steels with circular and elliptical rotating flux density vectors are reported. These formulations are used together with the finite element method to calculate the core losses in rotating
ERIC Educational Resources Information Center
Michigan State Univ., East Lansing. Inst. for International Studies in Education.
This collection of core bibliographies, which expands on an initial bibliography published in 1979 of the core resources housed in the Non-Formal Education Information Center at Michigan State University, comprises a basic stock of materials on nonformal education and women in development that have been contributed by development planners,…
Secure Core Contact Information
Secure Core Contact Information C. E. Irvine irvine@nps.edu 831-656-2461 Department of Computer for the secure management of local and/or remote information in multiple contexts. The SecureCore project Science Graduate School of Operations and Information Sciences www.cisr.nps.edu Project Description
ERIC Educational Resources Information Center
Chan, Monnica
2013-01-01
From a higher education perspective, new "Common Core" standards could improve student college-readiness levels, reduce institutional remediation rates, and close education gaps in and between states. As a national initiative to create common educational standards for students across multiple states, the Common Core State Standards…
ERIC Educational Resources Information Center
Krim, Jessica; Brody, Michael
2008-01-01
What can glaciers tell us about volcanoes and atmospheric conditions? How does this information relate to our understanding of climate change? Ice Core Investigations is an original and innovative activity that explores these types of questions. It brings together popular science issues such as research, climate change, ice core drilling, and air…
ERIC Educational Resources Information Center
Kopaska-Merkel, David C.
1995-01-01
Explains an activity in which students construct a simulated ice core. Materials required include only a freezer, food coloring, a bottle, and water. This hands-on exercise demonstrates how a glacier is formed, how ice cores are studied, and the nature of precision and accuracy in measurement. Suitable for grades three through eight. (Author/PVD)
Benker, D.E.; Bigelow, J.E.; Bond, W.D.; Chattin, F.R.; King, L.J.; Kitts, F.G.; Ross, R.G.; Stacy, R.G.
1985-01-01
A batch of irradiated Fast Flux Test Facility (FFTF) fuel was processed for the first time in the Solvent Extraction Test Facility (SETF) at the Oak Ridge National Laboratory (ORNL) during Campaign 7. The average burnup of the fuel was only 0.2 atom %, but the cooling time was short enough ({similar_to}2 years) so that {sup 95}Zr was detected in the feed. This short cooling permitted our first measurement of {sup 95}Zr decontamination factors (DFs) without having to use tracers. No operational problems were noted in the operation of the extraction-scrubbing contactor, and low uranium and plutonium losses (< 0.01%) were measured. Fission product DFs were improved noticeably by increasing the number of scrub stages from six to eight. Two flowsheet options for making pure uranium and plutonium products (total partitioning) were tested. Each flowsheet used hydroxylamine nitrate for reducing plutonium. Good products were obtained (uranium DFs of > 10{sup 3} and plutonium DFs of > 10{sup 4}), but each flowsheet was troubled with plutonium reoxidation. Adding hydrazine and lowering the plutonium concentration lessened the problem but did not eliminate it. About 370 g of plutonium was recovered from these tests, purified by anion exchange, converted to PuO{sub 2}, and transferred to the fuel refabrication program. 7 references.
NASA Astrophysics Data System (ADS)
Uwaba, Tomoyuki; Ito, Masahiro; Maeda, Koji
2011-09-01
The C3M irradiation test, which was conducted in the experimental fast reactor, "Joyo", demonstrated that mixed oxide (MOX) fuel pins with austenitic steel cladding could attain a peak pellet burnup of about 130 GWd/t safely. The test fuel assembly consisted of 61 fuel pins, whose design specifications were similar to those of driver fuel pins of a prototype fast breeder reactor, "Monju". The irradiated fuel pins exhibited diametral strain due to cladding void swelling and irradiation creep. The cladding irradiation creep strain were due to the pellet-cladding mechanical interaction (PCMI) as well as the internal gas pressure. From the fuel pin ceramographs and 137Cs gamma scanning, it was found that the PCMI was associated with the pellet swelling which was enhanced by the rim structure formation or by cesium uranate formation. The PCMI due to cesium uranate, which occurred near the top of the MOX fuel column, significantly affected cladding hoop stress and thermal creep, and the latter effect tended to increase the cumulative damage fraction (CDF) of the cladding though the CDF indicated that the cladding still had some margin to failure due to the creep damage.
Protostellar core instabilities
NASA Astrophysics Data System (ADS)
Tscharnuter, W. M.
An axisymmetric collapse model for the formation of the presolar nebula and a spherically symmetric, Larson-type protostellar evolution is discussed. Particular attention is paid to the dynamical behavior of the starlike core. It has been found that, subsequent to their formation, protostellar core embryos containing only a few percents of a solar mass tend to undergo oscillations of large amplitudes. Dynamical oscillations and even disruptive core expansion ('hiccups') have also been observed to occur in spherically symmetric models. On the basis of Baker's (1966) one-zone-model, it can be argued that protostellar cores are vibrationally unstable and may even become dynamically unstable. Thus, oscillations of protostellar cores are to be expected, but their amplitudes, as observed in the model sequences, could be very sensitive to the accuracy achieved by the numerical solution of the discretized structure equations.
Colle, E.A. Jr.; Yates, D.N. Jr.; Brieger, E.F.
1986-09-02
An apparatus is described for taking core samples from the sidewall of a borehole in a well, the apparatus comprising: a string of drill pipe; at least one gun housing connected to the downhole end of the drill string; at least one coring bullet radially disposed within the gun housing, the coring bullet arranged for securing formation samples from the sidewall of the borehole; a charge assembly for propelling the coring bullet toward the sidewall, the charge assembly comprising: a detonatable cord having a diameter substantially in the range of approximately 0.125 to 0.150 inches extending generally axially through the housing from the uphole to the downhole end thereof; at least one cartridge assembly disposed within the housing between the cord and the bullet; the cartridge assembly including a pyrotechnic charge for propelling the bullet, a cable connecting the coring bullet to the housing, whereby the bullet may be retrieved from the sidewall.
On the Minimum Core Mass for Giant Planet Formation
NASA Astrophysics Data System (ADS)
Piso, Ana-Maria; Youdin, Andrew; Murray-Clay, Ruth
2013-07-01
The core accretion model proposes that giant planets form by the accretion of gas onto a solid protoplanetary core. Previous studies have found that there exists a "critical core mass" past which hydrostatic solutions can no longer be found and unstable atmosphere collapse occurs. This core mass is typically quoted to be around 10Me. In standard calculations of the critical core mass, planetesimal accretion deposits enough heat to alter the luminosity of the atmosphere, increasing the core mass required for the atmosphere to collapse. In this study we consider the limiting case in which planetesimal accretion is negligible and Kelvin-Helmholtz contraction dominates the luminosity evolution of the planet. We develop a two-layer atmosphere model with an inner convective region and an outer radiative zone that matches onto the protoplanetary disk, and we determine the minimum core mass for a giant planet to form within the typical disk lifetime for a variety of disk conditions. We denote this mass as critical core mass. The absolute minimum core mass required to nucleate atmosphere collapse is ˜ 8Me at 5 AU and steadily decreases to ˜ 3.5Me at 100 AU, for an ideal diatomic gas with a solar composition and a standard ISM opacity law. Lower opacity and disk temperature significantly reduce the critical core mass, while a decrease in the mean molecular weight of the nebular gas results in a larger critical core mass. Our results yield lower mass cores than corresponding studies for large planetesimal accretion rates.
34. DESPATCH CORE OVENS, GREY IRON FOUNDRY CORE ROOM, BAKES ...
34. DESPATCH CORE OVENS, GREY IRON FOUNDRY CORE ROOM, BAKES CORES THAT ARE NOT MADE ON HEATED OR COLD BOX CORE MACHINES, TO SET BINDING AGENTS MIXED WITH THE SAND CREATING CORES HARD ENOUGH TO WITHSTAND THE FLOW OF MOLTEN IRON INSIDE A MOLD. - Stockham Pipe & Fittings Company, Grey Iron Foundry, 4000 Tenth Avenue North, Birmingham, Jefferson County, AL
Collapse and fragmentation of molecular cloud cores. I - Moderately centrally condensed cores
NASA Technical Reports Server (NTRS)
Boss, Alan P.
1993-01-01
3D calculations of the collapse of moderately centrally condensed molecular cloud cores with varied thermal and rotational energies are presented. The calculations are carried out using a newly developed and tested second-order accurate radiative hydrodynamics code. Because of the use of a second-order accurate numerical scheme and initial clouds that resemble both observed prolate molecular cloud cores and magnetically supported clouds at the initiation of the dynamic collapse phase, the new models provide a superior estimate of the likelihood of fragmentation as a mechanism for binary star formation.
Playful calculation : tangible coding for visual calculation
Ham, Derek (Derek Allen)
2015-01-01
Play and calculation are often considered to be at odds. Play embraces the wildness of youth, imagination, and a sense of freedom. Calculation, to most, represents rigor, mechanistic behavior, and following inflexible ...
Enhanced transferability for Bethe-Salpeter Calculations
NASA Astrophysics Data System (ADS)
Shirley, Eric L.
2015-03-01
We have systematized projector-augmented-wave methods to reliably augment plane-wave/pseudopotential Bloch functions in atomic core regions for purposes of performing screening calculations, evaluating transition matrix elements, and evaluating Slater integrals in the condensed matter environment. This has improved the accuracy of core-hole screening, adherence to sum rules, and control of the strength of absorption features. This also ensures that transition matrix elements and concomitant core excitation spectra are reliable over significant energy ranges. To accomplish this, we improve the quality of the pseudopotentials (which become harder), extending norm conservation, and increasing the number of ``valence electrons.'' We present results for both insulators and metals, and for both core and valence excitations. Comparison to experimental data is a key part of this work. We also emphasize what approximations remain to be tackled in the treatment of electronic excitation spectra, many of which are more difficult to treat than what is within the scope of this work.
NASA Astrophysics Data System (ADS)
Teague, Melissa C.; Fromm, Bradley S.; Tonks, Michael R.; Field, David P.
2014-12-01
Nuclear energy is a mature technology with a small carbon footprint. However, work is needed to make current reactor technology more accident tolerant and to allow reactor fuel to be burned in a reactor for longer periods of time. Optimizing the reactor fuel performance is essentially a materials science problem. The current understanding of fuel microstructure have been limited by the difficulty in studying the structure and chemistry of irradiated fuel samples at the mesoscale. Here, we take advantage of recent advances in experimental capabilities to characterize the microstructure in 3D of irradiated mixed oxide (MOX) fuel taken from two radial positions in the fuel pellet. We also reconstruct these microstructures using Idaho National Laboratory's MARMOT code and calculate the impact of microstructure heterogeneities on the effective thermal conductivity using mesoscale heat conduction simulations. The thermal conductivities of both samples are higher than the bulk MOX thermal conductivity because of the formation of metallic precipitates and because we do not currently consider phonon scattering due to defects smaller than the experimental resolution. We also used the results to investigate the accuracy of simple thermal conductivity approximations and equations to convert 2D thermal conductivities to 3D. It was found that these approximations struggle to predict the complex thermal transport interactions between metal precipitates and voids.
NASA Technical Reports Server (NTRS)
Miller, R.H.; Morrison, David (Technical Monitor)
1994-01-01
Nuclei of galaxies often show complicated density structures and perplexing kinematic signatures. In the past we have reported numerical experiments indicating a natural tendency for galaxies to show nuclei offset with respect to nearby isophotes and for the nucleus to have a radial velocity different from the galaxy's systemic velocity. Other experiments show normal mode oscillations in galaxies with large amplitudes. These oscillations do not damp appreciably over a Hubble time. The common thread running through all these is that galaxies often show evidence of ringing, bouncing, or sloshing around in unexpected ways, even though they have not been disturbed by any external event. Recent observational evidence shows yet another phenomenon indicating the dynamical complexity of central regions of galaxies: multiple cores (M31, Markarian 315 and 463 for example). These systems can hardly be static. We noted long-lived multiple core systems in galaxies in numerical experiments some years ago, and we have more recently followed up with a series of experiments on multiple core galaxies, starting with two cores. The relevant parameters are the energy in the orbiting clumps, their relative.masses, the (local) strength of the potential well representing the parent galaxy, and the number of cores. We have studied the dependence of the merger rates and the nature of the final merger product on these parameters. Individual cores survive much longer in stronger background potentials. Cores can survive for a substantial fraction of a Hubble time if they travel on reasonable orbits.
NASA Technical Reports Server (NTRS)
Gallagher, Dennis L.; Craven, P. D.; Comfort, R. H.
1999-01-01
Abstract. The Global Core Plasma Model (GCPM) provides, empirically derived, core plasma density as a function of geomagnetic and solar conditions throughout the inner magnetosphere. It is continuous in value and gradient and is composed of separate models for the ionosphere, the plasmasphere, the plasmapause, the trough, and the polar cap. The relative composition of plasmaspheric H+, He+, and O+ is included in the GCPM. A blunt plasmaspheric bulge and rotation of the bulge with changing geomagnetic conditions is included. The GCPM is an amalgam of density models, intended to serve as a framework for continued improvement as new measurements become available and are used to characterize core plasma density, composition, and temperature.
The iron alloys of the Earth's core
NASA Astrophysics Data System (ADS)
Caracas, R.; Verstraete, M. J.; Vargas Calderon, A.; Labrosse, S.; Hernlund, J. W.; Gomi, H.; Ohta, K.; Hirose, K.
2012-12-01
We estimate the necessary amount of several light elements - C, S, P, O, Si - as major alloying components to match the observed seismic properties of the Earth's inner core. For this we compute the elastic constants tensors and determine the seismic properties of Fe3X compounds, with X = C, S, P, O and Si, using first-principles calculations. Assuming linear relations and similar temperature corrections of velocities, we obtain as most reasonable silicon and oxygen. We perform the same exercise on Fe-Ni alloys and see a minor effect of Ni on the seismic properties of iron. We compute the electrical conductivity of iron and iron alloys at Earth's core conditions from electron-phonon coupling in the ABINIT implementation. We find an excellent agreement with experimental results for pure hcp iron below 1 mbars. We confidently use our results up to core pressure conditions. We show that the conductivity exhibits saturation at high pressures. We treat in detail the effect of Si on hcp iron and show a marked saturation effect, an increase in anisotropy and a strong dependence with the substitution pattern. The computed values suggest that the outer core should have conductivities in excess of 90 W/K/m, which is considerably larger than current estimates. This implies an inner core younger than 1 bil. years and stratification of the outer core.
Core radii and common-envelope evolution
NASA Astrophysics Data System (ADS)
Hall, Philip D.; Tout, Christopher A.
2014-11-01
Many classes of objects and events are thought to form in binary star systems after a phase in which a core and companion spiral to smaller separation inside a common envelope (CE). Such a phase can end with the merging of the two stars or with the ejection of the envelope to leave a surviving binary system. The outcome is usually predicted by calculating the separation to which the stars must spiral to eject the envelope, assuming that the ratio of the core-envelope binding energy to the change in orbital energy is equal to a constant efficiency factor ?. If either object would overfill its Roche lobe at this end-of-CE separation, then the stars are assumed to merge. It is unclear what critical radius should be compared to the end-of-CE Roche lobe for stars which have developed cores before the start of a CE phase. After improving the core radius formulae in the widely used BSE rapid evolution code, we compare the properties of populations in which the critical radius is chosen to be the pre-CE core radius or the post-CE stripped remnant radius. Our improvements to the core radius formulae and the uncertainty in the critical radius significantly affect the rates of merging in CE phases of most types. We find the types of systems for which these changes are most important.
Experimental constraints on Mercury's core composition
NASA Astrophysics Data System (ADS)
Chabot, Nancy L.; Wollack, E. Alex; Klima, Rachel L.; Minitti, Michelle E.
2014-03-01
The recent discovery of high S concentrations on the surface of Mercury by spacecraft measurements from the MESSENGER mission provides the potential to place new constraints on the composition of Mercury's large metallic core. In this work, we conducted a set of systematic equilibrium metal-silicate experiments that determined the effect of different metallic compositions in the Fe-S-Si system on the S concentration in the coexisting silicate melt. We find that metallic melts with a range of S and Si combinations can be in equilibrium with silicate melts with S contents consistent with Mercury's surface, but that such silicate melts contain Fe contents lower than measured for Mercury's surface. If Mercury's surface S abundance is representative of the planet's bulk silicate composition and if the planet experienced metal-silicate equilibrium during planetary core formation, then these results place boundaries on the range of possible combinations of Si and S that could be present as the light elements in Mercury's core and suggest that Mercury's core likely contains Si. Except for core compositions with extreme abundances of Si, bulk Mercury compositions calculated by using the newly determined range of potential S and Si core compositions do not resemble primitive meteorite compositions.
The Cancer Genome Atlas (TCGA) Biospecimen Core Resource centralized laboratory reviews and processes blood and tissue samples and their associated data using optimized standard operating procedures for the entire TCGA Research Network.
Core assembly storage structure
Jones, Jr., Charles E. (Northridge, CA); Brunings, Jay E. (Chatsworth, CA)
1988-01-01
A structure for the storage of core assemblies from a liquid metal-cooled nuclear reactor. The structure comprises an enclosed housing having a substantially flat horizontal top plate, a bottom plate and substantially vertical wall members extending therebetween. A plurality of thimble members extend downwardly through the top plate. Each thimble member is closed at its bottom end and has an open end adjacent said top plate. Each thimble member has a length and diameter greater than that of the core assembly to be stored therein. The housing is provided with an inlet duct for the admission of cooling air and an exhaust duct for the discharge of air therefrom, such that when hot core assemblies are placed in the thimbles, the heat generated will by convection cause air to flow from the inlet duct around the thimbles and out the exhaust duct maintaining the core assemblies at a safe temperature without the necessity of auxiliary powered cooling equipment.
Contaminated Sediment Core Profiling
Evaluating the environmental risk of sites containing contaminated sediments often poses major challenges due in part to the absence of detailed information available for a given location. Sediment core profiling is often utilized during preliminary environmental investigations ...
NASA Technical Reports Server (NTRS)
Hultgren, Lennart S.
2012-01-01
This presentation is a technical summary of and outlook for NASA-internal and NASA-sponsored external research on core noise funded by the Fundamental Aeronautics Program Subsonic Fixed Wing (SFW) Project. Sections of the presentation cover: the SFW system-level noise metrics for the 2015 (N+1), 2020 (N+2), and 2025 (N+3) timeframes; SFW strategic thrusts and technical challenges; SFW advanced subsystems that are broadly applicable to N+3 vehicle concepts, with an indication where further noise research is needed; the components of core noise (compressor, combustor and turbine noise) and a rationale for NASA's current emphasis on the combustor-noise component; the increase in the relative importance of core noise due to turbofan design trends; the need to understand and mitigate core-noise sources for high-efficiency small gas generators; and the current research activities in the core-noise area, with additional details given about forthcoming updates to NASA's Aircraft Noise Prediction Program (ANOPP) core-noise prediction capabilities, two NRA efforts (Honeywell International, Phoenix, AZ and University of Illinois at Urbana-Champaign, respectively) to improve the understanding of core-noise sources and noise propagation through the engine core, and an effort to develop oxide/oxide ceramic-matrix-composite (CMC) liners for broadband noise attenuation suitable for turbofan-core application. Core noise must be addressed to ensure that the N+3 noise goals are met. Focused, but long-term, core-noise research is carried out to enable the advanced high-efficiency small gas-generator subsystem, common to several N+3 conceptual designs, needed to meet NASA's technical challenges. Intermediate updates to prediction tools are implemented as the understanding of the source structure and engine-internal propagation effects is improved. The NASA Fundamental Aeronautics Program has the principal objective of overcoming today's national challenges in air transportation. The SFW Quiet-Aircraft Subproject aims to develop concepts and technologies to reduce perceived community noise attributable to aircraft with minimal impact on weight and performance. This reduction of aircraft noise is critical to enabling the anticipated large increase in future air traffic.
NASA Technical Reports Server (NTRS)
Collins, David; Brooks, Marshall; Chen, Paul; Dwelle, Paul; Fischer, Ben
1989-01-01
A micro-coring apparatus for lunar exploration applications, that is compatible with the other components of the Walking Mobile Platform, was designed. The primary purpose of core sampling is to gain an understanding of the geological composition and properties of the prescribed environment. This procedure has been used extensively for Earth studies and in limited applications during lunar explorations. The corer is described and analyzed for effectiveness.
Damaged TMI-2 core consolidated region crust behavior
Kuan, P.
1988-01-01
This paper describes the thermal analyses that were performed in the study of the behavior of crusts around the consolidated region of the damaged TMI-2 core. The heat conduction equation was solved to determine: (1) the equilibrium thickness of the crucible-like lower crust during the initial stage of core damage, (2) the growth of the top crust as water was pumped through the core, and (3) the subsequent erosion of the lower and upper crusts by the enclosed molten material. The calculated minimum thickness of the lower crust (between 25 and 130 mm) compares well with the thickness of samples of the lower crust. The top crust was calculated to reach an equilibrium thickness of less than 20 mm at about 20 min after core cooling. The major core relocation at 224 min into the accident could have been initiated by the failure of such a thin top crust. 10 refs., 8 figs., 1 tab.
Schenewerk, William E. (Sherman Oaks, CA); Glasgow, Lyle E. (Westlake Village, CA)
1983-01-01
A liquid metal cooled fast breeder reactor provided with an emergency core cooling system includes a reactor vessel which contains a reactor core comprising an array of fuel assemblies and a plurality of blanket assemblies. The reactor core is immersed in a pool of liquid metal coolant. The reactor also includes a primary coolant system comprising a pump and conduits for circulating liquid metal coolant to the reactor core and through the fuel and blanket assemblies of the core. A converging-diverging venturi nozzle with an intermediate throat section is provided in between the assemblies and the pump. The intermediate throat section of the nozzle is provided with at least one opening which is in fluid communication with the pool of liquid sodium. In normal operation, coolant flows from the pump through the nozzle to the assemblies with very little fluid flowing through the opening in the throat. However, when the pump is not running, residual heat in the core causes fluid from the pool to flow through the opening in the throat of the nozzle and outwardly through the nozzle to the assemblies, thus providing a means of removing decay heat.
NASA Astrophysics Data System (ADS)
Suwardi
2010-06-01
In recent years, a novel class of zirconium alloys having the melting temperature of 990-1160 K has been developed. Based on novel zirconium matrix alloys, high uranium content fuel pin with U-9Mo has been developed according to capillary impregnation technique. The pin shows it is thermal conductivity ranging from 18 to 22 w/m/K that is comparably higher than UO2 pellet pin. The paper presents the met-met fabrication and thermal performance analysis of the fuel in typical PWR. The fabrication consists of mixing UO2 powder or granules and a novel Zr-alloy powder having low melting point, filling the mixture in a cladding tube that one of its end has been plugged, heating the pin to above melting temperature of Zr-alloy for an hour, natural cooling and heat treating at 300 K for 12 hr. The thermal analysis takes into account the pore and temperature distribution and high burn up effect to pellet conductivity. The thermal diffusivity ratio of novel to conventional fuel has been used as correction factor for the novel fuel conductivity. The results show a significant lowering pellet temperature along the radius until 1000 K at the hottest position. The analysis underestimates since the gap conductivity has been treated as decreased by 2% fission gas released that is not real since the use of lower temperature, and also decreasing thermal conductivity by porosity formation will much lower. The analysis shows that the novel fuel has very good thermal properties which able to pass the barrier of 65 MWD/kg-U, the limit to day commercial fuel. The burn-up extension means fewer fresh fuel is needed to produce electricity, preserve natural uranium resource, easier fuel handling operational per energy produced
Industrial Technology Core (IT Core) Guide
NSDL National Science Digital Library
This resource, created by the South Carolina Advanced Technological Education (SC ATE) National Resource Center, introduces students to core projects of industrial technology. The lesson involves five different activities, the topics include: an introduction to technology careers, basic hand tools, mechanical advantage, basic electricity and hydraulic systems. A suggested equipment list, instructors notes, and objectives are included to guide instructors in preparing these lessons plans. Each one of these topics includes a worksheet for students to actively participate in these lessons. This is a comprehensive set of lessons to help students better understand the different elements in industrial technology.
Engineering Technology Core (ET Core) Guide
NSDL National Science Digital Library
"The ET Core is designed to prepare students for the study of courses specific to any engineering technology major. The curriculum provides hands-on work with technology and workplace relevance as students complete their study of physics, communications, and mathematics (through introductory calculus)." In this 140-page PDF, visitors will find an introduction to the course, the competencies it covers, equipment needed, and detailed instructions for all sixteen modules. The modules cover all sorts of engineering technology including Electrical, Thermal, Mechanical, Fluids, Optics, and Materials. Each module also contains any students handouts necessary to teach it.
NASA Technical Reports Server (NTRS)
Hultgren, Lennart S.
2010-01-01
This presentation is a technical progress report and near-term outlook for NASA-internal and NASA-sponsored external work on core (combustor and turbine) noise funded by the Fundamental Aeronautics Program Subsonic Fixed Wing (SFW) Project. Sections of the presentation cover: the SFW system level noise metrics for the 2015, 2020, and 2025 timeframes; the emerging importance of core noise and its relevance to the SFW Reduced-Noise-Aircraft Technical Challenge; the current research activities in the core-noise area, with some additional details given about the development of a high-fidelity combustion-noise prediction capability; the need for a core-noise diagnostic capability to generate benchmark data for validation of both high-fidelity work and improved models, as well as testing of future noise-reduction technologies; relevant existing core-noise tests using real engines and auxiliary power units; and examples of possible scenarios for a future diagnostic facility. The NASA Fundamental Aeronautics Program has the principal objective of overcoming today's national challenges in air transportation. The SFW Reduced-Noise-Aircraft Technical Challenge aims to enable concepts and technologies to dramatically reduce the perceived aircraft noise outside of airport boundaries. This reduction of aircraft noise is critical for enabling the anticipated large increase in future air traffic. Noise generated in the jet engine core, by sources such as the compressor, combustor, and turbine, can be a significant contribution to the overall noise signature at low-power conditions, typical of approach flight. At high engine power during takeoff, jet and fan noise have traditionally dominated over core noise. However, current design trends and expected technological advances in engine-cycle design as well as noise-reduction methods are likely to reduce non-core noise even at engine-power points higher than approach. In addition, future low-emission combustor designs could increase the combustion-noise component. The trend towards high-power-density cores also means that the noise generated in the low-pressure turbine will likely increase. Consequently, the combined result from these emerging changes will be to elevate the overall importance of turbomachinery core noise, which will need to be addressed in order to meet future noise goals.
Mechanisms and Geochemical Models of Core Formation
Rubie, David C
2015-01-01
The formation of the Earth's core is a consequence of planetary accretion and processes in the Earth's interior. The mechanical process of planetary differentiation is likely to occur in large, if not global, magma oceans created by the collisions of planetary embryos. Metal-silicate segregation in magma oceans occurs rapidly and efficiently unlike grain scale percolation according to laboratory experiments and calculations. Geochemical models of the core formation process as planetary accretion proceeds are becoming increasingly realistic. Single stage and continuous core formation models have evolved into multi-stage models that are couple to the output of dynamical models of the giant impact phase of planet formation. The models that are most successful in matching the chemical composition of the Earth's mantle, based on experimentally-derived element partition coefficients, show that the temperature and pressure of metal-silicate equilibration must increase as a function of time and mass accreted and so m...
Core Transitions in the Breakup of Exotic Nuclei
N. C. Summers; F. M. Nunes; I. J. Thompson
2006-02-10
An interesting physical process has been unveiled: dynamical core excitation during a breakup reaction of loosely bound $core+N$ systems. These reactions are typically used to extract spectroscopic information and/or astrophysical information. A new method, the eXtended Continuum Discretized Coupled Channel (XCDCC) method, was developed to incorporate, in a consistent way and to all orders, core excitation in the bound and scattering states of the projectile, as well as dynamical excitation of the core as it interacts with the target. The model predicts cross sections to specific states of the core. It is applied to the breakup of $^{11}$Be on $^9$Be at 60 MeV/u, and the calculated cross sections are in improved agreement with the data. The distribution of the cross section amongst the various core states is shown to depend on the reaction model used, and not simply on the ground state spectroscopic factors.
Protostellar formation in rotating interstellar clouds. VIII - Inner core formation
NASA Astrophysics Data System (ADS)
Boss, Alan P.
1989-11-01
The results are presented of a variety of spherically symmetric one-dimensional (1D) calculations intended to determine the robustness of the dynamical hiccup phenomenon in protostellar cores. The 1D models show that the phenomenon is relatively insensitive to changes in the equations of state, numerical resolution, initial density and temperature, and the radiative transfer approximation. In 1D, the hiccup results in an explosive destruction of the entire inner protostellar core. Inner core formation is studied with a sequence of three-dimensional models which show that rapid inner core rotation stabilizes the hiccup instability. Instead, the inner core becomes quite flat and undergoes a cycle of binary fragmentation, binary decay into a single object surrounded by a bar, breakup of the bar into a binary, etc. When lesser amounts of rotation are involved, the inner core does hiccup somewhat, but mass is ejected in only a few directions, leading to several broad streams of ejecta.
Protostellar formation in rotating interstellar clouds. VIII - Inner core formation
NASA Technical Reports Server (NTRS)
Boss, Alan P.
1989-01-01
The results are presented of a variety of spherically symmetric one-dimensional (1D) calculations intended to determine the robustness of the dynamical hiccup phenomenon in protostellar cores. The 1D models show that the phenomenon is relatively insensitive to changes in the equations of state, numerical resolution, initial density and temperature, and the radiative transfer approximation. In 1D, the hiccup results in an explosive destruction of the entire inner protostellar core. Inner core formation is studied with a sequence of three-dimensional models which show that rapid inner core rotation stabilizes the hiccup instability. Instead, the inner core becomes quite flat and undergoes a cycle of binary fragmentation, binary decay into a single object surrounded by a bar, breakup of the bar into a binary, etc. When lesser amounts of rotation are involved, the inner core does hiccup somewhat, but mass is ejected in only a few directions, leading to several broad streams of ejecta.