For comprehensive and current results, perform a real-time search at Science.gov.

1

Review of Axial Burnup Distribution Considerations for Burnup Credit Calculations

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.

Wagner, J.C.; DeHart, M.D.

2000-03-01

2

Detailed Burnup Calculations for Testing Nuclear Data

NASA Astrophysics Data System (ADS)

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.

Leszczynski, F.

2005-05-01

3

Detailed Burnup Calculations for Testing Nuclear Data

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.

Leszczynski, F. [Centro Atomico Bariloche (CNEA), 8400 S.C.de Bariloche (Argentina)

2005-05-24

4

Triton burnup measurements and calculations on TFTR

NASA Astrophysics Data System (ADS)

Measurements of the burnup of fusion product tritons in TFTR are presented. Interpretation of triton burnup experiments requires three accurate components: the measurement of the 2.5 MeV neutron emission, the measurement of the 14 MeV neutron emission and a calculation of the expected burnup ratio from the measured plasma parameters. The absolute calibration for the 14 MeV neutron measurements is provided by an NE213 proton recoil spectrometer. Time dependent burnup measurements for three plasma conditions selected for optimum detector operation are shown. Measurements of the time integrated triton burnup from copper activation foils (cross-calibrated to the NE213 measurements) are presented. Descriptions are provided of the neutron detectors and the plasma diagnostics whose data are used as input to the calculation of the expected burnup. All these measurements find that the triton burnup on TFTR is 1/2 +/- 1/4 the classical expectations for a wide variety of discharges. The burnup decreases for relatively longer triton slowing down times, implying possible fast ion diffusion coefficients of ~0.1 m2/s. Alternatively, burnup appears to decrease with increasing major radius of the triton source and edge safety factor qcyl, implying that ripple losses may be playing a role. Triton burnup is a very sensitive measure of anomalous fast ion transport; similar levels of diffusive transport in an ignited reactor would have minimal impact on the alpha particles.

Barnes, C. W.; Bosch, H.-S.; Hendel, H. W.; Huibers, A. G. A.; Jassby, D. L.; Motley, R. W.; Nieschmidt, E. B.; Saito, T.; Strachan, J. D.; Bitter, M.; Budny, R. V.; Hill, K. W.; Mansfield, D. K.; McCune, D. C.; Nazikian, R.; Park, H. K.; Ramsey, A. T.; Scott, S. D.; Taylor, G.; Zarnstorff, M. C.

1998-04-01

5

Detailed Burnup Calculations for Testing Nuclear Data

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

F. Leszczynski

2005-01-01

6

Triton burnup measurements and calculations on TFTR

Measurements of the burnup of fusion product tritons in TFTR are presented. Interpretation of triton burnup experiments requires three accurate components: the measurement of the 2.5 MeV neutron emission, the measurement of the 14 MeV neutron emission and a calculation of the expected burnup ratio from the measured plasma parameters. The absolute calibration for the 14 MeV neutron measurements is

C. W. Barnes; H.-S. Bosch; H. W. Hendel; A. G. A. Huibers; D. L. Jassby; R. W. Motley; E. B. Nieschmidt; T. Saito; J. D. Strachan; M. Bitter; R. V. Budny; K. W. Hill; D. K. Mansfield; D. C. McCune; R. Nazikian; H. K. Park; A. T. Ramsey; S. D. Scott; G. Taylor; M. C. Zarnstorff

1998-01-01

7

Sensitivity Study of Fuel Cost in Extended Burnup BWR Core

A sensitivity study on the fuel cost of an extended burnup BWR core has been carried out on the basis of a realistic model of discharge burnup extension. Full power operating length in months in a refueling cycle and the number of refueling batches are chosen as independent variables in the model to describe extended burnup cores of various types.

Yasuhiro KOBAYASHI; Kikuo UMEGAKI

1984-01-01

8

Burnup calculation methodology in the serpent 2 Monte Carlo code

This paper presents two topics related to the burnup calculation capabilities in the Serpent 2 Monte Carlo code: advanced time-integration methods and improved memory management, accomplished by the use of different optimization modes. The development of the introduced methods is an important part of re-writing the Serpent source code, carried out for the purpose of extending the burnup calculation capabilities from 2D assembly-level calculations to large 3D reactor-scale problems. The progress is demonstrated by repeating a PWR test case, originally carried out in 2009 for the validation of the newly-implemented burnup calculation routines in Serpent 1. (authors)

Leppaenen, J. [VTT Technical Research Centre of Finland, P.O.Box 1000, FI-02044 VTT (Finland); Isotalo, A. [Aalto Univ., Dept. of Applied Physics, P.O.Box 14100, FI-00076 AALTO (Finland)

2012-07-01

9

MONTE-CARLO BURNUP CALCULATION UNCERTAINTY QUANTIFICATION AND PROPAGATION DETERMINATION

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.

Sternat, M.; Nichols, T.

2011-06-09

10

MONTE-CARLO BURNUP CALCULATION UNCERTAINTY QUANTIFICATION AND PROPAGATION DETERMINATION

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.

Nichols, T.; Sternat, M.; Charlton, W.

2011-05-08

11

Methodology for embedded transport core calculation

NASA Astrophysics Data System (ADS)

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.

Ivanov, Boyan D.

12

Burnup concept for a long-life fast reactor core using MCNPX.

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.

Holschuh, Thomas Vernon,; Lewis, Tom Goslee,; Parma, Edward J.,

2013-02-01

13

Accident source terms for boiling water reactors with high burnup cores.

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.

Gauntt, Randall O.; Powers, Dana Auburn; Leonard, Mark Thomas

2007-11-01

14

The increasing tendency towards fuel lifetime extension in thermal nuclear reactors motivated validation work for available evaluation tools for nuclear fuel burnup calculations. In this study two deterministic codes with different transport solvers and one Monte Carlo method are investigated. The code system KAPROS/KARBUS uses the classical deterministic First Collision Probability method utilizing a cylinderized Wigner-Seitz cell. In the SCALES.0/TRITON/NEWT code the Extended Step Characteristic method is applied. In a first step the two deterministic codes are compared with experimental results from the KWO-Isotope Correlation Experiment up to 30 MWD/kg HM burnup, published in 1981. Two pin cell calculations are analyzed by comparison of calculated and experimental results for important heavy isotope vectors. The results are very satisfactory. Subsequently, further validation at higher burnup (< 80 MWD/kg HM) is provided by comparison of the two deterministic codes and the Monte Carlo based burnup code MONTEBURNS for PWR UO{sub 2} fuel assembly calculations. Possible reasons for differences in the results are analyzed and discussed. Especially the influence of cross section data and processing is presented. (authors)

Oberle, P.; Broeders, C. H. M.; Dagan, R. [Forschungszentrum Karlsruhe, Institut for Reactor Safety, Hermann-von-Helmholtz-Platz-1, 76344 Eggenstein-Leopoldshafen (Germany)

2006-07-01

15

Using Laguerre polynomials to compute the matrix exponential in burnup calculations

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)

She, D.; Zhu, A.; Wang, K. [Dept. of Engineering Physics, Tsinghua Univ., Beijing, 100084 (China)

2012-07-01

16

Calculations on fission gas behaviour in the high burnup structure

The behaviour of fission gas in high burnup fuel during steady-state and transient conditions is of special interest for safety reasons. Despite this, mechanistic models that reflect the fission gas transport processes and reliably predict the evolution of the remaining fission gas in the high burnup structure (HBS) are largely missing today. We start to address this problem by developing

P. Blair; A. Romano; Ch. Hellwig; R. Chawla

2006-01-01

17

Experimental and operational validation of burn-up calculations for the Syrian MNSR

The calculation of the uranium-235 burn-up in the Syrian Miniature Research Reactor was conducted in this paper using the WIMS-D\\/4 and CITATION codes. The uranium-235 burn-up was measured experimentally using the measured photoneutron flux in the Be reflector of MNSR subcritical state as well. Good agreements were obtained between the calculated and measured results. The results of the CITATION code

H. Omar; Kh. Haddad; N. Ghazi; N. Alsomel

2010-01-01

18

Calculations on fission gas behaviour in the high burnup structure

NASA Astrophysics Data System (ADS)

The behaviour of fission gas in high burnup fuel during steady-state and transient conditions is of special interest for safety reasons. Despite this, mechanistic models that reflect the fission gas transport processes and reliably predict the evolution of the remaining fission gas in the high burnup structure (HBS) are largely missing today. We start to address this problem by developing a one-dimensional, mass balance model and apply it to LWR UO 2 fuel at the moderate temperatures found in the rim region. We examine the quantity of gas remaining in the HBS fuel matrix at steady state and compare it with experimental values. We find that the current model reproduces the 0.2 wt% observed xenon concentration under certain conditions, viz. fast grain boundary diffusion and an effective volume diffusion coefficient. A sensitivity analysis is also conducted for the model parameters, the relative importance for which is not well established a priori.

Blair, P.; Romano, A.; Hellwig, Ch.; Chawla, R.

2006-05-01

19

Fuel burnup calculation of Ghana MNSR using ORIGEN2 and REBUS3 codes.

Ghana Research Reactor-1 core is to be converted from HEU fuel to LEU fuel in the near future and managing the spent nuclear fuel is very important. A fuel depletion analysis of the GHARR-1 core was performed using ORIGEN2 and REBUS3 codes to estimate the isotopic inventory at end-of-cycle in order to help in the design of an appropriate spent fuel cask. The results obtained for both codes were consistent for U-235 burnup weight percent and Pu-239 build up as a result of burnup. PMID:23796663

Abrefah, R G; Nyarko, B J B; Fletcher, J J; Akaho, E H K

2013-10-01

20

The Large Core Code Evaluation Working Group Benchmark Problem Four was specified to provide a stringent test of the current methods which are used in the nuclear design and analyses process. The benchmark specifications provided a base for performing detailed burnup calculations over the first two irradiation cycles for a large heterogeneous fast reactor. Particular emphasis was placed on the

C. L. Cowan; R. Protsik; J. W. Lewellen

1984-01-01

21

Criticality reference benchmark calculations for burnup credit using spent fuel isotopics

To date, criticality analyses performed in support of the certification of spent fuel casks in the United States do not take credit for the reactivity reduction that results from burnup. By taking credit for the fuel burnup, commonly referred to as burnup credit,'' the fuel loading capacity of these casks can be increased. One of the difficulties in implementing burnup credit in criticality analyses is that there have been no critical experiments performed with spent fuel which can be used for computer code validation. In lieu of that, a reference problem set of fresh fuel critical experiments which model various conditions typical of light water reactor (LWR) transportation and storage casks has been identified and used in the validation of SCALE-4. This report documents the use of this same problem set to perform spent fuel criticality benchmark calculations by replacing the actual fresh fuel isotopics from the experiments with six different sets of calculated spent fuel isotopics. The SCALE-4 modules SAS2H and CSAS4 were used to perform the analyses. These calculations do not model actual critical experiments. The calculated k-effectives are not supposed to equal unity and will vary depending on the initial enrichment and burnup of the calculated spent fuel isotopics. 12 refs., 11 tabs.

Bowman, S.M.

1991-04-01

22

Spent fuel pool storage calculations using the ISOCRIT burnup credit tool

In order to conservatively apply burnup credit in spent fuel pool criticality safety analyses, Westinghouse has developed a software tool, ISOCRIT, for generating depletion isotopics. This tool is used to create isotopics data based on specific reactor input parameters, such as design basis assembly type; bounding power/burnup profiles; reactor specific moderator temperature profiles; pellet percent theoretical density; burnable absorbers, axial blanket regions, and bounding ppm boron concentration. ISOCRIT generates burnup dependent isotopics using PARAGON; Westinghouse's state-of-the-art and licensed lattice physics code. Generation of isotopics and passing the data to the subsequent 3D KENO calculations are performed in an automated fashion, thus reducing the chance for human error. Furthermore, ISOCRIT provides the means for responding to any customer request regarding re-analysis due to changed parameters (e.g., power uprate, exit temperature changes, etc.) with a quick turnaround.

Kucukboyaci, Vefa [Westinghouse Electric Company, Cranberry Township, PA; Marshall, William BJ J [ORNL

2012-01-01

23

NASA Astrophysics Data System (ADS)

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.

Karpushkin, T. Yu.

2012-12-01

24

OECD/NEA burnup credit calculational criticality benchmark Phase I-B results

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.

DeHart, M.D.; Parks, C.V. [Oak Ridge National Lab., TN (United States); Brady, M.C. [Sandia National Labs., Las Vegas, NV (United States)

1996-06-01

25

Group Constants Generation of the Pseudo Fission Products for Fast Reactor Burnup Calculations

The pseudo fission products for the burnup calculations of the liquid metal fast reactor were generated. The cross-section data and fission product yield data of ENDF/B-VI were used for the pseudo fission product data of U-235, U-238, Pu-239, Pu-240, Pu-241, and Pu-242. The pseudo fission product data can be used with the KAFAX-F22 or -E66, which are the MATXS-format libraries for analyses of the liquid metal fast reactor at KAERI and were distributed through the OECD/NEA. The 80-group MATXS-format libraries of the 172 fission products were generated and the burnup chains for generation of the pseudo fission products were prepared.

Gil, Choong-Sup; Kim, Do Heon; Chang, Jonghwa [Korea Atomic Energy Research Institute, 150 Dukjin-dong, Yusung-gu, Taejon (Korea, Republic of)

2005-05-24

26

Startup of “Candle” burnup in fast reactor from enriched uranium core

A new reactor burnup strategy CANDLE was proposed, where shapes of neutron flux, nuclide densities and power density distributions remain constant but move to an axial direction. Here important points are that the solid fuel is fixed at each position and that any movable burnup reactivity control mechanisms such as control rods are not required.This burnup strategy can derive many

Hiroshi Sekimoto; Seiichi Miyashita

2006-01-01

27

Advanced Fuel Cycle Initiative - Projected Linear Heat Generation Rate and Burnup Calculations

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.

Richard G. Ambrosek; Gray S. Chang; Debbie J. Utterbeck

2005-02-01

28

NASA Astrophysics Data System (ADS)

For all the physical components that comprise a nuclear system there is an uncertainty. Assessing the impact of uncertainties in the simulation of fissionable material systems is essential for a best estimate calculation that has been replacing the conservative model calculations as the computational power increases. The propagation of uncertainty in a simulation using a Monte Carlo code by sampling the input parameters is recent because of the huge computational effort required. In this work a sample space of MCNPX calculations was used to propagate the uncertainty. The sample size was optimized using the Wilks formula for a 95th percentile and a two-sided statistical tolerance interval of 95%. Uncertainties in input parameters of the reactor considered included geometry dimensions and densities. It was showed the capacity of the sampling-based method for burnup when the calculations sample size is optimized and many parameter uncertainties are investigated together, in the same input.

Campolina, Daniel de A. M.; Lima, Claubia P. B.; Veloso, Maria Auxiliadora F.

2014-06-01

29

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

Mohammad Javed Khan; Aslam; Nasir Ahmad

2005-01-01

30

Accuracy considerations for Chebyshev rational approximation method (CRAM) in Burnup calculations

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)

Pusa, M. [VTT Technical Research Centre of Finland, P.O. Box 1000, FI-02044 VTT (Finland)

2013-07-01

31

Spent fuel pool storage calculations using the ISOCRIT burnup credit tool

In order to conservatively apply burnup credit in spent fuel pool criticality safety analyses, Westinghouse has developed a software tool, ISOCRIT, for generating depletion isotopics. This tool is used to create isotopics data based on specific reactor input parameters, such as design basis assembly type; bounding power\\/burnup profiles; reactor specific moderator temperature profiles; pellet percent theoretical density; burnable absorbers, axial

Vefa Kucukboyaci; William BJ J Marshall

2012-01-01

32

The Syrian Miniature Neutron Source Reactor (MNSR), a 30kW, 89.8% HEU fueled (U-Al), went critical in March, 1996. By operating the reactor at nominal power for 2.5h\\/day, the estimated core life is 10years. This paper presents the results of fuel burn-up and depletion analysis of the MNSR fuel lattice using the ORIGEN 2 code. A one-group cross-section data base for

H. Omar; N. Ghazi

2011-01-01

33

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.

Holly R. Trellue

1998-12-01

34

Argonne National Laboratory (ANL) of USA and Kharkov Institute of Physics and Technology (KIPT) of Ukraine have been collaborating on the conceptual design development of an electron accelerator driven subcritical (ADS) facility, using the KIPT electron accelerator. The neutron source of the subcritical assembly is generated from the interaction of 100 KW electron beam with a natural uranium target. The electron beam has a uniform spatial distribution and electron energy in the range of 100 to 200 MeV. The main functions of the subcritical assembly are the production of medical isotopes and the support of the Ukraine nuclear power industry. Neutron physics experiments and material structure analyses are planned using this facility. With the 100 KW electron beam power, the total thermal power of the facility is {approx}375 kW including the fission power of {approx}260 kW. The burnup of the fissile materials and the buildup of fission products reduce continuously the reactivity during the operation, which reduces the neutron flux level and consequently the facility performance. To preserve the neutron flux level during the operation, fuel assemblies should be added after long operating periods to compensate for the lost reactivity. This process requires accurate prediction of the fuel burnup, the decay behavior of the fission produces, and the introduced reactivity from adding fresh fuel assemblies. The recent developments of the Monte Carlo computer codes, the high speed capability of the computer processors, and the parallel computation techniques made it possible to perform three-dimensional detailed burnup simulations. A full detailed three-dimensional geometrical model is used for the burnup simulations with continuous energy nuclear data libraries for the transport calculations and 63-multigroup or one group cross sections libraries for the depletion calculations. Monte Carlo Computer code MCNPX and MCB are utilized for this study. MCNPX transports the electrons and the produced neutrons and photons but the current version of MCNPX doesn't support depletion/burnup calculation of the subcritical system with the generated neutron source from the target. MCB can perform neutron transport and burnup calculation for subcritical system using external neutron source, however it cannot perform electron transport calculations. To solve this problem, a hybrid procedure is developed by coupling these two computer codes. The user tally subroutine of MCNPX is developed and utilized to record the information of the each generated neutron from the photonuclear reactions resulted from the electron beam interactions. MCB reads the recorded information of each generated neutron thorough the user source subroutine. In this way, the neutron source generated by electron reactions could be utilized in MCB calculations, without the need for MCB to transport the electrons. Using the source subroutines, MCB could get the external neutron source, which is prepared by MCNPX, and perform depletion calculation for the driven subcritical facility.

Gohar, Y.; Zhong, Z.; Talamo, A.; Nuclear Engineering Division

2009-06-09

35

Fluence-limited burnup as a function of fast reactor core parameters

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 ...

Kersting, Alyssa (Alyssa Rae)

2011-01-01

36

ORPHEE research reactor: 3D core depletion calculation using Monte-Carlo code TRIPOLI-4®

NASA Astrophysics Data System (ADS)

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.

Damian, F.; Brun, E.

2014-06-01

37

A chemical isotopic analysis of the actinides and fission products of a high-burnup PWR-UO2 fuel with an average burnup of 60.2 MWd\\/kgHM was carried out to accumulate extensive nuclide composition data. Furthermore, computational analysis was performed using the integrated burnup calculation code SWAT. The differences between the amounts obtained by the chemical isotopic analysis and SWAT calculation using JENDL-3.2, JENDL-3.3,

Akihiro SASAHARA; Tetsuo MATSUMURA; Giorgos NICOLAOU; Yoshiaki KIYANAGI

2008-01-01

38

SOURCE OF BURNUP VALUES FOR COMMERCIAL SPENT NUCLEAR FUEL ASSEMBLIES

Waste packages are loaded with commercial spent nuclear fuel (SNF) that satisfies the minimum burnup requirements of a criticality loading curve. The burnup value assigned by the originating nuclear utility to each SNF assembly (assigned burnup) is used to load waste packages in compliance with a criticality loading curve. The burnup provided by a nuclear utility has uncertainties, so conservative calculation methods are used to characterize those uncertainties for incorporation into the criticality loading curves. Procedural safety controls ensure that the correct assembly is loaded into each waste package to prevent a misload that could create a condition affecting the safety margins. Probabilistic analyses show that procedural safety controls can minimize the chance of a misload but can not completely eliminate the possibility. Physical measurements of burnup with instrumentation in the surface facility are not necessary due to the conservative calculation methods used to produce the criticality loading curves. The reactor records assigned burnup of a commercial SNF assembly contains about two percent uncertainty, which is increased to five-percent to ensure conservatism. This five-percent uncertainty is accommodated by adjusting the criticality loading curve. Also, the record keeping methods of nuclear utilities are not uniform and the level of detail required by the NRC has varied over the last several decades. Thus, some SNF assemblies may have assigned burnups that are averages for a batch of assemblies with similar characteristics. Utilities typically have access to more detailed core-follow records that allow the batch average burnup to be changed to an assembly specific burnup. Alternatively, an additional safety margin is incorporated into the criticality loading curve to accommodate SNF assemblies with batch average burnups or greater uncertainties due to the methodology used by the nuclear utility. The utility records provide the assembly identifier, initial {sup 235}U enrichment, and time of discharge from the reactor as well as the assigned burnup, but the distribution. of burnup axially along the assembly length is not provided. The axial burnup profile is maintained within acceptable bounds by the operating conditions of the nuclear reactor and is calculated during preparations to reload a reactor, but the actual burnup profile is not measured. The axial burnup profile is important to the determination of the reactivity of a waste package, so a conservative evaluation of the calculated axial profiles for a large database of SNF has been performed. The product of the axial profile evaluation is a profile that is conservative. Thus, there is no need for physical measurement of the axial profile. The assembly identifier is legible on each SNF assembly and the utility records provide the associated characteristics of the assembly. The conservative methodologies used to determine the criticality loading curve for a waste package provide sufficient margin so that criticality safety is assured for preclosure operations even in the event of a misload. Consideration of misload effects for postclosure time periods is provided by the criticality Features, Events, and Processes (FEPs) analysis. The conservative approaches used to develop and apply the criticality loading curve are thus sufficiently robust that the utility assigned burnup is an adequate source of burnup values, and additional means of verification of assigned burnup through physical measurements are not needed.

BSC

2004-12-01

39

Numerical calculation of modes of oscillation of the Earth's core.

NASA Astrophysics Data System (ADS)

The authors outline theoretical and numerical methods they have developed and implemented for the accurate calculation of core modes, and illustrate the application of their methods with the example of the recent identification (Smylie, 1992) of the translational triplet of the solid inner core in the spectra of superconducting gravimeter data.

Smylie, D. E.; Jiang, X.

40

Burnup simulations of an inert matrix fuel using a two region, multigroup reactor physics model

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)

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

41

Perturbation and sensitivity theory for burnup analysis

Perturbation theory is developed for the nonlinear burnup equations describing the time-dependent behavior of the neutron and nuclide fields in a reactor core. General aspects of adjoint equations for nonliner systems are first discussed and then various approximations to the burnup equations are rigorously derived and their areas for application presented. In particular, the concept of coupled neutron\\/nuclide fields (in

1979-01-01

42

Quasiparticle calculations of surface core-level shifts

NASA Astrophysics Data System (ADS)

We report quasiparticle calculations of chemical shifts of core levels at clean and adsorbate-covered Si surfaces. Core-state excitation energies are given as poles of the electronic one-particle Green's function of the many-electron system. To calculate the Green's function, we employ the GW approximation for evaluating the necessary electronic self-energy operator. The core states whose shifts are addressed in this work are explicitly included in the valence shell. We present results for three different surfaces. For the As:Si(111)-(1×1) and H:Si(111)-(1×1) surfaces, we obtain core-level shifts in good agreement with experimental data. For the clean Si(001) surface a high sensitivity of the chemical shifts on the actual dimer structure of the surface is observed.

Rohlfing, Michael; Krüger, Peter; Pollmann, Johannes

1997-07-01

43

PWR AXIAL BURNUP PROFILE ANALYSIS

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).

J.M. Acaglione

2003-09-17

44

Fuel burn-up fraction in RBMK-1000 reactor

The authors calculate fuel burnup fractions for the four RBMK reactors of the Leningrad plant for both unloaded and loaded fuel and graph the predicted dependence of average burnup for both scenarios on reactor operation time, the distribution function for a steady-state mode of continuous fuel recharging, and a histogram of fuel element distribution with burnup fraction at intervals of

A. P. Eperin; V. S. Romanenko; A. V. Zavyalov; A. V. Krayushkin; Yu. V. Garusov; G. F. Yaroslavtsev; M. V. Shavlov

1987-01-01

45

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).

Berna, G.A.; Beyer, G.A.; Davis, K.L.; Lanning, D.D.

1997-12-01

46

FUEL BURNUP STUDIES FOR A 225 Mwe ADVANCED SODIUM GRAPHITE REACTOR

Reactivity and fuel burnup studies were performed for a 255 Mw(e) sodium-;\\u000a graphite reactor of the advanced calandria core type. This reactor is briefly ;\\u000a described. Initial criticality calculations and flux distributions were ;\\u000a obtained, using two-group theory for enrichments between 2.0 at.% UÂ³Â²âµ and ;\\u000a 4.0 at.% U235. A four-group burnup study was performed for enrichments between ;\\u000a 2.5

1960-01-01

47

TOPICAL REPORT ON ACTINIDE-ONLY BURNUP CREDIT FOR PWR SPENT NUCLEAR FUEL PACKAGES

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.

DOE

1997-04-01

48

Effect of Fuel Fraction on Small Modified CANDLE Burn-up Based Gas Cooled Fast Reactors

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.

Ariani, Menik [Departmen of Physics Bandung Institute of Technology, Jl. Ganesha 10, Bandung 40134 (Indonesia); Physics Department, Sriwijaya University, Kampus Indralaya, Ogan Ilir, Sumatera Selatan (Indonesia); Su'ud, Zaki; Waris, Abdul; Asiah, Nur [Departmen of Physics Bandung Institute of Technology, Jl. Ganesha 10, Bandung 40134 (Indonesia); Shafii, M. Ali [Departmen of Physics Bandung Institute of Technology, Jl. Ganesha 10, Bandung 40134 (Indonesia); Physics Department, Andalas University, Kampus Limau Manis, Padang, Sumatera Barat (Indonesia); Khairurrijal

2010-12-23

49

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.

Parish, T.A.

1995-03-02

50

BURNUP IN A SUBCRITICAL SYSTEM WITH FLAT POWER DENSITY

Burnup characteristics of a sub-critical system devoted to the incineration of LWR spent fuel have been studied. Depleted uranium and burnable absorbers were introduced into the core in order to minimize power peaking and reactivity loss during a burnup period of 300 days. In addition, the burnable absorber served the purpose of blocking unwanted neutron captures in even neutron number

Kamil Tu; Jan Wallenius; Waclaw Gudowski; Charlotta Sanders

51

Perturbation and sensitivity theory for reactor burnup analysis

Perturbation theory is developed for the nonlinear burnup equations describing the time-dependent behavior of the neutron and nuclide fields in a reactor core. General aspects of adjoint equations for nonlinear systems are first discussed and then various approximations to the burnup equations are rigorously derived and their areas for application presented. In particular, the concept of coupled neutron\\/nuclide fields (in

1979-01-01

52

Fuel-Cycle of 'CANDLE' Burnup with Depleted Uranium

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.

Hiroshi; Sekimoto

2006-01-01

53

Preliminary Study of Burnup Characteristics for a Simplified Small Pebble Bed Reactor

NASA Astrophysics Data System (ADS)

Simplification of the pebble bed reactor by removing the unloading device from the system was peformed. For this reactor design, a suitable fuel-loading scheme is the Peu à Peu (little by little) fueling scheme. In the Peu à Peu modus, there is no unloading device; as such, the fuels are never discharged and remain at the bottom of the core during reactor operation. This means that the burnup cycle and reactivity is controlled by the addition of fuel. The objectives of the the present study were to find a means of carrying out the exact calculations needed to analyze the Peu à Peu fuel-loading scheme and to optimize the fuel composition, and fuel-loading scheme to achieve better burnup characteristics. The Monte Carlo method is used to perform calculations with high accuracy. Before the calculation of the whole core, the analysis for the infinite geometry was performed. The power generated per mass consumed for each combination of the uranium enrichment and packing fraction was analyzed from the parametric survey. By using the optimal value obtained, a whole-core calculation for the small 20 MWth reactor was performed and the criticality and burnup of this design was analyzed.

Irwanto, Dwi; Kato, Yukikata; Yamanaka, Ichiro; Obara, Toru

2010-06-01

54

Fuel burn-up fraction in RBMK-1000 reactor

The authors calculate fuel burnup fractions for the four RBMK reactors of the Leningrad plant for both unloaded and loaded fuel and graph the predicted dependence of average burnup for both scenarios on reactor operation time, the distribution function for a steady-state mode of continuous fuel recharging, and a histogram of fuel element distribution with burnup fraction at intervals of 100 MW per day for each of the four reactors.

Eperin, A.P.; Romanenko, V.S.; Zav'yalov, A.V.; Krayushkin, A.V.; Garusov, Yu.V.; Yaroslavtsev, G.F.; Shavlov, M.V.

1987-03-01

55

The 5-MW highly enriched uranium (HEU)-fueled Tehran Research Reactor is considered for conversion to high-density, low-enriched uranium (LEU) fuel. A preliminary neutronics calculation is performed as part of the conversion goal. In this study, two cores are considered: the HEU reference core and a proposed LEU core similar to the reference core, and a proposed LEU core similar to the

Nejat; S. M. R

1993-01-01

56

NASA Astrophysics Data System (ADS)

Minimum offset of 7 km across the Pinaleño 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.

Naruk, Stephen J.

1987-07-01

57

NASA Astrophysics Data System (ADS)

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.

Sloma, Tanya Noel

58

Nuclide analysis in high burnup fuel samples irradiated in Vandellós 2

NASA Astrophysics Data System (ADS)

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.

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

59

Approximate Calculation Method for Second Order Sensitivity Coefficient

A simple method has been developed for calculating the second order sensitivity coefficient of static and burnup-dependent core performance parameters. The method is applied to a small and a large fast breeder reactors. Changes in core performance parameters due to 10% cross section changes are compared with that predicted by the first and the second order sensitivity analyses. Numerical results

Kazuhisa MATSUMOTO; Toshikazu TAKEDA; Tomoaki MASUDA

1994-01-01

60

New Soft-Core Potential Function for Molecular Dynamics Based Alchemical Free Energy Calculations

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

de Groot, Bert

61

A 300 kW small medical reactor was designed to be used for boron neutron capture therapy (BNCT) at KAIST in 1996. In this paper, analysis for the core life cycle of the original design of the BNCT facility and modifications of the fuel assembly configuration and enrichment to get a proper life cycle were performed and a criticality, neutron flux distribution and fuel burnup calculations were carried out.

Sambuu, Odmaa; Nanzad, Norov [Nuclear Research Center National University of Mongolia Ulaanbaatar (Mongolia)

2009-03-31

62

NASA Astrophysics Data System (ADS)

A 300 kW small medical reactor was designed to be used for boron neutron capture therapy (BNCT) at KAIST in 1996 [1]. In this paper, analysis for the core life cycle of the original design of the BNCT facility and modifications of the fuel assembly configuration and enrichment to get a proper life cycle were performed and a criticality, neutron flux distribution and fuel burnup calculations were carried out.

Sambuu, Odmaa; Nanzad, Norov

2009-03-01

63

NASA Astrophysics Data System (ADS)

A detailed investigation of the atomic structure and radiative parameters involving the lowest states within the 6p4, 6p36d, 6p37s, 6p37p and 6p37d configurations of neutral polonium is reported in the present paper. Using different physical models based on the pseudo-relativistic Hartree-Fock approach, the influence of intravalence, core-valence and core-core electron correlation on the atomic parameters is discussed in detail. This work allowed us to fix the spectroscopic designation of some experimental level energy values and to provide for the first time a set of reliable oscillator strengths corresponding to 31 Po I spectral lines in the wavelength region from 175 to 987 nm.

Quinet, Pascal

2014-09-01

64

Molecular Evolution of A First Core in 3 Dimensional Hydrodynamic Calculations

NASA Astrophysics Data System (ADS)

It is well established that stars are formed by gravitational collapse of molecular cloud cores. Collapsing cores initially undergo isothermal collapse. The isothermal condition breaks down at the density of ˜ 10-13 g cm-3, and the temperature starts rising. Increasing gas pressure decelerates the contraction, and the cores come to hydrostatic equilibrium with a radius of a few AU and a mass of ˜ 0.01 M?, which is called the first cores (e.g. Larson 1969). Observation of the first cores is important but challenging, since their lifetime is short (˜ 1000 yr). The mechanical property of the first cores have been studied by multi-dimensional hydrodynamic calculations considering interstellar magnetic fields and radiative transfer (e.g. Tomisaka 2002; Machida et al.2008; Tomida et al. 2010). In contrast, their chemical property is yet to be understood. It is important to reveal their chemical property in terms of which lines we should use to observe the first cores. In addition, the first cores evolve to protoplanetary disks (Saigo et al. 2008; Machida et al. 2010), hence the compositions of the first cores restrict the initial compositions of disks. We investigate molecular evolution of star forming cores that are initially rotating molecular cloud cores and collapse to form the first cores. The results of three dimensional hydrodynamic calculations (Matsumoto & Hanawa 2003) are adopted as physical models of the core. We trace trajectories of test particles in the hydrodynamic calculations, and molecular evolution is solved using low temperature chemical network (Garrod & Herbst 2006) at T < 100 K and high temperature network (Harada et al. 2010) at T > 100 K along the trajectories. We also consider three body reactions and collisional dissociations (Willacy et al. 1998). Trace particles fall into the first core almost spherically, and rotate in the first core where the spiral arms transports angular momentum. In our model with barotropic approximation, we find that in outer regions (R > 5 AU), the composition is similar to the low temperature chemistry. In intermediate regions (R ˜3 AU), hot-core like species, such as HCOOCH_3 and CH_3OCH_3 are generated. In central regions (R < 1 AU), complex molecules, such as HC_7N, HC_9N and NH_2CN, are formed in the gas phase.

Furuya, K.; Aikawa, Y.; Matsumoto, T.; Tomida, K.; Saigo, K.; Tomisaka, K.; Hersant, F.; Wakelam, V.

2011-05-01

65

Explicit Calculation of the Current Rating of Conductor-Cooled Single-Core Power Cables

The known basic mathematical principles and calculation formulae for cable installations with forced cooling are extended for conductor-cooled single-core power cables to give explicit specificationi of the current rating. The thermal coupling of the cables with each other and the ambient area enter into the calculation.

G. Mainka

1980-01-01

66

Space-time kinetics calculations have been done for transient fuel behavior tests that were conducted during the late 1960's in the Capsule Driver Core in PBF. The purpose of the calculations was to determine the amount of energy deposited by delayed-neutron-fission in the test fuel after the power burst was terminated by control rod insertion. The kinetics calculations were done in

A. J. Scott; D. W. Nigg; J. L. Judd; S. A. Easson

1981-01-01

67

Non-Invasive Characterization of Burnup for PWR Spent Fuel Rods with Burnups > 80 GWd\\/t

High-resolution gamma spectroscopy has been employed for the measurement of Â¹Â³â´Cs\\/Â¹Â³â·Cs, Â¹âµâ´Eu\\/Â¹Â³â·Cs and Â¹Â³â´Cs\\/Â¹âµâ´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 UOâ 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

S. Caruso; M. Murphy; F. Jatuff; R. Chawla

2006-01-01

68

Performance of relativistic effective core potentials in DFT calculations on actinide compounds.

Density functional theory (DFT) calculations using relativistic effective core potentials (RECPs) have emerged as a robust and fast method of calculating the structural parameters and energy changes of the thermochemical reactions of actinide complexes. A comparative investigation of the performance of the Stuttgart small-core and large-core RECPs in DFT calculations has been carried out. The vibrational frequencies and reaction enthalpy changes of several uranium(VI) compounds computed using these RECPs were compared to those obtained using DFT and a four-component one-electron scalar relativistic approximation of the full Dirac equation with large all-electron basis sets (AE). The relativistic AE method is a full solution of the Dirac equation with all spin components separated out. This method gives the "correct" answer (with respect to scalar relativity) which should be closest to experimental values when an adequate density functional is used and in the absence of significant spin-orbit effects. The small-core RECP always show better agreement with the four-component scalar- relativistic AE method than the large-core RECP. We conclude that the 5s, 5p, and 5d orbitals are of great importance in determining the chemistry of actinide complexes. Instances in which large-core RECPs give better agreement with experimental data are attributed to either experimental uncertainties or error cancellations. PMID:20039716

Odoh, Samuel O; Schreckenbach, Georg

2010-02-01

69

Burnup studies of spent fuels of varying types and enrichment

This paper describes the results of fuel burnup measurements, made over a period of several years on discharged fuel from nuclear power plant and research reactor. The measured and calculated burnup of different spent fuel types, viz.: Natural uranium CANDU fuel bundles; 10.5% enriched booster rods; 20% enriched MTR fuel elements have been presented. High-resolution gamma spectrometry, using 137Cs and

S. A. Ansari; M. Asif; T. Rashid; K. G. Qasim

2007-01-01

70

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)

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

71

NASA Astrophysics Data System (ADS)

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.

Hartini, Entin; Andiwijayakusuma, Dinan

2014-09-01

72

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.

J. W. Sterbentz

1999-08-01

73

Faddeev-type calculation of three-body nuclear reactions including core excitation

NASA Astrophysics Data System (ADS)

The core excitation, being an important reaction mechanism, so far is not properly included in most calculations of three-body nuclear reactions. We aim to include the excitation of the core nucleus using an exact Faddeev-type framework for nuclear reactions in the three-body (core+neutron+proton) system. We employ Alt, Grassberger, and Sandhas (AGS) integral equations for the three-particle transition operators and solve them in the momentum-space framework. The Coulomb interaction is included via the method of screening and renormalization. We calculate elastic, inelastic, and transfer reactions involving 10Be and 24Mg nuclear cores. Important effects of the core excitation are found, often improving the description of the experimental data. In the neutron transfer reactions the core excitation effect is by far not just a simple reduction of the cross section by the respective spectroscopic factor. This indicates that widely used extraction of the spectroscopic factors from the ratio of the experimental and theoretical transfer cross sections is an unreliable approach.

Deltuva, A.

2013-07-01

74

Two-level algorithm for efficient space-time reactor core calculations

In order to perform more sophisticated transient analyses, Siemens has coupled the nodal core simulator PANBOX2 with the plant analysis code RELAP5\\/MOD2. The coupling replaces the point-kinetics approximation, which is used in the RELAP5 model with the transient three-dimensional neutron diffusion calculations of PANBOX2. This coupling produces more accurate results, but calculation times become very long due to the complexity

C. Jackson; H. Finnemann; D. Cacuci; R. Boeer

1994-01-01

75

Electronic Structure Calculations and Adaptation Scheme in Multi-core Computing Environments

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.

Seshagiri, Lakshminarasimhan; Sosonkina, Masha; Zhang, Zhao

2009-05-20

76

Neutronic calculations for the conversion to LEU of a research reactor core

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)

Varvayanni, M.; Catsaros, N.; Stakakis, E. [National Center for Scientific Research 'DEMOKRITOS', 153 10 Aghia Paraskevi (Greece); Grigoriadis, D. [National Center for Scientific Research 'DEMOKRITOS', 153 10 Aghia Paraskevi (Greece); Department of Mechanical and Manufacturing Engineering, University of Cyprus, P.O. Box 20537, Nicosia 1678 (Cyprus)

2008-07-15

77

NASA Technical Reports Server (NTRS)

Based on the Binary-Encounter-Bethe (BEB) model, the advantage of using relativistic effective core potentials (RECP) in the calculation of total ionization cross sections of heavy atoms or molecules containing heavy atoms is discussed. Numerical examples for Ar, Kr, Xe, and WF6 are presented.

Huo, Winifred M.; Kim, Yong-Ki

1999-01-01

78

A unified approach for calculating the core melt frequency of a specific reactor caused by both internal and external accident initiators is demonstrated. Two classes of internal initiators are examined: transients, of which turbine trip is the chosen example; and loss-of-coolant events of various sizes. The concepts of hazard and fragility analysis first proposed for seismic risk analysis are linked

Carolyn D. Heising; Virgilio Lopes Oliveira

1995-01-01

79

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)

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

80

A Variational Principle for the Calculation of Core Modes Directly in the Displacement Field

NASA Astrophysics Data System (ADS)

We present a new functional for the variational calculation of core modes expressed directly in the vector displacement field. The Earth's outer core is a rotating, compressible, self-gravitating, stratified fluid, contained within the elastic spherical boundaries of the shell and inner core. The calculation of its long period modes of oscillation is a challenging problem first solved by Smylie et al (1992) using a variational principle based on the scalar generalized potential. While the accurate computations this formulation affords led to the detection of the three translational modes of oscillation of the inner core in the spectra of superconducting gravimeter observations (Smylie, Francis and Merriam, 2001), it is a complicated formulation and leads to a lambda matrix problem of degree eight, even in the case of small non-neutral stratification. The functional presented here gives the displacement field directly and leads to only a quadratic eigenvalue problem for arbitrary non-neutral stratification. It is easily converted to a linear problem of twice the dimension opening the possibility of computing all eigenvalues and eigenvectors below a given degree of spatial complexity, and the construction of a catalogue of core modes.

Ma, H.; Smylie, D. E.; de Viron, O.

2001-12-01

81

Calculation of ex-core physical quantities using the 3D importance functions

NASA Astrophysics Data System (ADS)

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.

Trakas, Christos; De Laubiere, Xavier

2014-06-01

82

3D Neutron Transport PWR Full-core Calculation with RMC code

NASA Astrophysics Data System (ADS)

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.

Qiu, Yishu; She, Ding; Fan, Xiao; Wang, Kan; Li, Zeguang; Liang, Jingang; Leroyer, Hadrien

2014-06-01

83

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)

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

84

NASA Astrophysics Data System (ADS)

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.

Antsiferova, E. V.; Bogdanov, V. V.; Derebenko, E. V.; Lagutina, A. V.; Khmelnikov, E. A.

2006-08-01

85

TORT/MCNP coupling method for the calculation of neutron flux around a core of BWR.

For the analysis of BWR neutronics performance, accurate data are required for neutron flux distribution over the In-Reactor Pressure Vessel equipments taking into account the detailed geometrical arrangement. The TORT code can calculate neutron flux around a core of BWR in a three-dimensional geometry model, but has difficulties in fine geometrical modelling and lacks huge computer resource. On the other hand, the MCNP code enables the calculation of the neutron flux with a detailed geometry model, but requires very long sampling time to give enough number of particles. Therefore, a TORT/MCNP coupling method has been developed to eliminate the two problems mentioned above in each code. In this method, the TORT code calculates angular flux distribution on the core surface and the MCNP code calculates neutron spectrum at the points of interest using the flux distribution. The coupling method will be used as the DOT-DOMINO-MORSE code system. This TORT/MCNP coupling method was applied to calculate the neutron flux at points where induced radioactivity data were measured for 54Mn and 60Co and the radioactivity calculations based on the neutron flux obtained from the above method were compared with the measured data. PMID:16604689

Kurosawa, Masahiko

2005-01-01

86

A multi-platform linking code for fuel burnup and radiotoxicity analysis

NASA Astrophysics Data System (ADS)

A linking code between ORIGEN2.1 and MCNP has been developed at the Departamento de Engenharia Nuclear/UFMG to calculate coupled neutronic/isotopic results for nuclear systems and to produce a large number of criticality, burnup and radiotoxicity results. In its previous version, it evaluated the isotopic composition evolution in a Heat Pipe Power System model as well as the radiotoxicity and radioactivity during lifetime cycles. In the new version, the code presents features such as multi-platform execution and automatic results analysis. Improvements made in the code allow it to perform simulations in a simpler and faster way without compromising accuracy. Initially, the code generates a new input for MCNP based on the decisions of the user. After that, MCNP is run and data, such as recoverable energy per prompt fission neutron, reaction rates and keff, are automatically extracted from the output and used to calculate neutron flux and cross sections. These data are then used to construct new ORIGEN inputs, one for each cell in the core. Each new input is run on ORIGEN and generates outputs that represent the complete isotopic composition of the core on that time step. The results show good agreement between GB (Coupled Neutronic/Isotopic code) and Monteburns (Automated, Multi-Step Monte Carlo Burnup Code System), developed by the Los Alamos National Laboratory.

Cunha, R.; Pereira, C.; Veloso, M. A. F.; Cardoso, F.; Costa, A. L.

2014-02-01

87

NASA Astrophysics Data System (ADS)

The ability of liquid iron to transport heat and electric charge by conduction at extreme pressure and temperature is of paramount importance to the thermal history of the core. Thermal conductivity determines the amount of heat conducted along the core adiabat, i.e. heat not available for generation of the magnetic field, and also strongly controls the time required for the inner core to reach its current size. Electrical conductivity sets the rate of magnetic field dissipation, and consequently the amount of energy required to sustain the geodynamo. Also, because these properties tightly control the heat budget within the core, they dictate the extent to which radiogenic heat need to be invoked to obtain thermal history models that are in agreement with geophysical and paleomagnetic observations. Current estimates for electrical conductivity of iron at conditions characteristic of Earth's core are rather uncertain, constraining the value only to within a factor of three. Thermal conductivity values are subsequently obtained by applying the Wiedemann-Franz relation, the validity of which has not been rigorously shown at extreme pressures. In addition, electronic transport properties are expected to depend strongly on pressure (P) and temperature (T), as well as on the concentration (X) of light elements in the liquid metal. However, with no data available on these variations, geophysical studies in which these values are applied invariably assume them to be constant. In an effort to improve our understanding of the P-T-X behavior of electronic transport properties in the core, and also to test the various assumptions made in their determination, we have performed first-principles calculations of the electrical and thermal conductivity of liquid iron over a large range of pressure and temperature conditions, including those characteristic of Earth's core. Compositions respectively doped with silicon, oxygen and sulphur are also considered. These calculations involve using first-principles molecular dynamics to generate a series of uncorrelated liquid structures at constant temperature and density, for which the electronic transport properties are then computed using the Kubo-Greenwood equation. Our aim is to construct a parameterized model for the thermal and electrical conductivity of liquid iron as a function of pressure, temperature and light element composition, which can be applied in geodynamo simulations and thermal history models for planetary cores. Preliminary results indicate a strong pressure and temperature dependence, with the Wiedemann-Franz relation only approximately satisfied. Implications of these results for models of the thermal history of the core will be considered and discussed.

de Koker, N.; Steinle-Neumann, G.; Vl?ek, V.

2010-12-01

88

Liquid iron-sulfur alloys at outer core conditions by first-principles calculations

NASA Astrophysics Data System (ADS)

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.

Umemoto, Koichiro; Hirose, Kei; Imada, Saori; Nakajima, Yoichi; Komabayashi, Tetsuya; Tsutsui, Satoshi; Baron, Alfred Q. R.

2014-10-01

89

NASA Astrophysics Data System (ADS)

We present results from three-dimensional, self-gravitating radiation hydrodynamical models of gas accretion by planetary cores. In some cases, the accretion flow is resolved down to the surface of the solid core - the first time such simulations have been performed. We investigate the dependence of the gas accretion rate upon the planetary core mass, and the surface density and opacity of the encompassing protoplanetary disc. Accretion of planetesimals is neglected. We find that high-mass protoplanets are surrounded by thick circumplanetary discs during their gas accretion phase but, contrary to locally isothermal calculations, discs do not form around accreting protoplanets with masses <~ when radiation hydrodynamical simulations are performed, even if the grain opacity is reduced from interstellar values by a factor of 100. We find that the opacity of the gas plays a large role in determining the accretion rates for low-mass planetary cores. For example, reducing the opacities from interstellar values by a factor of 100 leads to roughly an order of magnitude increase in the accretion rates for protoplanets. The dependence on opacity becomes less important in determining the accretion rate for more massive cores where gravity dominates the effects of thermal support and the protoplanet is essentially accreting at the runaway rate. Increasing the core mass from 10 to 100 M increases the accretion rate by a factor of ~50 for interstellar opacities. Beyond , the ability of the protoplanetary disc to supply material to the accreting protoplanet limits the accretion rate, independent of the opacity. Finally, for low-mass planetary cores (<~), we obtain accretion rates that are in agreement with previous one-dimensional quasi-static models. This indicates that three-dimensional hydrodynamical effects may not significantly alter the gas accretion time-scales that have been obtained from quasi-static models.

Ayliffe, Ben A.; Bate, Matthew R.

2009-02-01

90

NASA Astrophysics Data System (ADS)

An analytical method for calculating the magnetostatic field of a pulse transformer with open magnetic cores is put forward in this paper, and formulas for calculating inductances of a small aspect-ratio transformer are derived. In comparison to results calculated by finite element magnetostatic-field simulations, the calculated values of inductance of primary winding L1 and the inductance of secondary winding L2 have a relative error of about 5%, while the error of the coupling coefficient (k) is less than 2%. Meanwhile, the effect of current nonuniformity in the primary winding on magnetizing inductance is studied. According to the calculated results, this effect reduces the magnetizing inductance and the coupling coefficient of the transformer, and can lead to an overvoltage phenomenon on the secondary winding. A small aspect-ratio pulse transformer with open magnetic cores is developed, which has a small size of 250mm×150mm in length and diameter, respectively. Inductances of the transformer are measured. The measured results conform to the law obtained in this work. Tests of the pulsed transformer are carried out. Experimental results show that the transformer can export a high-voltage pulse with an amplitude of 310 kV and full width at half maximum of 1?s.

Yu, Bin-xiong; Liu, Jin-liang

2013-01-01

91

Sensitivity of ex-core neutron detectors to vibrations of PWR fuel assemblies

The response of an ex-core neutron detector to fuel assembly vibrations in an 1150-MWe Westinghouse pressurized-water reactor (PWR) was determined by performing space-dependent reactor-kinetics calculations. The effect on the detector response of reducing the soluble-boron concentration in the coolant and fuel burnup over the first fuel cycle was also determined. The results of the calculations indicate that the ex-core neutron

F. J. Sweeney; J. P. Renier

1983-01-01

92

Numerical methods for nuclear fuel burnup calculations.

??The material composition of nuclear fuel changes constantly due to nuclides transforming to other nuclides via neutron-induced transmutation reactions and spontaneous radioactive decay. The objective… (more)

Pusa, Maria

2013-01-01

93

Assessment of US NRC fuel rod behavior codes to extended burnup

The purpose of this paper is to report the status of assessing the capabilities of the NRC fuel rod performance codes for calculating extended burnup rod behavior. As part of this effort, a large spectrum of fuel rod behavior phenomena was examined, and the phenomena deemed as being influential during extended burnup operation were identified. Then, the experiment data base addressing these identified phenomena was examined for availability and completeness at extended burnups. Calculational capabilities of the NRC's steady state FRAPCON-2 and transient FRAP-T6 fuel rod behavior codes were examined for each of the identified phenomenon. Parameters calculated by the codes were compared with the available data base, and judgments were made regarding model performance. Overall, the FRAPCON-2 code was found to be moderately well assessed to extended burnups, but the FRAP-T6 code cannot be adequately assessed until more transient high burnup data are available.

Laats, E.T.; Croucher, D.W.; Haggag, F.M.

1982-01-01

94

Sensitivity and Uncertainty Analysis to Burnup Estimates on ADS using the ACAB Code

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

O. Cabellos; J. Sanz; A. Rodríguez; E. González; M. Embid; F. Alvarez; S. Reyes

2005-01-01

95

Calculation of Design Parameters for an Equilibrium LEU Core in the NBSR

A plan is being developed for the conversion of the NIST research reactor (NBSR) from high-enriched uranium (HEU) fuel to low-enriched uranium (LEU) fuel. Previously, the design of the LEU fuel had been determined in order to provide the users of the NBSR with the same cycle length as exists for the current HEU fueled reactor. The fuel composition at different points within an equilibrium fuel cycle had also been determined. In the present study, neutronics parameters have been calculated for these times in the fuel cycle for both the existing HEU and the proposed LEU equilibrium cores. The results showed differences between the HEU and LEU cores that would not lead to any significant changes in the safety analysis for the converted core. In general the changes were reasonable except that the figure-of-merit for neutrons that can be used by experimentalists shows there will be a 10% reduction in performance. The calculations included kinetics parameters, reactivity coefficients, reactivity worths of control elements and abnormal configurations, and power distributions.

Hanson, A.L.; Diamond, D.

2011-09-30

96

Neutron-physics characteristics of the VVÉR core which affect the operability of the fuel elements

UDC 621.039.5 When one calculates the thermophysical features of a fuel element, one must know: the distribution of the energy liberation over the radius of the fuel core, since the distribution affects the temperature field in the fuel element [1]; the burnup at a \\

V. D. Sidorenko; A. S. Shcheglov

1993-01-01

97

Calculation of the reactivity feedback due to core assembly bowing in LMFBRs

A computational model to calculate the reactivity feedback due to material displacements induced by assembly bowing effects has been developed and embodied in a new code called BOWPERT. While previous bowing feedback models were based on an R-Z representation of the core with user defined worth tables, the BOWPERT model is Hex-Z and requires only unambiguously defined quantities such as cross sections and fluxes. The nonuniformity of the temperature distribution in an LMFBR leads to differential thermal expansion of the walls of the assembly hexcans. These thermal expansion differentials cause the hexcan to distort or bow. Consequentially, the assembly experiences a spatial displacement, thereby resulting in a change in reactivity for the core. Although bowing effects are not expected to be sizable in large heterogeneous LMFBRs, it is important to quantify these effects.

Greenman, G.M.

1984-01-01

98

Emergence of rotational bands in ab initio no-core configuration interaction calculations

Rotational bands have been observed to emerge in ab initio no-core configuration interaction (NCCI) calculations for p-shell nuclei, as evidenced by rotational patterns for excitation energies, electromagnetic moments, and electromagnetic transitions. We investigate the ab initio emergence of nuclear rotation in the Be isotopes, focusing on 9Be for illustration, and make use of basis extrapolation methods to obtain ab initio predictions of rotational band parameters for comparison with experiment. We find robust signatures for rotational motion, which reproduce both qualitative and quantitative features of the experimentally observed bands.

M. A. Caprio; P. Maris; J. P. Vary; R. Smith

2015-02-04

99

Strategies for Application of Isotopic Uncertainties in Burnup Credit

Uncertainties in the predicted isotopic concentrations in spent nuclear fuel represent one of the largest sources of overall uncertainty in criticality calculations that use burnup credit. The methods used to propagate the uncertainties in the calculated nuclide concentrations to the uncertainty in the predicted neutron multiplication factor (k{sub eff}) of the system can have a significant effect on the uncertainty in the safety margin in criticality calculations and ultimately affect the potential capacity of spent fuel transport and storage casks employing burnup credit. Methods that can provide a more accurate and realistic estimate of the uncertainty may enable increased spent fuel cask capacity and fewer casks needing to be transported, thereby reducing regulatory burden on licensee while maintaining safety for transporting spent fuel. This report surveys several different best-estimate strategies for considering the effects of nuclide uncertainties in burnup-credit analyses. The potential benefits of these strategies are illustrated for a prototypical burnup-credit cask design. The subcritical margin estimated using best-estimate methods is discussed in comparison to the margin estimated using conventional bounding methods of uncertainty propagation. To quantify the comparison, each of the strategies for estimating uncertainty has been performed using a common database of spent fuel isotopic assay measurements for pressurized-light-water reactor fuels and predicted nuclide concentrations obtained using the current version of the SCALE code system. The experimental database applied in this study has been significantly expanded to include new high-enrichment and high-burnup spent fuel assay data recently published for a wide range of important burnup-credit actinides and fission products. Expanded rare earth fission-product measurements performed at the Khlopin Radium Institute in Russia that contain the only known publicly-available measurement for {sup 103}Rh have also been included.

Gauld, I.C.

2002-12-23

100

NASA Astrophysics Data System (ADS)

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

Espel, Federico Puente

101

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

Su, Bingjing; Hawari, Ayman, I.

2004-03-30

102

First-principle calculation of core level binding energies of LixPOyNz solid electrolyte

NASA Astrophysics Data System (ADS)

We present first-principle calculations of core-level binding energies for the study of insulating, bulk phase, compounds, based on the Slater-Janak transition state model. Those calculations were performed in order to find a reliable model of the amorphous LixPOyNz solid electrolyte which is able to reproduce its electronic properties gathered from X-ray photoemission spectroscopy (XPS) experiments. As a starting point, Li2PO2N models were investigated. These models, proposed by Du et al. on the basis of thermodynamics and vibrational properties, were the first structural models of LixPOyNz. Thanks to chemical and structural modifications applied to Li2PO2N structures, which allow to demonstrate the relevance of our computational approach, we raise an issue concerning the possibility of encountering a non-bridging kind of nitrogen atoms (=N-) in LixPOyNz compounds.

Guille, Émilie; Vallverdu, Germain; Baraille, Isabelle

2014-12-01

103

First-principle calculation of core level binding energies of LixPOyNz solid electrolyte.

We present first-principle calculations of core-level binding energies for the study of insulating, bulk phase, compounds, based on the Slater-Janak transition state model. Those calculations were performed in order to find a reliable model of the amorphous LixPOyNz solid electrolyte which is able to reproduce its electronic properties gathered from X-ray photoemission spectroscopy (XPS) experiments. As a starting point, Li2PO2N models were investigated. These models, proposed by Du et al. on the basis of thermodynamics and vibrational properties, were the first structural models of LixPOyNz. Thanks to chemical and structural modifications applied to Li2PO2N structures, which allow to demonstrate the relevance of our computational approach, we raise an issue concerning the possibility of encountering a non-bridging kind of nitrogen atoms (=N(-)) in LixPOyNz compounds. PMID:25554171

Guille, Émilie; Vallverdu, Germain; Baraille, Isabelle

2014-12-28

104

Calculation for core-s-XPS of transition metals on graphite

NASA Astrophysics Data System (ADS)

A few years ago, the appearance of surface magnetism of vanadium upon graphite was pointed out experimentally and was assumed to be quenched after exposure to CO pollutant: only for freshly evaporated V clusters on graphite, a satellite structure had been found in the V 3s-XPS spectrum and it was attributed to the presence of magnetic moments on the V surface. More recently the appearance of magnetism was observed (by means of other techniques) in a Ru monolayer grown laterally on the C(0001) graphite surface. In the present paper, we extend the impurity Anderson model for the calculation of transition metal core-s-XPS spectra by taking into account the exchange terms, especially between the core hole spin and the transition metal magnetic moment. Our results on core-s-XPS are discussed in terms of the various parameters entering the present model Hamiltonian and a new interpretation of the experimental V 3s-XPS spectrum is proposed.

Parlebas, J. C.; Krüger, P.; Taguchi, M.; Demangeat, C.; Kotani, A.

1997-04-01

105

Hybrid parallel code acceleration methods in full-core reactor physics calculations

When dealing with nuclear reactor calculation schemes, the need for three dimensional (3D) transport-based reference solutions is essential for both validation and optimization purposes. Considering a benchmark problem, this work investigates the potential of discrete ordinates (Sn) transport methods applied to 3D pressurized water reactor (PWR) full-core calculations. First, the benchmark problem is described. It involves a pin-by-pin description of a 3D PWR first core, and uses a 8-group cross-section library prepared with the DRAGON cell code. Then, a convergence analysis is performed using the PENTRAN parallel Sn Cartesian code. It discusses the spatial refinement and the associated angular quadrature required to properly describe the problem physics. It also shows that initializing the Sn solution with the EDF SPN solver COCAGNE reduces the number of iterations required to converge by nearly a factor of 6. Using a best estimate model, PENTRAN results are then compared to multigroup Monte Carlo results obtained with the MCNP5 code. Good consistency is observed between the two methods (Sn and Monte Carlo), with discrepancies that are less than 25 pcm for the k{sub eff}, and less than 2.1% and 1.6% for the flux at the pin-cell level and for the pin-power distribution, respectively. (authors)

Courau, T.; Plagne, L.; Ponicot, A. [EDF R and D, 1, Avenue du General de Gaulle, 92141 Clamart Cedex (France); Sjoden, G. [Nuclear and Radiological Engineering, Georgia Inst. of Technology, Atlanta, GA 30332 (United States)

2012-07-01

106

Analysis on burnup step effect for evaluating reactor criticality and fuel breeding ratio

NASA Astrophysics Data System (ADS)

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 thereactor 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 runthe 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 typeas 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.

Saputra, Geby; Purnama, Aditya Rizki; Permana, Sidik; Suzuki, Mitsutoshi

2014-09-01

107

Thermal-hydraulic calculations for the conversion to LEU of a research reactor core

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)

Grigoriadis, D. [National Center for Scientific Research 'DEMOKRITOS', 153 10 Aghia Paraskevi (Greece); Department of Mechanical and Manufacturing Engineering, University of Cyprus, P.O. Box 20537, Nicosia 1678 (Cyprus); Varvayanni, M.; Catsaros, N.; Stakakis, E. [National Center for Scientific Research 'DEMOKRITOS', 153 10 Aghia Paraskevi (Greece)

2008-07-15

108

Non-Invasive Characterization of Burnup for PWR Spent Fuel Rods with Burnups > 80 GWd/t

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)

Caruso, S.; Murphy, M.; Jatuff, F.; Chawla, R. [Paul Scherrer Institute, Laboratory for Reactor Physics and Systems Behavior, CH-5232 Villigen PSI (Switzerland)

2006-07-01

109

Applicability of 3D Monte Carlo simulations for local values calculations in a PWR core

NASA Astrophysics Data System (ADS)

As technical support of the French Nuclear Safety Authority, IRSN has been developing the MORET Monte Carlo code for many years in the framework of criticality safety assessment and is now working to extend its application to reactor physics. For that purpose, beside the validation for criticality safety (more than 2000 benchmarks from the ICSBEP Handbook have been modeled and analyzed), a complementary validation phase for reactor physics has been started, with benchmarks from IRPHEP Handbook and others. In particular, to evaluate the applicability of MORET and other Monte Carlo codes for local flux or power density calculations in large power reactors, it has been decided to contribute to the "Monte Carlo Performance Benchmark" (hosted by OECD/NEA). The aim of this benchmark is to monitor, in forthcoming decades, the performance progress of detailed Monte Carlo full core calculations. More precisely, it measures their advancement towards achieving high statistical accuracy in reasonable computation time for local power at fuel pellet level. A full PWR reactor core is modeled to compute local power densities for more than 6 million fuel regions. This paper presents results obtained at IRSN for this benchmark with MORET and comparisons with MCNP. The number of fuel elements is so large that source convergence as well as statistical convergence issues could cause large errors in local tallies, especially in peripheral zones. Various sampling or tracking methods have been implemented in MORET, and their operational effects on such a complex case have been studied. Beyond convergence issues, to compute local values in so many fuel regions could cause prohibitive slowing down of neutron tracking. To avoid this, energy grid unification and tallies preparation before tracking have been implemented, tested and proved to be successful. In this particular case, IRSN obtained promising results with MORET compared to MCNP, in terms of local power densities, standard deviations and computing times.

Bernard, Franck; Cochet, Bertrand; Jinaphanh, Alexis; Jacquet, Olivier

2014-06-01

110

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)

Jung, Y. S.; Joo, H. G. [Seoul National University, 1 Gwanak-ro, Gwanak-gu, Seoul (Korea, Republic of); Yoon, J. I. [KEPCO Nuclear Fuel, 1047 Daedukdae-ro, Yuseong-gu, Daejeon (Korea, Republic of)

2013-07-01

111

Burnup study for Pakistan Research Reactor1 utilizing high density low enriched uranium fuel

Burnup study for Pakistan Research Reactor-1 (PARR-1), which is a typical swimming pool type MTR utilizing high density low enriched uranium fuel, was performed by using Fuel Cycle Analysis Program (FCAP). Existing equilibrium core of PARR-1, which is relatively economical but provides less neutron fluxes per unit power than the first equilibrium core, was formed by adding five more fuel

Rizwan Ahmed; Aslam; Nasir Ahmad

2005-01-01

112

NASA Astrophysics Data System (ADS)

The main characteristics of the neutron field formed within the massive (512 kg) natural uranium target assembly (TA) QUINTA irradiated by deuteron beam of JINR Nuclotron with energies 1,2,4, and 8 GeV as well as the spatial distributions and the integral numbers of (n,f), (n,?) and (n,xn)- reactions were calculated and compared with experimental data [1] . The MCNPX 27e code with ISABEL/ABLA/FLUKA and INCL4/ABLA models of intra-nuclear cascade (INC) and experimental cross-sections of the corresponding reactions were used. Special attention was paid to the elucidation of the role of charged particles (protons and pions) in the fission of natural uranium of TA QUINTA. Extensive calculations have been done for quasi-infinite (with very small neutron leakage) depleted uranium TA BURAN having mass about 20 t which are intended to be used in experiments at Nuclotron in 2014-2016. As in the case of TA QUINTA which really models the central zone of TA BURAN the total numbers of fissions, produced 239Pu nuclei and total neutron multiplicities are predicted to be proportional to proton or deuteron energy up to 12 GeV. But obtained values of beam power gain are practically constant in studied incident energy range and are approximately four. These values are in contradiction with the experimental result [2] obtained for the depleted uranium core weighting three tons at incident proton energy 0.66 GeV.

Zhivkov, P.; Furman, W.; Stoyanov, Ch

2014-09-01

113

Review of Technical Studies in the United States in Support of Burnup Credit Regulatory Guidance

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.

Wagner, John C [ORNL] [ORNL; Parks, Cecil V [ORNL] [ORNL; Mueller, Don [ORNL] [ORNL; Gauld, Ian C [ORNL] [ORNL

2010-01-01

114

SENSITIVITY AND UNCERTAINTY ANALYSIS OF COMMERCIAL REACTOR CRITICALS FOR BURNUP CREDIT

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).

Radulescu, Georgeta [ORNL; Mueller, Don [ORNL; Wagner, John C [ORNL

2009-01-01

115

NASA Astrophysics Data System (ADS)

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.

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

116

The burnup calculational procedures for fast breeder reactors are briefly reviewed; disadvantages of the procedures are discussed and several improvements are outlined and implemented. Microscopic group constants, which are usually considered time independent over each burnup interval, are replaced by time dependent expressions. These time dependencies are evaluated in terms of polynomials of the actinide number densities with predetermined coefficients.

R. C. Borg; K. O. Ott

1978-01-01

117

NASA Astrophysics Data System (ADS)

We have employed the improved Peierls-Nabarro (P-N) equation to study the properties of 1/2lang110rang edge dislocation in the {111} plane in face-centered cubic (FCC) metals Al, Cu, Ir, Pd and Pt. The generalized-stacking-fault energy surface entering the equation is calculated by using first-principles density functional theory (DFT). The accuracy of the method has been tested by calculating the values for various stacking fault energies that favorably compare with previous theoretical and experimental results. The core structures, including the core widths of the edge and screw components, and dissociation behavior have been investigated. The dissociated distance between two partials for Al in our calculation agrees well with the values obtained from numerical simulation with DFT and molecular dynamics simulation, as well as experiment. Our calculations show that it is preferred to create partial dislocations in Cu, and easily observed full dislocations in Al, Ir, Pd and especially Pt.

Wang, Rui; Wang, Shaofeng; Wu, Xiaozhi

2011-04-01

118

Accident source terms for light-water nuclear power plants using high-burnup or MOX fuel.

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.

Salay, Michael (U.S. Nuclear Regulatory Commission, Washington, D.C.); Gauntt, Randall O.; Lee, Richard Y. (U.S. Nuclear Regulatory Commission, Washington, D.C.); Powers, Dana Auburn; Leonard, Mark Thomas

2011-01-01

119

Methodologies to assess potential lifetime limits for extended burnup nuclear fuel

predict no increased propensity for cladding failure due to PCI induced SCC at extended burnups. The other model predicts a significant probability of cladding failure occurring for a control rod ejection event at extended burnups. Generally, rod... 9 Forcing functions for the Control Rod Ejection at full power transient 168 10 Transient event cladding hoop stress calculations. . 182 11 Model results for cladding failure. . . . 185 LIST OF FIGURES Figure Page 1 High buinup total data...

De Vore, Curtis Vincent

2012-06-07

120

Dependence of Fast Reactor Fuel Burnup Characteristics on Nuclear Data Libraries

In the framework of the development of burnup calculation method for commercial fast reactors, a sensitivity analysis was carried out to clarify the dependence of fuel burnup characteristics on nuclear data libraries (NDLs). The following NDLs were compared: JEF-2.2, ENDF\\/B-VI Release 5, JENDL-3.2 and JENDL-3.3. The NDL-dependence of material balance for main heavy metal nuclides (U, U, Pu, Pu and

Shigeo OHKI; Tomoyuki JIN

2005-01-01

121

C 1s and N 1s core excitation of aniline: Experiment by electron impact and ab initio calculations

Core shell excitation spectra of aniline at the carbon and nitrogen 1s edges have been obtained by inner-shell electron energy-loss spectroscopy recorded under scattering conditions where electric dipolar conditions dominate, with higher resolution than in the previous studies. They are interpreted with the aid of ab initio configuration interaction calculations. The spectrum at the C 1s edge is dominated by an intense {pi}{sup *} band. The calculated chemical shift due to the different chemical environment at the carbon 1s edge calculated is in agreement with the experimental observations within a few tenths of an eV. The transition energies of the most intense bands in the C 1s excitation spectrum are discussed at different levels of calculations. In the nitrogen 1s excitation spectrum the most intense bands are due to Rydberg-valence transitions involving the {sigma}{sup *}-type molecular orbitals, in agreement with the experiment. This assignment is different from that of extended Hueckel molecular orbital calculations. The geometries of the core excited states have been calculated and compared to their equivalent core molecules and benzene.

Duflot, D.; Flament, J.-P.; Giuliani, A.; Heinesch, J.; Grogna, M.; Hubin-Franskin, M.-J. [Laboratoire de Physique des Lasers, Atomes et Molecules (PhLAM), UMR CNRS 8523, Centre d'Etudes et de Recherches Lasers et Applications (CERLA, FR CNRS 2416), Universite des Sciences et Technologies de Lille, F-59655 Villeneuve d'Ascq Cedex (France); DISCO Beamline, SOLEIL Synchrotron, BP 48, L'Orme des Merisiers, 91192 Gif-sur-Yvette Cedex (France); Cepia, Institut National de la Recherche Agronomique, BP 71627, 44316 Nantes Cedex 3 (France); Laboratoire de Spectroscopie d'Electrons diffuses, Universite de Liege, Institut de Chimie B6c, B-4000 Liege 1 (Belgium)

2007-05-15

122

NASA Astrophysics Data System (ADS)

Temperature dependence of the dispersion properties of liquid crystal core photonic crystal fibers with large air fraction in clads between 300 and 2000 nm for different core diameters have been calculated by a multipole method, which is modified to treat anisotropic inclusions rigorously. In calculations, air holes are assumed to be arranged in a regular hexagonal array in fused silica and a central hole is filled with liquid crystal to create a core. Below the clearing point temperature, the liquid crystal molecules are assumed to be oriented parallel to the cylindrical axis of the holes, where the liquid crystal is highly anisotropic. The large changes of the dispersion properties are found at the clearing point temperature, where the liquid crystal becomes isotropic.

Karasawa, Naoki

2015-03-01

123

PHYSICAL REVIEW B 89, 035120 (2014) Electronic stopping power from first-principles calculations 11100, FI-00076 AALTO, Finland 2 Materials Physics Division, University of Helsinki, P.O. Box 43, FI electronic stopping power Se of energetic ions in graphitic targets from first principles. By treating core

Krasheninnikov, Arkady V.

124

NASA Astrophysics Data System (ADS)

A few white dwarfs, located in binary systems, may acquire sufficiently high mass-accretion rates resulting in the burning of carbon and oxygen under nondegenerate conditions forming an O+Ne+Mg core. These O+Ne+Mg cores are gravitationally less bound than more massive progenitor stars and can release more energy due to the nuclear burning. They are also amongst the probable candidates for low entropy r-process sites. Recent observations of subluminous Type II-P supernovae (e.g. 2005cs, 2003gd, 1999br and 1997D) were able to rekindle the interest in 8-10 Modot which develop O+Ne+Mg cores. Microscopic calculations of capture rates on 24Mg, which may contribute significantly to the collapse of O+Ne+Mg cores, using the shell model and the proton-neutron quasiparticle random-phase approximation (pn-QRPA) theory, were performed earlier and comparisons made. Simulators, however, may require these capture rates on a fine scale. For the first time, a detailed microscopic calculation of the electron and positron capture rates on 24Mg on an extensive temperature-density scale is presented here. This type of scale is more appropriate for interpolation purposes and of greater utility for simulation codes. The calculations are done using the pn-QRPA theory using a separable interaction. The deformation parameter, believed to be a key parameter in QRPA calculations, is adopted from experimental data to increase the reliability of the QRPA results further. The resulting calculated rates are up to a factor of 14 or more enhanced as compared to shell model rates and may lead to some interesting scenarios for core collapse simulators.

Nabi, Jameel-Un

2008-09-01

125

Few white dwarfs, located in binary systems, may acquire sufficiently high mass accretion rates resulting in the burning of carbon and oxygen under nondegenerate conditions forming a O+Ne+Mg core. These O+Ne+Mg cores are gravitationally less bound than more massive progenitor stars and can release more energy due to the nuclear burning. They are also amongst the probable candidates for low entropy r-process sites. Recent observations of subluminous Type II-P supernovae (e.g., 2005cs, 2003gd, 1999br, 1997D) were able to rekindle the interest in 8 -- 10 M$_{\\odot}$ which develop O+Ne+Mg cores. Microscopic calculations of capture rates on $^{24}$Mg, which may contribute significantly to the collapse of O+Ne+Mg cores, using shell model and proton-neutron quasiparticle random phase approximation (pn-QRPA) theory, were performed earlier and comparisons made. Simulators, however, may require these capture rates on a fine scale. For the first time a detailed microscopic calculation of the electron and positron capture rates on $^{24}$Mg on an extensive temperature-density scale is presented here. This type of scale is more appropriate for interpolation purposes and of greater utility for simulation codes. The calculations are done using the pn-QRPA theory using a separable interaction. The deformation parameter, believed to be a key parameter in QRPA calculations, is adopted from experimental data to further increase the reliability of the QRPA results. The resulting calculated rates are up to a factor of 14 or more enhanced as compared to shell model rates and may lead to some interesting scenario for core collapse simulators.

Jameel-Un Nabi

2014-08-15

126

The ability of liquid iron to transport heat and electric charge by conduction at extreme pressure and temperature is of paramount importance to the thermal history of the core. Thermal conductivity determines the amount of heat conducted along the core adiabat, i.e. heat not available for generation of the magnetic field, and also strongly controls the time required for the

N. de Koker; G. Steinle-Neumann; V. Vlcek

2010-01-01

127

NASA Astrophysics Data System (ADS)

We implemented the calculation of hyperfine tensors into such plane wave supercell code working with the projector augmentation wave method that incorporates hybrid density functional theory and the contribution of the spin polarization of the core states. We show that the combination of HSE06 hybrid density functional together with the contribution of the core spin polarization provides accurate results on prominent point defects in various semiconductors, where the latter effect may be enormously large, in contrast to previous expectations. We briefly discuss the relevance of our results in the light of realization of solid-state quantum bits by paramagnetic point defects.

Szász, Krisztián; Hornos, Tamás; Marsman, Martijn; Gali, Adam

2013-08-01

128

ATR PDQ and MCWO Fuel Burnup Analysis Codes Evaluation

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.

G.S. Chang; P. A. Roth; M. A. Lillo

2009-11-01

129

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)

Mai, V. T.; Fujii, T.; Wada, K.; Kitada, T. [Osaka Univ., 2-1, Yamadaoka, Suita-shi, Osaka 565-0871 (Japan); Takaki, N.; Yamaguchi, A.; Watanabe, H. [Tokai Univ., 4-1-1 Kitakaname, Hiratsuka-shi, Kanagawa, 259-1292 (Japan); Unesaki, H. [Kyoto Univ. Research Reactor Inst., Asahiro-nishi, Kumatori-cho, Sennan-gun, Osaka 590-0494 (Japan)

2012-07-01

130

A variational transport theory method for two-dimensional reactor core calculations

NASA Astrophysics Data System (ADS)

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.

Mosher, Scott W.

131

Recommendations for Addressing Axial Burnup in the PWR Burnup Credit Analyses

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.

Wagner, J.C.

2002-10-23

132

Lattice Cell Calculations, Slowing Down Theory and Computer Code Wims; Vver Type Reactors

NASA Astrophysics Data System (ADS)

The following sections are included: * INTRODUCTION * WIMS AS A TOOL FOR REACTOR CORE CALCULATIONS * GENERAL STRUCTURE OF THE WIMS CODE * WIMS APPROACH TO THE SLOWING DOWN CALCULATIONS * MULTIGROUP OSCOPIC CROSS SECTIONS, RESONANCE TREATMENT * DETERMINATION OF MULTIGROUP SPECTRA * PHYSICAL MODELS IN MAIN TRANSPORT CALCULATIONS * BURNUP CALCULATIONS * APPLICATION OF WIMSD-4 TO VVER TYPE LATTICES * FINAL REMARKS * REFERENCES * APPENDIX A: DANCOFF FACTOR - STANDARD APPROACH * APPENDIX B: FORMULAS FOR DANCOFF AND BELL FACTORS CALCULATIONS APPLIED IN PREWIM * APPENDIX C: CALCULATION OF ONE GROUP PROBABILITIES Pij IN AN ANNULAR SYSTEM * APPENDIX D: SCHAEFER'S METHOD

Moen, J.; Brekke, A.; Hall, C.

1991-01-01

133

The licensing impact of high-burnup fuel in the ex-core fuel cycle steps of transportation, fabrication, and storage has been determined. Fuel performance evaluations and model code verification are presented. Scoping type safety analyses have been performed for a number of design transients and postulated accidents for the five-batch, extended-burnup fuel cycle. Loss-of-coolant accident methods are reviewed.

Matzie, R. A.; Liu, Y.

1980-01-01

134

The sensitivity and uncertainty of various core burnup performance quantities (e.g., k[sub eff], burnup reactivity swing, local power density, etc.) to the heavy isotope fission spectrum parameters was investigated using depletion perturbation methods and ENDF\\/B-V covariance data. A brief description of the methods is followed by results of a 900-MW(thermal) fast reactor. The analysis here indicates that for a 900-MW(thermal)

T. J. Downar; J. Broda; J. Kritzer

1990-01-01

135

The main objective of the development of multifield, multicomponent thermohydrodynamic computer codes is the detailed study of hypothetical core disruptive accidents (HCDAs) in liquid-metal fast breeder reactors. The main contributions such codes are expected to make are the inclusion of detailed modeling of the relative motion of liquid and vapor (slip), the inclusion of modeling of nonequilibrium\\/nonsaturation thermodynamics, and the

J. J. Sienicki; P. B. Abramson

1978-01-01

136

Characterization and modeling of high burn-up mixed oxide fuel

NASA Astrophysics Data System (ADS)

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.

Teague, Melissa Christine

137

One of the challenges associated with implementation of burnup credit is the validation of criticality calculations used in the safety evaluation; in particular the availability and use of applicable critical experiment data. The purpose of the validation is to quantify the relationship between reality and calculated results. Validation and determination of bias and bias uncertainty require the identification of sets

Don Mueller; Bradley T Rearden; Davis Allan Reed

2010-01-01

138

Depletion analysis of the UMLRR reactor core using MCNP6

NASA Astrophysics Data System (ADS)

Accurate knowledge of the neutron flux and temporal nuclide inventory in reactor physics calculations is necessary for a variety of application in nuclear engineering such as criticality safety, safeguards, and spent fuel storage. The Monte Carlo N- Particle (MCNP6) code with integrated buildup depletion code (CINDER90) provides a high-fidelity tool that can be used to perform 3D, full core simulation to evaluate fissile material utilization, and nuclide inventory calculations as a function of burnup. The University of Massachusetts Lowell Research Reactor (UMLRR) reactor has been modeled with the deterministic based code, VENTURE and with an older version of MCNP (MCNP5). The MIT developed MCODE (MCNP ORIGEN DEPLETION CODE) was used previously to perform some limited depletion calculations. This work chronicles the use of MCNP6, released in June 2013, to perform coupled neutronics and depletion calculation. The results are compared to previously benchmarked results. Furthermore, the code is used to determine the ratio of fission products 134Cs and 137Cs (burnup indicators), and the resultant ratio is compared to the burnup of the UMLRR.

Odera, Dim Udochukwu

139

MCWO - Linking MCNP And ORIGEN2 For Fuel Burnup Analysis

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.

Gray S Chang

2005-04-01

140

NASA Astrophysics Data System (ADS)

Modular Pebble Bed Reactor (MPBR) is a high temperature gas-cooled nuclear power reactor. This project investigated the feasibility of using the passive neutron counting and active neutron/gamma counting for the on line fuel burnup measurement for MPBR. To investigate whether there is a correlation between neutron emission and fuel burnup, the MPBR fuel depletion was simulated under different irradiation conditions by ORIGEN2. It was found that the neutron emission from an irradiated pebble increases with burnup super-linearly and reaches to 104 neutron/sec/pebble at the discharge burnup. The photon emission from an irradiated pebble was found to be in the order of 1013 photon/sec/pebble at all burnup levels. Analysis shows that the neutron emission rate of an irradiated pebble is sensitive to its burnup history and the spectral-averaged one-group cross sections used in the depletion calculations, which consequently leads to large uncertainty in the correlation between neutron emission and burnup. At low burnup levels, the uncertainty in the neutron emission/burnup correlation is too high and the neutron emission rate is too low so that it is impossible to determine a pebble's burnup by on-line neutron counting at low burnup levels. At high burnup levels, the uncertainty in the neutron emission rate becomes less but is still large in quantity. However, considering the super-linear feature of the correlation, the uncertainty in burnup determination was found to be ˜7% at the discharge burnup, which is acceptable. Therefore, total neutron emission rate of a pebble can be used as a burnup indicator to determine whether a pebble should be discharged or not. The feasibility of using passive neutron counting methods for the on-line burnup measurement was investigated by using a general Monte Carlo code, MCNP, to assess the detectability of the neutron emission and the capability to discriminate gamma noise by commonly used neutron detectors. It was found that both He-3 and BF3 detector systems are able to satisfy the requirement on detection efficiency; but their gamma discrimination capability is only marginal. Even using thick gamma shielding, these two types of detectors shall deteriorate in performance after a limited period of operation time because of excess accumulative gamma exposures. Thus, two or more detector systems must be used alternatively for continuous measurement. On the other hand, fission counters were found that they can effectively discriminate gamma interference for this on-line application even without using gamma shield. However, detection efficiency of fission counters is low; thus a multi-fission-counter system (using at least 12 commercially available fission chambers) must be used to achieve the required detection efficiency. Overall, passive neutron counting could be used to provide an on-line, go/no-go decision on fuel disposition on a pebble-by-pebble basis for MPBR, if the detection system is well designed. (Abstract shortened by UMI.)

Zhao, Zhongxiang

141

Simulating the Dynamics of Earth's Core: Using NCCS Supercomputers Speeds Calculations

NASA Technical Reports Server (NTRS)

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.

2002-01-01

142

NASA Astrophysics Data System (ADS)

By combining density functional molecular dynamics simulations with a thermodynamic integration technique, we determine the free energy of metallic hydrogen and silica, SiO2, 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.

González-Cataldo, F.; Wilson, Hugh F.; Militzer, B.

2014-05-01

143

In 2010 life test of three LEU (19.7%) lead test assemblies (LTA) is expected in the existing WWR-K reactor core with regular WWR-C-type fuel assemblies and a smaller core with a beryllium insert. Preliminary analysis of test safety is to be carried out. It implies reconstruction of the reactor core history for last three years, including burnup calculation for each regular fuel assembly (FA), as well as calculation of characteristics of the test core. For the planned configuration of the test core a number of characteristics have been calculated. The obtained data will be used as input for calculations on LTA test core steady-state thermal hydraulics and on transient analysis.

Arinkin, F.; Chakrov, P.; Chekushina, L.; Gizatulin,, Sh.; Koltochnik, S.; Hanan, N.; Garner, P.; Nuclear Engineering Division; Kazakhstan Ministry of Energy and Mineral Resources

2010-03-01

144

The calculation of phase equilibria of oxide core-concrete systems

NASA Astrophysics Data System (ADS)

Thermodynamic models have been developed to describe the phase equilibria of oxide solutions appropriate for the understanding of the chemical interactions between nuclear reactor core debris and concrete. For this purpose, the Gibbs energy of the liquid phase is described by the inclusion of associate species and nonideal interactions between the components and associate species. Assessments of the thermodynamic and phase equilibrium data for the subsystems of the CaO- Al2O3- SiO2- UO2- ZrO2 system have been used to obtain a thermodynamic description of the crystalline and liquid phases in good agreement with published data. The data for the subsystems have then been combined, using well established principles, to predict the phase relationships in the ternary and quaternary systems and in the overall quinary system. The results show that the overall system cannot properly be treated as a pseudo-ideal liquid and solid solution, as used in some computer codes which attempt to model the physics and chemistry of core-concrete interactions. The limitations of the current model are discussed.

Ball, R. G. J.; Mignanelli, M. A.; Barry, T. I.; Gisby, J. A.

145

Some Nuclear Calculations of U-235-D2O Gaseous-Core Cavity Reactors

NASA Technical Reports Server (NTRS)

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.

Ragsdale, Robert G.; Hyland, Robert E.

1961-01-01

146

NASA Astrophysics Data System (ADS)

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.

Matthiä, Daniel; Meier, Matthias M.; Reitz, Günther

2014-03-01

147

NASA Astrophysics Data System (ADS)

Monthly means of the magnetic field measurements taken by ground observatories are a useful data source for studying temporal changes of the core magnetic field and the underlying core flow. However, the usual way of calculating monthly means as the arithmetic mean of all days (geomagnetic quiet as well as disturbed) and all local times (day and night) may result in contributions from external (magnetospheric and ionospheric) origin in the (ordinary, omm) monthly means. Such contamination makes monthly means less favourable for core studies. We calculated revised monthly means (rmm), and their uncertainties, from observatory hourly means using robust means and after removal of external field predictions, using an improved method for characterising the magnetospheric ring current. The utility of the new method for calculating observatory monthly means is demonstrated by inverting their first differences for core surface advective flows. The flow is assumed steady over three consecutive months to ensure uniqueness; the effects of more rapid changes should be attenuated by the weakly conducting mantle. Observatory data are inverted directly for a regularised core flow, rather than deriving it from a secular variation spherical harmonic model. The main field is specified by the CHAOS-4 model. Data from up to 128 observatories between 1997 and 2013 were used to calculate 185 flow models from the omm and rmm, for each possible set of three consecutive months. The full 3x3 (non-diagonal) data covariance matrix was used, and two-norm (least squares) minimisation performed. We are able to fit the data to the target (weighted) misfit of 1, for both omm and rmm inversions, provided we incorporate the full data covariance matrix, and produce consistent, plausible flows. Fits are better for rmm flows. The flows exhibit noticeable changes over timescales of a few months. However, they follow rapid excursions in the omm that we suspect result from external field contamination; this tends to cause more erratic flow speeds rather than a change in the flow pattern. We resolve temporal changes in flows derived from the rmm associated with two geomagnetic jerks that occurred around 2003.5 and 2004.5. Throughout the interval investigated, the band of westward flow straddling the equator in the hemisphere centred on the Greenwich meridian is well developed, and flows are considerably weaker beneath the Pacific Ocean. At most times, including at the start and end of our period of interest, an anti-clockwise gyre is seen beneath the southern Indian Ocean. These are the well-established long-term features of the flow. However, the gyre disappears and re-develops twice in the mid-2000s. These changes imply quite rapid and significant changes in length-of-day (assuming such changes set up torsional oscillations), which mimics changes thought to be associated with geomagnetic jerks. The bulk westward drift speed decreases throughout the interval, with oscillations superimposed. Sharp minima in 2003, 2006, 2009 and 2011 are at times Chulliat and Maus identified secular acceleration pulses at the core surface, with particularly prominent signatures at low latitudes.

Olsen, Nils; Whaler, Kathryn A.; Finlay, Christopher C.

2014-05-01

148

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.

DeHart, M.D.

1996-05-01

149

Ab initio no core calculations of light nuclei and preludes to Hamiltonian quantum field theory

Recent advances in ab initio quantum many-body methods and growth in computer power now enable highly precise calculations of nuclear structure. The precision has attained a level sufficient to make clear statements on the nature of 3-body forces in nuclear physics. Total binding energies, spin-dependent structure effects, and electroweak properties of light nuclei play major roles in pinpointing properties of the underlying strong interaction. Eventually,we anticipate a theory bridge with immense predictive power from QCD through nuclear forces to nuclear structure and nuclear reactions. Light front Hamiltonian quantum field theory offers an attractive pathway and we outline key elements.

Vary, J.P.; Maris, P.; /Iowa State U.; Shirokov, A.M.; /Iowa State U. /SINP, Moscow; Honkanen, H.; li, J.; /Iowa State U.; Brodsky, S.J.; /SLAC; Harindranath, A.; /Saha Inst.; Teramond, G.F.de; /Costa Rica U.

2009-08-03

150

Ab initio no core calculations of light nuclei and preludes to Hamiltonian quantum field theory

Recent advances in ab initio quantum many-body methods and growth in computer power now enable highly precise calculations of nuclear structure. The precision has attained a level sufficient to make clear statements on the nature of 3-body forces in nuclear physics. Total binding energies, spin-dependent structure effects, and electroweak properties of light nuclei play major roles in pinpointing properties of the underlying strong interaction. Eventually, we anticipate a theory bridge with immense predictive power from QCD through nuclear forces to nuclear structure and nuclear reactions. Light front Hamiltonian quantum field theory offers an attractive pathway and we outline key elements.

Vary, J. P.; Maris, P.; Honkanen, H.; Li, J. [Department of Physics and Astronomy, Iowa State University, Ames, Iowa, 50011 (United States); Shirokov, A. M. [Department of Physics and Astronomy, Iowa State University, Ames, Iowa, 50011 (United States); Skobeltsyn Institute of Nuclear Physics, Moscow State University, Moscow, 119991 (Russian Federation); Brodsky, S. J. [SLAC National Accelerator Laboratory, Stanford University, Menlo Park, California (United States); Harindranath, A. [Theory Group, Saha Institute of Nuclear Physics, 1/AF, Bidhannagar, Kolkata, 700064 (India); Teramond, G. F. de [Universidad de Costa Rica, San Jose (Costa Rica)

2009-12-17

151

ISOTOPIC MODEL FOR COMMERCIAL SNF BURNUP CREDIT

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).

A.H. Wells

2004-11-17

152

The purpose of this work is to establish a method for evaluating the burn-up values of the rod-type TRIGA spent fuel by using gamma-ray spectrometry of the short-lived fission products 97Zr/97Nb, 132I, and 140La. Fuel irradiation history is not needed in this method. Short-lived fission-product activities were established by reirradiating the spent fuels in a nuclear reactor. Based on the measured activities, 235U burn-up values can be deduced by iterative calculations. The complication caused by 239Pu production and fission is also discussed in detail. The burn-up values obtained by this method are in good agreement with those deduced from the conventional method based on long-lived fission products 137Cs, 134Cs/137Cs ratio and 106Ru/137Cs ratio. PMID:10670930

Wang, T K; Peir, J J

2000-01-01

153

Sensitivity and Uncertainty Analysis to Burnup Estimates on ADS using the ACAB Code

NASA Astrophysics Data System (ADS)

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.

Cabellos, O.; Sanz, J.; Rodríguez, A.; González, E.; Embid, M.; Alvarez, F.; Reyes, S.

2005-05-01

154

Sensitivity and Uncertainty Analysis to Burn-up Estimates on ADS Using ACAB Code

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 reevaluation of some uncertainty XSs for ADS.

Cabellos, O; Sanz, J; Rodriguez, A; Gonzalez, E; Embid, M; Alvarez, F; Reyes, S

2005-02-11

155

Sensitivity and Uncertainty Analysis to Burnup Estimates on ADS using the ACAB Code

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.

Cabellos, O. [Universidad Politecnica de Madrid, Dpto. Ingenieria Nuclear, Madrid (Spain); Sanz, J.; Rodriguez, A. [Univ. National Educacion a Distancia, Dpto. Ingenieria Energetica, Madrid (Spain); Gonzalez, E.; Embid, M.; Alvarez, F. [CIEMAT, Madrid (Spain); Reyes, S. [Lawrence Livermore National Laboratory, Livermore CA (United States)

2005-05-24

156

Auger decay of the C(2)H(2) double core-hole (DCH) states, including the single-site DCH (C1s(-2)), two-site DCH (C1s(-1)C1s(-1)), and satellite (C1s(-2)?(-1)??(+1)) states, has been investigated experimentally using synchrotron radiation combined with multi-electron coincidence method, and theoretically with the assumption of the two-step sequential model for Auger decay of the DCH states. The theoretical calculations can reproduce the experimental two-dimensional Auger spectra of the C(2)H(2) single-site DCH and satellite decays, and allow to assign the peaks appearing in the spectra in terms of sequential two-electron vacancy creations in the occupied valence orbitals. In case of the one-dimensional Auger spectrum of the C(2)H(2) two-site DCH decay, the experimental and calculated results agree well, but assignment of peaks is difficult because the first and second Auger components overlap each other. The theoretical calculations on the Auger decay of the N(2) single-site DCH state, approximately considering the effect of nuclear motion, suggest that the nuclear motion, together with the highly repulsive potential energy curves of the final states, makes an important effect on the energy distribution of the Auger electrons emitted in the second Auger decay. PMID:23249002

Tashiro, Motomichi; Nakano, Motoyoshi; Ehara, Masahiro; Penent, Francis; Andric, Lidija; Palaudoux, Jérôme; Ito, Kenji; Hikosaka, Yasumasa; Kouchi, Noriyuki; Lablanquie, Pascal

2012-12-14

157

NASA Astrophysics Data System (ADS)

Auger decay of the C2H2 double core-hole (DCH) states, including the single-site DCH (C1s-2), two-site DCH (C1s-1C1s-1), and satellite (C1s-2?-1?*+1) states, has been investigated experimentally using synchrotron radiation combined with multi-electron coincidence method, and theoretically with the assumption of the two-step sequential model for Auger decay of the DCH states. The theoretical calculations can reproduce the experimental two-dimensional Auger spectra of the C2H2 single-site DCH and satellite decays, and allow to assign the peaks appearing in the spectra in terms of sequential two-electron vacancy creations in the occupied valence orbitals. In case of the one-dimensional Auger spectrum of the C2H2 two-site DCH decay, the experimental and calculated results agree well, but assignment of peaks is difficult because the first and second Auger components overlap each other. The theoretical calculations on the Auger decay of the N2 single-site DCH state, approximately considering the effect of nuclear motion, suggest that the nuclear motion, together with the highly repulsive potential energy curves of the final states, makes an important effect on the energy distribution of the Auger electrons emitted in the second Auger decay.

Tashiro, Motomichi; Nakano, Motoyoshi; Ehara, Masahiro; Penent, Francis; Andric, Lidija; Palaudoux, Jérôme; Ito, Kenji; Hikosaka, Yasumasa; Kouchi, Noriyuki; Lablanquie, Pascal

2012-12-01

158

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.

Stout, R.B.; Merckx, K.R.; Holm, J.S.

1981-01-01

159

A Modal Expansion Equilibrium Cycle Perturbation Method for Optimizing High Burnup Fast Reactors

NASA Astrophysics Data System (ADS)

This dissertation develops a simulation tool capable of optimizing advanced nuclear reactors considering the multiobjective nature of their design. An Enhanced Equilibrium Cycle (EEC) method based on the classic equilibrium method is developed to evaluate the response of the equilibrium cycle to changes in the core design. Advances are made in the consideration of burnup-dependent cross sections and dynamic fuel performance (fission gas release, fuel growth, and bond squeeze-out) to allow accuracy in high-burnup reactors such as the Traveling Wave Reactor. EEC is accelerated for design changes near a reference state through a new modal expansion perturbation method that expands arbitrary flux perturbations on a basis of ?-eigenmodes. A code is developed to solve the 3-D, multigroup diffusion equation with an Arnoldi-based solver that determines hundreds of the reference flux harmonics and later uses these harmonics to determine expansion coefficients required to approximate the perturbed flux. The harmonics are only required for the reference state, and many substantial and localized perturbations from this state are shown to be well-approximated with efficient expressions after the reference calculation is performed. The modal expansion method is coupled to EEC to produce the later-in-time response of each design perturbation. Because the code determines the perturbed flux explicitly, a wide variety of core performance metrics may be monitored by working within a recently-developed data management system called the ARMI. Through ARMI, the response of each design perturbation may be evaluated not only for the flux and reactivity, but also for reactivity coefficients, thermal hydraulics parameters, economics, and transient performance. Considering the parameters available, an automated optimization framework is designed and implemented. A non-parametric surrogate model using the Alternating Conditional Expectation (ACE) algorithm is trained with many design perturbations and then transformed through the Physical Programming (PP) paradigm to build an aggregate objective function without iteratively determining weights. Finally, the design is optimized with standard gradient-based methods. Through the power of ACE and the transparency of PP, the optimization system allows users to locate designs that best suit their multiobjective preferences with ease.

Touran, Nicholas W.

160

NASA Astrophysics Data System (ADS)

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.

Botazzoli, Pietro; Luzzi, Lelio; Brémier, Stephane; Schubert, Arndt; Van Uffelen, Paul; Walker, Clive T.; Haeck, Wim; Goll, Wolfgang

2011-12-01

161

NASA Astrophysics Data System (ADS)

A numerical method for solving the inverse problem of determining the geometry of the multilayer shell of a Bragg waveguide that has the lowest waveguide losses for a given mode has been developed with the use of the genetic algorithm. For the calculated designs of waveguides, the distribution of the coordinates of the boundaries of shell layers has been found to be aperiodic under the condition r 1 ? ? due to the axial symmetry of the problem. Waveguide losses for the TE 01, TM 01, and HE 11 modes satisfy the conditions ? _{TM_{01} } > ? _{HE_{11} } > ? _{TE_{01} } . It follows from the dependence of n eff obtained for these modes on the core radius and number of shell layers that any change in the structure of the waveguide leads to the violation of the optimal propagation regime for these modes. A Bragg fiber waveguide with a hollow core that is designed for the TE 01 mode and directs light in the single-mode regime is presented. The main fraction of losses in this waveguide is attributed to material absorption.

Bogdanovich, D. V.

2007-10-01

162

Uncertainty in the burnup reactivity swing of liquid-metal fast reactors

The uncertainty in the burnup reactivity swing Ïk{sub b} attributable to nuclear data uncertainties is analyzed using depletion-dependent sensitivity coefficients for single- and multicycle equilibrium depletion. Four systems are analyzed with design features that encompass many of the design options considered for current U.S. advanced liquid-metal reactor cores. These systems, while characterized by very different Ïk{sub b} values in the

T. J. Downar; H. Khalil

1991-01-01

163

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

Lashkari, A; Khalafi, H; Kazeminejad, H

2013-05-01

164

Effective delayed neutron fraction and prompt neutron lifetime of Tehran research reactor mixed-core

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

Lashkari, A.; Khalafi, H.; Kazeminejad, H.

2013-01-01

165

Atmospheric burnup of the cosmos-954 reactor.

On 24 January 1978 the Russian satellite Cosmos-954 reentered the atmosphere over northern Canada. By use of high-altitude balloons, the atmosphere was sampled during 1978 up to an altitude of 39 kilometers to detect particulate debris from the reactor on board the satellite. Enriched uranium-bearing aerosols at concentrations and particle sizes compatible with partial burnup of the Cosmos-954 reactor were detected only in the high polar stratosphere. PMID:17729681

Krey, P W; Leifer, R; Benson, W K; Dietz, L A; Hendrikson, H C; Coluzza, J L

1979-08-10

166

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

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.

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

1985-06-01

167

NASA Astrophysics Data System (ADS)

Understanding the mechanisms of clustering in colloids, nanoparticles, and proteins is of significant interest in material science and both chemical and pharmaceutical industries. Recently, using an integral equation theory formalism, Bomont et al. [J. Chem. Phys. 132, 184508 (2010)] studied theoretically the temperature dependence, at a fixed density, of the cluster formation in systems where particles interact with a hard-core double Yukawa potential composed of a short-range attraction and a long-range repulsion. In this paper, we provide evidence that the low-q peak in the static structure factor, frequently associated with the formation of clusters, is a common behavior in systems with competing interactions. In particular, we demonstrate that, based on a thermodynamic self-consistency criterion, accurate structural functions are obtained for different choices of closure relations. Moreover, we explore the dependence of the low-q peak on the particle number density, temperature, and potential parameters. Our findings indicate that enforcing thermodynamic self-consistency is the key factor to calculate both thermodynamic properties and static structure factors, including the low-q behavior, for colloidal dispersions with both attractive and repulsive interactions. Additionally, a simple analysis of the mean number of neighboring particles provides a qualitative description of some of the cluster features.

Kim, Jung Min; Castañeda-Priego, Ramón; Liu, Yun; Wagner, Norman J.

2011-02-01

168

Assessment of Reactivity Margins and Loading Curves for PWR Burnup Credit Cask Designs

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).

Wagner, J.C.

2002-12-17

169

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

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)

Lancaster, D. [NuclearConsultants.com, 187 Faith Circle, Boalsburg, PA 16827 (United States); Machiels, A. [Electric Power Research Inst., Inc., 3420 Hillview Avenue, Palo Alto, CA 94304 (United States)

2012-07-01

170

A time-domain high frequency model of an induction machine, which is based on the design parameters, is proposed for calculation of common mode (CM) stator ground currents in inverter-fed AC machines. Basically, the model is an extension of the measured frequency response models and considers an adequate representation of the iron core and dielectric losses over a wide frequency range,

Oliver Magdun; Andreas Binder; Yves Gemeinder

2010-01-01

171

For electrical machine designers, core loss data are usually provided in the form of tables or curves of total loss versus flux density or frequency. These can be used to extract the loss coefficients of the core loss formulas. In this paper, three currently available formulas are discussed and compared with the loss data supplied by lamination steel manufacturers. It

Yicheng Chen; Pragasen Pillay

2002-01-01

172

Microstructural Modeling of Thermal Conductivity of High Burn-up Mixed Oxide Fuel

Predicting the thermal conductivity of oxide fuels as a function of burn-up and temperature is fundamental to the efficient and safe operation of nuclear reactors. However, modeling the thermal conductivity of fuel is greatly complicated by the radially inhomogeneous nature of irradiated fuel in both composition and microstructure. In this work, radially and temperature-dependent models for effective thermal conductivity were developed utilizing optical micrographs of high burn-up mixed oxide fuel. The micrographs were employed to create finite element meshes with the OOF2 software. The meshes were then used to calculate the effective thermal conductivity of the microstructures using the BISON fuel performance code. The new thermal conductivity models were used to calculate thermal profiles at end of life for the fuel pellets. These results were compared to thermal conductivity models from the literature, and comparison between the new finite element-based thermal conductivity model and the Duriez–Lucuta model was favorable.

Melissa Teague; Michael Tonks; Stephen Novascone; Steven Hayes

2014-01-01

173

One of the challenges associated with implementation of burnup credit is the validation of criticality calculations used in the safety evaluation; in particular the availability and use of applicable critical experiment data. The purpose of the validation is to quantify the relationship between reality and calculated results. Validation and determination of bias and bias uncertainty require the identification of sets of critical experiments that are similar to the criticality safety models. A principal challenge for crediting fission products (FP) in a burnup credit safety evaluation is the limited availability of relevant FP critical experiments for bias and bias uncertainty determination. This paper provides an evaluation of the available critical experiments that include FPs, along with bounding, burnup-dependent estimates of FP biases generated by combining energy dependent sensitivity data for a typical burnup credit application with the nuclear data uncertainty information distributed with SCALE 6. A method for determining separate bias and bias uncertainty values for individual FPs and illustrative results is presented. Finally, a FP bias calculation method based on data adjustment techniques and reactivity sensitivity coefficients calculated with the SCALE sensitivity/uncertainty tools and some typical results is presented. Using the methods described in this paper, the cross-section bias for a representative high-capacity spent fuel cask associated with the ENDF/B-VII nuclear data for 16 most important stable or near stable FPs is predicted to be no greater than 2% of the total worth of the 16 FPs, or less than 0.13 % k/k.

Mueller, Don [ORNL; Rearden, Bradley T [ORNL; Reed, Davis Allan [ORNL

2010-01-01

174

Analyzing the rod drop accident in a BWR with high burnup fuel

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

D. J. Diamond; L. Neymotin

1997-01-01

175

Development of HELIOS/CAPP code system for the analysis of block type VHTR cores

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)

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

176

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.

Saavedra, Steven F [ORNL; Charlton, William S [Texas A& M University; Solodov, Alexander A [ORNL; Ehinger, Michael H [ORNL

2010-01-01

177

NASA Technical Reports Server (NTRS)

The effective-index method and Marcatili's technique were utilized independently to calculate the electric field profile of a rib channel waveguide. Using the electric field profile calculated from each method, the theoretical coupling efficiency between a single-mode optical fiber and a rib waveguide was calculated using the overlap integral. Perfect alignment was assumed and the coupling efficiency calculated. The coupling efficiency calculation was then repeated for a range of transverse offsets.

Tuma, Margaret L.; Beheim, Glenn

1995-01-01

178

Designing Critical Experiments in Support of Full Burnup Credit

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

Don Mueller; Jeremy A Roberts

2008-01-01

179

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

John M Scaglione; Don Mueller; John C Wagner

2011-01-01

180

Analysis of burnup and isotopic compositions of BWR 9 x 9 UO{sub 2} fuel assemblies

In order to extend isotopic composition data focusing on fission product nuclides, measurements are progressing using facilities of JAEA for five samples taken from high burnup BWR 9 x 9 UO{sub 2} fuel assemblies. Neutronics analysis with an infinite assembly model was applied to the preliminary measurement data using a continuous-energy Monte Carlo burnup calculation code MVP-BURN with nuclear libraries based on JENDL-3.3 and JENDL-4.0. The burnups of the samples were determined to be 28.0, 39.3, 56.6, 68.1, and 64.0 GWd/t by the Nd-148 method. They were compared with those calculated using node-average irradiation histories of power and in-channel void fractions which were taken from the plant data. The comparison results showed that the deviations of the calculated burnups from the measurements were -4 to 3%. It was confirmed that adopting the nuclear data library based on JENDL-4.0 reduced the deviations of the calculated isotopic compositions from the measurements for {sup 238}Pu, {sup 144}Nd, {sup 145}Nd, {sup 146}Nd, {sup 148}Nd, {sup 134}Cs, {sup 154}Eu, {sup 152}Sm, {sup 154}Gd, and {sup 157}Gd. On the other hand, the effect of the revision in the nuclear. data library on the neutronics analysis was not significant for major U and Pu isotopes. (authors)

Suzuki, M.; Yamamoto, T.; Ando, Y.; Nakajima, T. [Nuclear Energy System Safety Div., Japan Nuclear Energy Safety Organization, 4-1-28 Toranomon, Minato-ku, Tokyo 105-0001 (Japan)

2012-07-01

181

Calculation of core-hole excitonic features on Al L23-edge x-ray-absorption spectra of ?-Al2O3

NASA Astrophysics Data System (ADS)

We carry out first-principles molecular-orbital calculations for model clusters composed of 21 to 41 atoms with and without inclusion of a core hole. The strongest peak that appears near the Al L23-edge x-ray-absorption spectrum and electron energy-loss spectrum of ?-Al2O3 is found to originate from the presence of a core hole. Such an effect is less significant in MgO and ?-quartz (SiO2). The cation-cation overlap population in the lowest unoccupied molecular orbital (LUMO) is found to be exceptionally strong at one of the Al-Al bonds in ?-Al2O3 because of its short Al-Al bond length. The LUMO strongly localizes when the core hole is introduced.

Tanaka, Isao; Adachi, Hirohiko

1996-08-01

182

Preparation of higher-actinide burnup and cross section samples. [LMFBR

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.

Adair, H.L.; Kobisk, E.H.; Quinby, T.C.; Thomas, D.K.; Dailey, J.M.

1981-01-01

183

NASA Astrophysics Data System (ADS)

The presented results show our consecutive steps in developing a neutron source with parameters required by Boron Neutron Capture Therapy (BNCT) at the Kyiv Research Reactor (KRR). The main goal of this work was to analyze the influence of installation of different types of uranium converters close to the reactor core on neutron beam characteristics and on level of reactor safety. The general Monte Carlo radiation transport code MCNP, version 4B, has been used for these calculations.

Gritzay, Olena; Kalchenko, Oleksandr; Klimova, Nataliya; Razbudey, Volodymyr; Sanzhur, Andriy; Binney, Stephen

2005-05-01

184

Technical Data to Justify Full Burnup Credit in Criticality Safety Licensing Analysis

Enercon Services, Inc. (ENERCON) was requested under Task Order No.2 to identify scientific and technical data needed to benchmark and justify Full Burnup Credit, which adds 16 fission products and 4 minor actinides1 to Actinide-Only burnup credit. The historical perspective for Full Burnup Credit is discussed, and interviews of organizations participating in burnup credit activities are summarized as a basis

2011-01-01

185

Benchmark of Atucha-2 PHWR RELAP5-3D control rod model by Monte Carlo MCNP5 core calculation

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)

Pecchia, M.; D'Auria, F. [San Piero A Grado Nuclear Research Group GRNSPG, Univ. of Pisa, via Diotisalvi, 2, 56122 - Pisa (Italy); Mazzantini, O. [Nucleo-electrica Argentina Societad Anonima NA-SA, Buenos Aires (Argentina)

2012-07-01

186

First 3-D calculation of core disruptive accident in a large-scale sodium-cooled fast reactor

The SIMMER-IV computer code is a three-dimensional fluid-dynamics code coupled with a fuel-pin model and a space- and energy-dependent neutron transport kinetics model. The present study has attempted the first application of SIMMER-IV to a core disruptive accident in a large-scale sodium-cooled fast reactor. A principal point of this study was to investigate reactivity effects with fuel relocation under three-dimensional

Hidemasa Yamano; Yoshiharu Tobita; Satoshi Fujita; Werner Maschek

2009-01-01

187

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)

Zhang, D.; Rahnema, F. [Georgia Institute of Technology, 770 State Street NW, Atlanta, GA 30332-0745 (United States)

2013-07-01

188

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)

Navarro, J.; Aryaeinejad, R.; Nigg, D.W. [Idaho National Laboratory, P. O. Box 1625, Idaho Falls, ID 83415 (United States)

2011-07-01

189

Iron Loss Calculation Method of Filter Inductor Core on a Single-Phase PWM Voltage Source Inverter

NASA Astrophysics Data System (ADS)

The authors have reported a novel iron loss calculation method based on a loss-map of the magnetic materials. A distinctive feature of this method is that the iron loss on the inductors can easily be calculated in many kinds of converters. In this paper, a novel iron loss calculation method of the ac filter inductor on the PWM inverter by using the loss-map is described. This method enables to calculate the iron loss due to the dynamic minor loops by using the loss map and executing the easy circuit-simulation. The relation between the control method for the PWM inverter and the iron loss is discussed. Finally, the effectiveness of this method is verified through 500W experimental set-up.

Iyasu, Seiji; Shimizu, Toshihisa; Ishii, Ken-Ichiro

190

PWR cores with silicon carbide cladding

The feasibility of using silicon carbide rather than Zircaloy cladding, to reach higher power levels and higher discharge burnups in PWRs has been evaluated. A preliminary fuel design using fuel rods with the same dimensions as in the Westinghouse Robust Fuel Assembly but with fuel pellets having 10 vol% central void has been adopted to mitigate the higher fuel temperatures that occur due to the lower thermal conductivity of the silicon carbide and to the persistence of the open clad-pellet gap over most of the fuel life. With this modified fuel design, it is possible to achieve 18 month cycles that meet present-day operating constraints on peaking factor, boron concentration, reactivity coefficients and shutdown margin, while allowing batch average discharge burnups up to 80 MWD/kgU and peak rod burnups up to 100 MWD/kgU. Power uprates of 10% and possibly 20% also appear feasible. For non-uprated cores, the silicon carbide-clad fuel has a clear advantage that increases with increasing discharge burnup. Even for comparable discharge burnups, there is a savings in enriched uranium. Control rod configuration modifications may be required to meet the shutdown margin criterion for the 20% up-rate. Silicon carbide's ability to sustain higher burnups than Zircaloy also allows the design of a licensable two year cycle with only 96 fresh assemblies, avoiding the enriched uranium penalty incurred with use of larger batch sizes due to their excessive leakage. (authors)

Dobisesky, J. P.; Carpenter, D.; Pilat, E.; Kazimi, M. S. [Center for Advanced Nuclear Energy Systems, Dept. of Nuclear Science and Engineering, Massachusetts Inst. of Technology, 77 Massachusetts Avenue 24-215, Cambridge, MA 02139-4307 (United States)

2012-07-01

191

Transverse buckling effects on solitary burn-up waves

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

Xue-Nong Chen; Werner Maschek

2005-01-01

192

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)

Dieudonne, C.; Dumonteil, E.; Malvagi, F.; Diop, C. M. [Commissariat a l'Energie Atomique et aux Energies Alternatives CEA, Service d'Etude des Reacteurs et de Mathematiques Appliquees, DEN/DANS/DM2S/SERMA/LTSD, F91191 Gif-sur-Yvette cedex (France)

2013-07-01

193

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)

Armstrong, J.; Hamilton, H.; Hyland, B. [Atomic Energy of Canada Limited, Chalk River Laboratories, Chalk River, Ontario, K0J 1J0 (Canada)

2013-07-01

194

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.

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

2014-01-01

195

A plan is being developed for the conversion of the NIST research reactor (NBSR) from high-enriched uranium (HEU) fuel to low-enriched uranium (LEU) fuel. The LEU fuel may be a monolithic foil (LEUm) of U10Mo (10% molybdenum by weight in an alloy with uranium) or a dispersion of U7Mo in aluminum (LEUd). A previous report provided neutronic calculations for the LEUm fuel and this report presents the neutronics parameters for the LEUd fuel. The neutronics parameters for the LEUd fuel are compared to those previously obtained for the present HEU fuel and the proposed LEUm fuel. The results show no significant differences between the LEUm and the LEUd other than the LEUd fuel requires slightly less uranium than the LEUm fuel due to less molybdenum being present. The calculations include kinetics parameters, reactivity coefficients, reactivity worths of control elements and abnormal configurations, and power distributions under normal operation and with misloaded fuel elements.

Hanson A. L.; Diamond D.

2014-06-30

196

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.

Cerne, S.P.; Hermann, O.W.; Westfall, R.M.

1987-10-01

197

NASA Astrophysics Data System (ADS)

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.

Muratov, V. G.; Lopatkin, A. V.

198

An empirical formulation to describe the evolution of the high burnup structure

NASA Astrophysics Data System (ADS)

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.

Lemes, Martín; Soba, Alejandro; Denis, Alicia

2015-01-01

199

NASA Astrophysics Data System (ADS)

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.

Richards, Simon D.; Davies, Nigel; Armishaw, Malcolm J.; Dobson, Geoff P.; Wright, George A.

2014-06-01

200

Design and analysis of a nuclear reactor core for innovative small light water reactors

NASA Astrophysics Data System (ADS)

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.

Soldatov, Alexey I.

201

A highly heterogeneous 3D PWR core benchmark: deterministic and Monte Carlo method comparison

NASA Astrophysics Data System (ADS)

Physical analyses of the LWR potential performances with regards to the fuel utilization require an important part of the work dedicated to the validation of the deterministic models used for theses analyses. Advances in both codes and computer technology give the opportunity to perform the validation of these models on complex 3D core configurations closed to the physical situations encountered (both steady-state and transient configurations). In this paper, we used the Monte Carlo Transport code TRIPOLI-4®; to describe a whole 3D large-scale and highly-heterogeneous LWR core. The aim of this study is to validate the deterministic CRONOS2 code to Monte Carlo code TRIPOLI-4®; in a relevant PWR core configuration. As a consequence, a 3D pin by pin model with a consistent number of volumes (4.3 millions) and media (around 23,000) is established to precisely characterize the core at equilibrium cycle, namely using a refined burn-up and moderator density maps. The configuration selected for this analysis is a very heterogeneous PWR high conversion core with fissile (MOX fuel) and fertile zones (depleted uranium). Furthermore, a tight pitch lattice is selcted (to increase conversion of 238U in 239Pu) that leads to harder neutron spectrum compared to standard PWR assembly. In these conditions two main subjects will be discussed: the Monte Carlo variance calculation and the assessment of the diffusion operator with two energy groups for the core calculation.

Jaboulay, J.-C.; Damian, F.; Douce, S.; Lopez, F.; Guenaut, C.; Aggery, A.; Poinot-Salanon, C.

2014-06-01

202

Thermal regimes of high burn-up nuclear fuel rod

NASA Astrophysics Data System (ADS)

The temperature distribution in the nuclear fuel rods for high burn-up is studied. We use the numerical and analytical approaches. It is shown that the time taken to have the stationary thermal regime of nuclear fuel rod is less than one minute. We can make the inference that the behavior of the nuclear fuel rod can be considered as a stationary task. Exact solutions of the temperature distribution in the fuel rods in the stationary case are found. Thermal regimes of high burn-up the nuclear fuel rods are analyzed.

Kudryashov, Nikolai A.; Khlunov, Aleksandr V.; Chmykhov, Mikhail A.

2010-05-01

203

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).

B. Boer; A. M. Ougouag

2010-09-01

204

Application of CANDLE burnup to block-type high temperature gas cooled reactor

The CANDLE burnup strategy, where the distributions of fuel nuclide densities, neutron flux, and power density move with the same constant speed and without any change in their shapes, is applied to the block-type high temperature gas cooled reactor. If it is successful, a burnup control rod can be eliminated, and several merits are expected. This burnup may be realized

Y. Ohoka; H. Sekimoto

2004-01-01

205

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.

Radulescu, Georgeta [ORNL] [ORNL; Gauld, Ian C [ORNL] [ORNL; Ilas, Germina [ORNL] [ORNL; Wagner, John C [ORNL] [ORNL

2011-01-01

206

Issues related to criticality safety analysis for burnup credit applications

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. Parametric analyses are required to characterize the importance of fuel assembly and fuel cycle parameters on spent fuel composition and reactivity. Numerical models are evaluated to determine the sensitivity of criticality

M. D. DeHart; C. V. Parks

1995-01-01

207

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)

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

208

Calculation of fuel pin failure timing under LOCA conditions

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.

Jones, K.R.; Wade, N.L.; Siefken, L.J.; Straka, M.; Katsma, K.R.

1991-10-01

209

Mathematical modeling of the heat treatment and combustion of a coal particle. V. Burn-up stage

NASA Astrophysics Data System (ADS)

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.

Enkhjargal, Kh.; Salomatov, V. V.

2011-07-01

210

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.

Yu. V. Petrov; A. I. Nazarov; M. S. Onegin; V. Yu. Petrov; E. G. Sakhnovsky

2005-09-15

211

A multi-group Monte Carlo core analysis method and its application in SCWR design

Complex geometry and spectrum have been the characteristics of many newly developed nuclear energy systems, so the suitability and precision of the traditional deterministic codes are doubtable while being applied to simulate these systems. On the contrary, the Monte Carlo method has the inherent advantages of dealing with complex geometry and spectrum. The main disadvantage of Monte Carlo method is that it takes long time to get reliable results, so the efficiency is too low for the ordinary core designs. A new Monte Carlo core analysis scheme is developed, aimed to increase the calculation efficiency. It is finished in two steps: Firstly, the assembly level simulation is performed by continuous energy Monte Carlo method, which is suitable for any geometry and spectrum configuration, and the assembly multi-group constants are tallied at the same time; Secondly, the core level calculation is performed by multi-group Monte Carlo method, using the assembly group constants generated in the first step. Compared with the heterogeneous Monte Carlo calculations of the whole core, this two-step scheme is more efficient, and the precision is acceptable for the preliminary analysis of novel nuclear systems. Using this core analysis scheme, a SCWR core was designed based on a new SCWR assembly design. The core output is about 1,100 MWe, and a cycle length of about 550 EFPDs can be achieved with 3-batch refueling pattern. The average and maximum discharge burn-up are about 53.5 and 60.9 MWD/kgU respectively. (authors)

Zhang, P.; Wang, K.; Yu, G. [Dept. of Engineering Physics, Tsinghua Univ., Beijing, 100084 (China)

2012-07-01

212

Structure of high-burnup-fuel Zircaloy cladding. [PWR; BWR

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

Chung

1983-01-01

213

Analysis of burnup credit in fuel storage with CASMO

Recent trends in nuclear power plant operation have tended toward longer cycles with reload fuel of high (> 3.5 wt% Â²Â³âµU) enrichments. At the same time, the need for greater spent-fuel pool capacity has reduced storage canister spacing to the point where maximum allowable fresh enrichments are lower than those necessary for longer cycles. As a result, burnup credit analysis

Napolitano

1987-01-01

214

Copyright Notice Triton Burnup Study in JT-60U

Copyright Notice Triton Burnup Study in JT-60U T. Nishitani, M. Hoek1, H. Harano2, G.A. Wurden3, R of 1 MeV tritons produced in the d(d,p)t reaction is important to predict the properties of D-T produced 3.5 MeV alphas because 1 MeV tritons and 3.5 MeV alphas have similar kinematic properties

215

In 2005 the Argentine Government took the decision to complete the construction of the Atucha-II nuclear power plant, which has been progressing slowly during the last ten years. Atucha-II is a 745 MWe nuclear station moderated and cooled with heavy water, of German (Siemens) design located in Argentina. It has a pressure vessel design with 451 vertical coolant channels and the fuel assemblies (FA) are clusters of 37 natural UO{sub 2} rods with an active length of 530 cm. For the reactor physics area, a revision and update of reactor physics calculation methods and models was recently carried out covering cell, supercell (control rod) and core calculations. This paper presents benchmark comparisons of core parameters of a slightly idealized model of the Atucha-I core obtained with the PUMA reactor code with MCNP5. The Atucha-I core was selected because it is smaller, similar from a neutronic point of view, more symmetric than Atucha-II, and has some experimental data available. To validate the new models benchmark comparisons of k-effective, channel power and axial power distributions obtained with PUMA and MCNP5 have been performed. In addition, a simple cell heterogeneity correction recently introduced in PUMA is presented, which improves significantly the agreement of calculated channel powers with MCNP5. To complete the validation, the calculation of some of the critical configurations of the Atucha-I reactor measured during the experiments performed at first criticality is also presented. (authors)

Grant, C. [Comision Nacional de Energia Atomica, Av del Libertador 8250, Buenos Aires 1429 (Argentina); Mollerach, R. [Nucleoelectrica Argentina S.A., Arribenos 3619, Buenos Aires 1429 (Argentina); Leszczynski, F.; Serra, O.; Marconi, J. [Comision Nacional de Energia Atomica, Av del Libertador 8250, Buenos Aires 1429 (Argentina); Fink, J. [Nucleoelectrica Argentina S.A., Arribenos 3619, Buenos Aires 1429 (Argentina)

2006-07-01

216

A Feasibility Study to Determine Cooling Time and Burnup of ATR Fuel Using a Nondestructive Technique1 Rahmat Aryaeinejad, Jorge Navarro, and David W Nigg Idaho National Laboratory Abstract Effective and efficient Advanced Test Reactor (ATR) fuel management require state of the art core modeling tools. These new tools will need isotopic and burnup validation data before they are put into production. To create isotopic, burn up validation libraries and to determine the setup for permanent fuel scanner system a feasibility study was perform. The study consisted in measuring short and long cooling time fuel elements at the ATR canal. Three gamma spectroscopy detectors (HPGe, LaBr3, and HPXe) and two system configurations (above and under water) were used in the feasibility study. The first stage of the study was to investigate which detector and system configuration would be better suited for different scenarios. The second stage of the feasibility study was to create burnup and cooling time calibrations using experimental isotopic data collected and ORIGEN 2.2 burnup data. The results of the study establish that a better spectra resolution is achieve with an above the water configuration and that three detectors can be used in the permanent fuel scanner system for different situations. In addition it was conclude that a number of isotopic ratios and absolute measurements could be used to predict ATR fuel burnup and cooling times. 1This work was supported by the U.S. Depart¬ment of Energy (DOE) under Battelle Energy Alliance, LLC Contract No. DE-AC07-05ID14517.

Jorge Navarro; Rahmat Aryaeinejad,; David W. Nigg

2011-05-01

217

High burnup effects on fuel behaviour under accident conditions: the tests CABRI REP-Na

NASA Astrophysics Data System (ADS)

A large, performance based, knowledge and experience in the field of nuclear fuel behaviour is available for nominal operation conditions. The database is continuously completed and precursor assembly irradiations are performed for testing of new materials and innovative designs. This procedure produces data and arguments to extend licencing limits in the permanent research for economic competitiveness. A similar effort must be devoted to the establishment of a database for fuel behaviour under off-normal and accident conditions. In particular, special attention must be given to the so-called design-basis-accident (DBA) conditions. Safety criteria are formulated for these situations and must be respected without consideration of the occurrence probability and the risk associated to the accident situation. The introduction of MOX fuel into the cores of light water reactors and the steadily increasing goal burnup of the fuel call for research work, both experimental and analytical, in the field of fuel response to DBA conditions. In 1992, a significant programme step, CABRI REP-Na, has been launched by the French Nuclear Safety and Protection Institute (IPSN) in the field of the reactivity initiated accident (RIA). After performing the nine experiments of the initial test matrix it can be concluded that important new findings have been evidenced. High burnup clad corrosion and the associated degradation of the mechanical properties of the ZIRCALOY4 clad is one of the key phenomena of the fuel behaviour under accident conditions. Equally important is the evidence that transient, dynamic fission gas effects resulting from the close to adiabatic heating introduces a new explosive loading mechanism which may lead to clad rupture under RIA conditions, especially in the case of heterogeneous MOX fuel.

Schmitz, Franz; Papin, Joelle

218

Assessment of Fission Product Cross-Section Data for Burnup Credit Applications

Past efforts by the Department of Energy (DOE), the Electric Power Research Institute (EPRI), the Nuclear Regulatory Commission (NRC), and others have provided sufficient technical information to enable the NRC to issue regulatory guidance for implementation of pressurized-water reactor (PWR) burnup credit; however, consideration of only the reactivity change due to the major actinides is recommended in the guidance. Moreover, DOE, NRC, and EPRI have noted the need for additional scientific and technical data to justify expanding PWR burnup credit to include fission product (FP) nuclides and enable burnup credit implementation for boiling-water reactor (BWR) spent nuclear fuel (SNF). The criticality safety assessment needed for burnup credit applications will utilize computational analyses of packages containing SNF with FP nuclides. Over the years, significant efforts have been devoted to the nuclear data evaluation of major isotopes pertinent to reactor applications (i.e., uranium, plutonium, etc.); however, efforts to evaluate FP cross-section data in the resonance region have been less thorough relative to actinide data. In particular, resonance region cross-section measurements with corresponding R-matrix resonance analyses have not been performed for FP nuclides. Therefore, the objective of this work is to assess the status and performance of existing FP cross-section and cross-section uncertainty data in the resonance region for use in burnup credit analyses. Recommendations for new cross-section measurements and/or evaluations are made based on the data assessment. The assessment focuses on seven primary FP isotopes (103Rh, 133Cs, 143Nd, 149Sm, 151Sm, 152Sm, and 155Gd) that impact reactivity analyses of transportation packages and two FP isotopes (153Eu and 155Eu) that impact prediction of 155Gd concentrations. Much of the assessment work was completed in 2005, and the assessment focused on the latest FP cross-section evaluations available in the international nuclear data community as of March 2005. The accuracy of the cross-section data was investigated by comparing existing cross-section evaluations against available measured cross-section data. When possible, benchmark calculations were also used to assess the performance of the latest FP cross-section data. Since March 2005, the U.S. and European data projects have released newer versions of their respective data files. Although there have been updates to the international data files and to some degree FP data, much of the updates have included nuclear cross-section modeling improvements at energies above the resonance region. The one exception is improved ENDF/B-VII cross-section uncertainty data or covariance data for gadolinium isotopes. In particular, ENDF/B-VII includes improved 155Gd resonance parameter covariance data, but they are based on previously measured resonance data. Although the new covariance data are available for 155Gd, the conclusions of the FP cross-section data assessment of this report still hold in lieu of the newer international cross-section data files. Based on the FP data assessment, there is judged to be a need for new total and capture cross-section measurements and corresponding cross-section evaluations, in a prioritized manner, for the nine FPs to provide the improved information and technical rigor needed for criticality safety analyses.

Leal, Luiz C [ORNL; Derrien, Herve [ORNL; Dunn, Michael E [ORNL; Mueller, Don [ORNL

2007-12-01

219

Spent LWR fuel dry storage in large transport and storage casks after extended burnup

NASA Astrophysics Data System (ADS)

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.

Spilker, Harry; Peehs, Martin; Dyck, Hans-Peter; Kaspar, Guenter; Nissen, Klaus

1997-11-01

220

NASA Astrophysics Data System (ADS)

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.

Su'ud, Zaki; Sekimoto, H.

2014-09-01

221

NASA Astrophysics Data System (ADS)

Statistical errors in sampling neutron fields in physically large systems like an RBMK are analyzed both qualitatively and quantitatively. Recommendations concerning the choice of parameters for calculations are given. A new procedure for Monte Carlo RBMK calculations with model corrections on the basis of data from in-core detectors is proposed. Dedicated software based on the CUDA software and hardware platform is developed for computational research. Results of testing the procedure and software in question via calculations for real RBMK reactors are discussed.

Ivanov, I. E.; Schukin, N. V.; Bychkov, S. A.; Druzhinin, V. E.; Lysov, D. A.; Shmonin, Yu. V.; Gurevich, M. I.

2014-12-01

222

Extended burnup fuel - a beneficial environmental step for the nuclear fuel cycle

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.

Tulenko, J.S. (Univ. of Florida, Gainesville (United States))

1992-01-01

223

Investigation of the Performance of D2O-Cooled High-Conversion Reactors for Fuel Cycle Calculations

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.

Hikaru Hiruta; Gilles Youinou

2013-09-01

224

Analysis of high burnup fuel behavior in Halden reactor by FEMAXI–V code

The author developed a code FEMAXI–V to analyze the behaviors of high burnup LWR fuels. FEMAXI–V succeeded the basic structure of code FEMAXI–IV, and incorporated such new models and functions as fuel thermal conductivity degradation with burnup, alliance with burnup analysis code which gives radial power profile and fast neutron flux, etc. In the present analysis, coolant conditions, detailed power

Suzuki Motoe

2000-01-01

225

ORIGEN2 was used to develop a data base of pressurized water reactor isotopic concentrations at various times after discharge with core burnup, specific power, enrichment, and neutron spectrum as variables. Results were analyzed to determine source term sensitivity to core management. Fuel rod power history was found to have an important effect on the source term. Activity and decay power

J. K. Wheeler; A. Sesonske

1986-01-01

226

Analyse de l'impact de l'environnement dans un schema de calcul a deux etapes avec DRAGON et DONJON

NASA Astrophysics Data System (ADS)

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.

Bodin, Christophe

227

Deep burn-up and transmutation of plutonium for a selected HTGR reference design

This article comprises results of equilibrium core calculations for a simplified HTGR (high tem- perature gas cooled reactor) based on the PBMR reactor design with a pure plutonium oxide fuel (first genera- tion plutonium of a light water reactor). Different aspects were investigated as isotope compositions of pluto- nium, minor actinides and fission products in the equilibrium core and the

Ch. Pohl; H. J. Rütten; K. Haas

228

NASA Technical Reports Server (NTRS)

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.

Triner, J. E.

1979-01-01

229

-2 Temperature Rise Calculations 2-3 Core Selector Charts TECHNICAL DATA 3-1 Material Properties 3-2 Conversion Tables 3-3 Normal Magnetization Curves 3-5 Core Loss Density Curves 3-12 Permeability versus Temperature versus Frequency Curves 3-21 Wire Table CORE DATA 4-1 Toroid Data 4-31 Kool MÂµÂ® E Core Data 4-33 MPP

230

Core-core and core-valence correlation

NASA Technical Reports Server (NTRS)

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.

Bauschlicher, Charles W., Jr.; Langhoff, Stephen R.; Taylor, Peter R.

1988-01-01

231

A Metal Fuel Core Concept for 1000 MWt Advanced Burner Reactor

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)

Yang, W.S.; Kim, T.K.; Grandy, C. [Argonne National Laboratory, 9700 S. Cass Avenue, Argonne IL 60439 (United States)

2007-07-01

232

Technical Data to Justify Full Burnup Credit in Criticality Safety Licensing Analysis

Enercon Services, Inc. (ENERCON) was requested under Task Order No.2 to identify scientific and technical data needed to benchmark and justify Full Burnup Credit, which adds 16 fission products and 4 minor actinides1 to Actinide-Only burnup credit. The historical perspective for Full Burnup Credit is discussed, and interviews of organizations participating in burnup credit activities are summarized as a basis for identifying additional data needs and making recommendation. Input from burnup credit participants representing two segments of the commercial nuclear industry is provided. First, the Electric Power Research Institute (EPRI) has been very active in the development of Full Burnup Credit, representing the interests of nuclear utilities in achieving capacity gains for storage and transport casks. EPRI and its utility customers are interested in a swift resolution of the validation issues that are delaying the implementation of Full Burnup Credit [EPRI 2010b]. Second, used nuclear fuel storage and transportation Cask Vendors favor improving burnup credit beyond Actinide-Only burnup credit, although their discussion of specific burnup credit achievements and data needs was limited citing business sensitive and technical proprietary concerns. While Cask Vendor proprietary items are not specifically identified in this report, the needs of all nuclear industry participants are reflected in the conclusions and recommendations of this report. In addition, Oak Ridge National Laboratory (ORNL) and Sandia National Laboratory (SNL) were interviewed for their input into additional data needs to achieve Full Burnup Credit. ORNL was very open to discussions of Full Burnup Credit, with several telecoms and a visit by ENERCON to ORNL. For many years, ORNL has provided extensive support to the NRC regarding burnup credit in all of its forms. Discussions with ORNL focused on potential resolutions to the validation issues for the use of fission products. SNL was helpful in ENERCON's understanding of the difficult issues related to obtaining and analyzing additional cross section test data to support Full Burnup Credit. A PIRT (Phenomena Identification and Ranking Table) analysis was performed by ENERCON to evaluate the costs and benefits of acquiring different types of nuclear data in support of Full Burnup Credit. A PIRT exercise is a formal expert elicitation process with the final output being the ranking tables. The PIRT analysis (Table 7-4: Results of PIRT Evaluation) showed that the acquisition of additional Actinide-Only experimental data, although beneficial, was associated with high cost and is not necessarily needed. The conclusion was that the existing Radiochemical Assay (RCA) data plus the French Haut Taux de Combustion (HTC)2 and handbook Laboratory Critical Experiment (LCE) data provide adequate benchmark validation for Actinide-Only Burnup Credit. The PIRT analysis indicated that the costs and schedule to obtain sufficient additional experimental data to support the addition of 16 fission products to Actinide-Only Burnup Credit to produce Full Burnup Credit are quite substantial. ENERCON estimates the cost to be $50M to $100M with a schedule of five or more years. The PIRT analysis highlights another option for fission product burnup credit, which is the application of computer-based uncertainty analyses (S/U - Sensitivity/Uncertainty methodologies), confirmed by the limited experimental data that is already available. S/U analyses essentially transform cross section uncertainty information contained in the cross section libraries into a reactivity bias and uncertainty. Recent work by ORNL and EPRI has shown that a methodology to support Full Burnup Credit is possible using a combination of traditional RCA and LCE validation plus S/U validation for fission product isotopics and cross sections. Further, the most recent cross section data (ENDF/B-VII) can be incorporated into the burnup credit codes at a reasonable cost compared to the acquisition of equivalent experimental data. ENERCON concludes that even with the cos

Enercon Services, Inc.

2011-03-14

233

Attainable Burnup in a LIFE Engine Loaded with Depleted Uranium

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.

Fratoni, M; Kramer, K J; Latkowski, J F

2009-11-30

234

Microstructural Characterization of High Burn-up Mixed Oxide Fast Reactor Fuel

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.

Melissa C. Teague; Brian P. Gorman; Steven L. Hayes; Douglas L. Porter; Jeffrey King

2013-10-01

235

Modular Pebble Bed Reactor (MPBR) is a high temperature gas-cooled nuclear power reactor. This project investigated the feasibility of using the passive neutron counting and active neutron\\/gamma counting for the on line fuel burnup measurement for MPBR. To investigate whether there is a correlation between neutron emission and fuel burnup, the MPBR fuel depletion was simulated under different irradiation conditions

Zhongxiang Zhao

2004-01-01

236

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

M. D. Dehart

1996-01-01

237

Burnup verification measurements on spent fuel assemblies at Arkansas Nuclear One

Burnup verification measurements have been performed using the Fork system at Arkansas Nuclear One, Units 1 and 2, operated by Energy Operations, Inc. Passive neutron and gamma-ray measurements on individual spent fuel assemblies were correlated with the reactor records for burnup, cooling time, and initial enrichment. The correlation generates an internal calibration for the system in the form of a

Ewing

1995-01-01

238

Burnup verification measurements at U.S. Nuclear Facilities using the Fork system

Burnup verification measurements have been performed using the Fork system at the Oconee Nuclear Station of Duke Power Company, and at Arkansas Nuclear One (Units 1 and 2), operated by Energy Operations, Inc. Passive neutron and gamma-ray measurements on individual spent fuel assemblies were correlated with the reactor records for burnup, cooling time, and initial enrichment. The correlation generates an

Ewing

1995-01-01

239

An analysis of burnup reactivity credit for reactor RA spent fuel storage

The need for increasing the spent fuel storage capacity has led to the development of validated methods for assessing the reactivity effects associated with fuel burnup. This paper gives an overview of the criticality safety analysis methodology used to investigate the sensitivity of storage system reactivities to changes in fuel burnup. Results representing the validation of the methods are also

M. J Miloševi?; M. P Peši?

1998-01-01

240

Federal Register 2010, 2011, 2012, 2013, 2014

...Draft Test Plan for the High Burnup Dry Storage Cask Research and Development Project...draft test plan for the High Burnup Dry Storage Cask Research and Development Project...the execution of the High Burnup Dry Storage Cask Research and Development...

2013-11-12

241

The burnup dependence of light water reactor spent fuel oxidation

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).

Hanson, B.D.

1998-07-01

242

The nodal discrete ordinates (SN) transport calculation code for three-dimensional hexagonal geometry NSHEX treats intranode flux distribution using a polynomial series and considers the angular dependence of flux by the SN method. For the improvement of calculation accuracy of NSHEX for practical use to large-size fast reactor plants, the maximum order of the polynomial series is extended from two to

Kazuteru SUGINO; Teruhiko KUGO

2011-01-01

243

Burnup and feasibility study of low power density PWR's

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

Molins-Bartra

1981-01-01

244

Temperature and burnup correlated fuel-cladding chemical interaction in U-10ZR metallic fuel

NASA Astrophysics Data System (ADS)

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.

Carmack, William J.

245

MCNPX Monte Carlo burnup simulations of the isotope correlation experiments in the NPP obrigheim.

This paper describes the simulation work of the Isotope Correlation Experiment (ICE) using the MCNPX Monte Carlo computer code package. The Monte Carlo simulation results are compared with the ICE-Experimental measurements for burnup up to 30 GWD/t. The comparison shows the good capabilities of the MCNPX computer code package for predicting the depletion of the uranium fuel and the buildup of the plutonium isotopes in a PWR thermal reactor. The Monte Carlo simulation results show also good agreements with the experimental data for calculating several long-lived and stable fission products. However, for the americium and curium actinides, it is difficult to judge the predication capabilities for these actinides due to the large uncertainties in the ICE-Experimental data. In the MCNPX numerical simulations, a pin cell model is utilized to simulate the fuel lattice of the nuclear power reactor. Temperature dependent libraries based on JEFF3.1 nuclear data files are utilized for the calculations. In addition, temperature dependent libraries based ENDF/B-VII nuclear data files are utilized and the obtained results are very close to the JEFF3.1 results, except for {approx}10% differences in the prediction of the minor actinide isotopes buildup.

Cao, Y.; Gohar, Y.; Broeders, C. (Nuclear Engineering Division); (Inst. for Neutron Physics and Reactor Technology)

2010-10-01

246

Generation of lumped fission product cross sections for high burnup, highly enriched uranium fuel

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.

Primm, R.T. III; Greene, N.M.

1988-01-01

247

Sensitivity of Candu-Scwr Reactor Physics Calculations to Nuclear Data Files

A simplified MCNP model of a CANDU-SCWR lattice was used to test the sensitivity of the calculated reactivity to various nuclear data files involving issues of current interest. These tests were performed for cooled and voided conditions, with and without axial neutron leakage, for a uniform lattice of mid-life fuel and a mixed lattice with high-burnup and low-burnup fuel in

K. S. Kozier; G. R. Dyck

2006-01-01

248

Using ORIGEN\\/KENO to calculate burnup credit for spent-fuel pool criticality analyses

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

C. T. Rombough; S. H. Martonak; J. Walkin

1994-01-01

249

NASA Astrophysics Data System (ADS)

We analyze the accuracy of the atomic force within the all-electron full-potential linearized augmented plane-wave (FLAPW) method using the force formalism of Yu et al. [Phys. Rev. B 43, 6411 (1991), 10.1103/PhysRevB.43.6411]. A refinement of this formalism is presented that explicitly takes into account the tail of high-lying core states leaking out of the muffin-tin sphere and considers the small discontinuities of LAPW wave function, density, and potential at the muffin-tin sphere boundaries. For MgO and EuTiO3 it is demonstrated that these amendments substantially improve the acoustic sum rule and the symmetry of the force constant matrix. Sum rule and symmetry are realized with an accuracy of ? Htr /aB .

Klüppelberg, Daniel A.; Betzinger, Markus; Blügel, Stefan

2015-01-01

250

Development of Technical Basis for Burnup Credit Regulatory Guidance in the United States

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.

Parks, Cecil V [ORNL] [ORNL; Wagner, John C [ORNL] [ORNL; Mueller, Don [ORNL] [ORNL; Gauld, Ian C [ORNL] [ORNL

2011-01-01

251

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.

Scaglione, John M [ORNL] [ORNL; Mueller, Don [ORNL] [ORNL; Wagner, John C [ORNL] [ORNL

2011-01-01

252

LWR fuel-cycle costs as a function of burnup. Final report

Utilities may be able to decrease fuel-cycle costs as much as 5% in PWRs and 6% in BWRs by increasing discharge burnup to optimum practical limits. With one exception, this analysis of 12- and 18-month fuel cycles indicated a potential for still further cost reductions at higher burnup rates than those considered (39 GWd/MtU for BWRs and 55 GWd/MtU for PWRs).

Franks, W.; Goldstein, L.; Joseph, L.; Nikmohammadian, N.

1984-11-01

253

The Effect of Pitch, Burnup, and Absorbers on a TRIGA Spent-Fuel Pool Criticality Safety

It has been shown that supercriticality might occur for some postulated accident conditions at the TRIGA spent-fuel pool. However, the effect of burnup was not accounted for in previous studies. In this work, the combined effect of fuel burnup, pitch among fuel elements, and number of uniformly mixed absorber rods for a square arrangement on the spent-fuel pool k{sub eff}

Marjan Logar; Robert Jeraj; Bogdan Glumac

2003-01-01

254

ACCURATE NUCLEAR FUEL BURNUP ANALYSES. First Quarterly Report December 1961February 1962

S> Activities in a program to develop mass spectrometric techniques for ; use in reactor fuel burnup analysis are reported. The program emphasis is on ; measurement of nonradioactive refractory fission products that can be related to ; burnup. A controlled irradiation program is being initiated to prepare foils of ; UÂ²Â³âµ, PuÂ²Â³â¹, and UÂ²Â³Â³ for use in development of

Rider

1962-01-01

255

Creep assessment of Zry-4 cladded high burnup fuel under dry storage

Cladding creep rupture is thought to be the most likely and limiting failure mechanism of spent fuel in dry storage. In spite of being highly unlikely, the current trend towards high burnups is drawing further attention to the potential creep effect on cladding integrity of fuels burnt over 45 GWd\\/tU.This paper explores the burnup influence on cladding creep during dry storage

F. Feria; L. E. Herranz

2011-01-01

256

A Simplified Approach for Evaluation of the Burnup Potential of Alternative Fuels

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

Stella Maris Oggianu; Hee Cheon No; Mujid S. Kazimi

2004-01-01

257

High Burnup Effects Program A State-of-the-Technology Assessment

Various analytical models and empirical correlations describing the fission gas release phenomenon were examined. An evaluation was made of the current pertinent experimental data on the subject of high burnup fission gas release. Data reported by individual investigators were compared and evaluated in relation to their applicability to the content and scope of the High Burnup Effects Program. These evaluations then form the bases for defining the data needs, and the selection of variables to be studied in this program.

Rising, K. H.; Bradley, E. R.; Williford, R. E.; Freshley, M D.

1982-06-01

258

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.

Hans D. Gougar

2009-08-01

259

Fuel Burnup and Fuel Pool Shielding Analysis for Bushehr Nuclear Reactor VVER-1000

NASA Astrophysics Data System (ADS)

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.

Hadad, Kamal; Ayobian, Navid

260

Sensitivity study on Xe depletion in the high burn-up structure of UO2

NASA Astrophysics Data System (ADS)

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.

Holt, L.; Schubert, A.; Van Uffelen, P.; Walker, C. T.; Fridman, E.; Sonoda, T.

2014-09-01

261

The Encapsulated Nuclear Heat Source (ENHS) Reactor Core Design

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.

Hong, Ser Gi [Korea Atomic Energy Research Institute (Korea, Republic of); Greenspan, Ehud [University of California, Berkeley (United States); Kim, Yeong Il [Korea Atomic Energy Research Institute (Korea, Republic of)

2005-01-15

262

A search for minimum volume of Breed and Burn cores

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)

Di Sanzo, C.; Greenspan, E. [Dept. of Nuclear Engineering, Univ. of California, Berkeley Etcheverry Hall, Berkeley, CA 94720 (United States)

2012-07-01

263

Sensitivity analysis of hot channel calculation methods

In safety analysis, the fulfillment of acceptance criteria is usually evaluated by separate hot channel or\\/and hot assembly thermal hydraulic\\/fuel behavior calculations. The whole range of the relevant input parameters (e.g. power distributions, burnup, heat conduction data, inlet temperature, etc.) must be taken into account. Concerning these parameters, the most frequent conservative approach is to select the limiting values, partly

I. Panka; M. Telbisz

2007-01-01

264

Minor Actinides-Loaded FBR Core Concept Suitable for the Introductory Period in Japan

NASA Astrophysics Data System (ADS)

According to the Japan's Framework for Nuclear Energy Policy(1), a basic scenario for fast breeder reactors (FBRs) is that they will be introduced on a commercial basis starting around 2050 replacing light water reactors (LWRs). During the FBR introduction period, the Pu from LWR spent fuel is used for FBR startup. Howerver, the FBR core loaded with this Pu has a larger burnup reactivity due to its larger isotopic content of Pu-241 than a core loaded with Pu from an FBR multi-recycling core. The increased burnup reactivity may reduce the cycle length of an FBR. We investigated, an FBR transitional core concept to confront the issues of the FBR introductory period in Japan. Core specifications are based on the compact-type sodium-cooled mixed oxide (MOX)-fueled core designed from the Japanese FBR cycle feasibility studies, because lower Pu inventory should be better for the FBR introductory period in view of its flexibility for the required reprocessing amount of LWR spent fuel to start up FBRs. The reference specifications were selected as follows. Output of 1500MWe and average discharge fuel burnup of about 150GWd/t. Minor Actinides (MAs) recovered from LWR spent fuels which provide Pu to startup FBRs are loaded to the initial loading fuels and exchanged fuels during few cycles until equilibrium. We made the MA content of the initial loading fuel four kinds like 0%, 3%, 4%, 5%. The average of the initial loading fuel is assumed to be 3%, and that of the exchange fuel is set as 5%. This 5% maximum of the MA content is based on the irradiation results of the experimental fast reactor Joyo. We evaluated the core performances including burnup characteristics and the reactivity coefficient and confirmed that transitional core from initial loading until equilibrium cycle with loaded Pu from LWR spent fuel performs similary to an FBR multi-recycling core.

Fujimura, Koji; Sasahira, Akira; Yamashita, Junichi; Fukasawa, Tetsuo; Hoshino, Kuniyoshi

265

The fast-neutron and photon space-energy distributions have been measured in an axially (1.25 m active height) and azimuthally (60 degree symmetry sector) shortened model of the WWER-1000 reactor assembled in the LR-0 experimental reactor. The space-energy distributions have been calculated with the stochastic code MCNP and the deterministic three-dimensional code TORT. Selected results are presented and discussed in the paper. This work has been done in the frame of the EU 5th FW project REDOS REDOS, Reactor Dosimetry: Accurate determination and benchmarking of radiation field parameters, relevant for reactor pressure vessel monitoring. EURATOM Programme, Call 2000/C 294/04). All geometry and material composition data of the model as well as the available experimental data were carefully checked and revised. PMID:16381689

Osmera, B; Cvachovec, F; Kyncl, J; Smutný, V

2005-01-01

266

NASA Astrophysics Data System (ADS)

We perform density functional theory calculations on a series of armchair and zigzag nanotubes of diameters less than 1 nm using the all-electron full-potential(-linearized)-augmented-plane-wave method. Emphasis is laid on the effects of curvature, the electron-beam orientation, and the inclusion of the core hole on the carbon electron-energy-loss K edge. The electron-energy-loss near-edge spectra of all the studied tubes show strong curvature effects compared to that of flat graphene. The curvature-induced ?-? hybridization is shown to have a more drastic effect on the electronic properties of zigzag tubes than on those of armchair tubes. We show that the core-hole effect must be accounted for in order to correctly reproduce electron-energy-loss measurements. We also find that the energy-loss near-edge spectra of these carbon systems are dominantly dipole selected and that they can be expressed simply as a proportionality with the local momentum projected density of states, thus portraying the weak energy dependence of the transition matrix elements. Compared to graphite, we report a reduction in the anisotropy as seen on the energy-loss near-edge spectra of carbon nanotubes.

Titantah, J. T.; Jorissen, K.; Lamoen, D.

2004-03-01

267

NASA Astrophysics Data System (ADS)

Interactions of eka-Hg (E112) and Hg atoms with small gold clusters were studied in the frame of the relativistic effective core potential model using the density functional theory (DFT) approach incorporating spin-dependent (magnetic) interactions. The choice of the exchange-correlation functional was based on a comparison of the results of DFT and large-scale coupled cluster calculations for E112Au and HgAu at the scalar relativistic level. A close similarity between the E112Aun and HgAun equilibrium structures was observed. The E112 binding energies on Aun are typically smaller than those for Hg by ca. 25%-32% and the equilibrium E112-Au separations are always slightly larger than their Hg-Au counterparts.

Rykova, E. A.; Zaitsevskii, A.; Mosyagin, N. S.; Isaev, T. A.; Titov, A. V.

2006-12-01

268

Interactions of eka-Hg (E112) and Hg atoms with small gold clusters were studied in the frame of the relativistic effective core potential model using the density functional theory (DFT) approach incorporating spin-dependent (magnetic) interactions. The choice of the exchange-correlation functional was based on a comparison of the results of DFT and large-scale coupled cluster calculations for E112Au and HgAu at the scalar relativistic level. A close similarity between the E112Aun and HgAun equilibrium structures was observed. The E112 binding energies on Aun are typically smaller than those for Hg by ca. 25%-32% and the equilibrium E112-Au separations are always slightly larger than their Hg-Au counterparts. PMID:17199333

Rykova, E A; Zaitsevskii, A; Mosyagin, N S; Isaev, T A; Titov, A V

2006-12-28

269

Development and preliminary verification of the 3D core neutronic code: COCO

As the recent blooming economic growth and following environmental concerns (China)) is proactively pushing forward nuclear power development and encouraging the tapping of clean energy. Under this situation, CGNPC, as one of the largest energy enterprises in China, is planning to develop its own nuclear related technology in order to support more and more nuclear plants either under construction or being operation. This paper introduces the recent progress in software development for CGNPC. The focus is placed on the physical models and preliminary verification results during the recent development of the 3D Core Neutronic Code: COCO. In the COCO code, the non-linear Green's function method is employed to calculate the neutron flux. In order to use the discontinuity factor, the Neumann (second kind) boundary condition is utilized in the Green's function nodal method. Additionally, the COCO code also includes the necessary physical models, e.g. single-channel thermal-hydraulic module, burnup module, pin power reconstruction module and cross-section interpolation module. The preliminary verification result shows that the COCO code is sufficient for reactor core design and analysis for pressurized water reactor (PWR). (authors)

Lu, H.; Mo, K.; Li, W.; Bai, N.; Li, J. [Reactor Design and Fuel Management Research Center, China Nuclear Power Technology Research Inst., 47F/A Jiangsu Bldg., Yitian Road, Futian District, Shenzhen (China)

2012-07-01

270

A Simplified Approach for Evaluation of the Burnup Potential of Alternative Fuels

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.

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

271

NASA Astrophysics Data System (ADS)

In order to investigate the effect on fuel thermophysical properties when adding americium and selected fission products to uranium-plutonium mixed oxide (MOX) fuel, simulated low decontamination MOX fuel with high burn-ups to 250 GWd/t, has been prepared and subjected to characterization tests, elastic moduli measurements and melting temperature measurement. Elastic moduli for the simulated low decontamination MOX fuel were almost the same level as fuel without americium and fission products and decrease in the moduli was slight with increasing simulated burn-up. The melting temperature of high burn-up, low decontamination MOX fuel may be estimated by using the findings on the effect of americium, plutonium addition and fission products accumulation.

Tanaka, Kosuke; Osaka, Masahiko; Miwa, Shuhei; Hirosawa, Takashi; Kurosaki, Ken; Muta, Hiroaki; Uno, Masayoshi; Yamanaka, Shinsuke

2012-01-01

272

Corrosion of high burn-up structured UO 2 fuel in presence of dissolved H 2

NASA Astrophysics Data System (ADS)

The influence of high burn-up structured material on UO 2 corrosion has been studied in an autoclave experiment. The experiment was conducted on spent fuel fragments with an average burn-up of 67 GWd/tHM. They were corroded in a simplified groundwater containing 33 mM dissolved H 2 for 502 days. All redox sensitive elements were reduced. The reduction continued until a steady-state concentration was reached in the leachate for U at 1.5 × 10 -10 M and for Pu at 7 × 10 -11 M. The instant release of Cs during the first 7 days was determined to 3.4% of the total inventory. However, the Cs release stopped after release of 3.5%. It was shown that the high burn-up structure did not enhance fuel corrosion.

Fors, P.; Carbol, P.; Van Winckel, S.; Spahiu, K.

2009-10-01

273

Identifying and bounding uncertainties in nuclear reactor thermal power calculations

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)

Phillips, J.; Hauser, E.; Estrada, H. [Cameron, 1000 McClaren Woods Drive, Coraopolis, PA 15108 (United States)

2012-07-01

274

International studies on burnup credit criticality safety by an OECD/NEA working group

The results and conclusions from a six-year study by an international benchmarking group in the comparison of computational methods for evaluating burnup credit in criticality safety analyses is presented. Approximately 20 participants from 12 countries have provided results for most problems. Four detailed benchmark problems for pressurized-water-reactor fuel have been completed. Results from work being finalized, addressing burnup credit for boiling-water-reactor fuel, are discussed, as well as planned activities for additional benchmarks, including mixed-oxide fuels, and other activities.

Brady, M.C. [Duke Engineering and Services, Inc., Richland, WA (United States); Okuno, H. [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan); DeHart, M.D. [Oak Ridge National Lab., TN (United States); Nouri, A. [Inst. de Protection et de Surete Nucleaire, Fontenay Aux Roses (France); Sartori, E. [Organization for Economic Cooperation and Development/Nuclear Energy Agency, Paris (France)

1998-11-01

275

NASA Astrophysics Data System (ADS)

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.

Hipp, J. R.; Encarnacao, A.; Ballard, S.; Young, C. J.; Phillips, W. S.; Begnaud, M. L.

2011-12-01

276

A mathematical model is proposed for calculating correction coefficients for neutron-measuring channels. The seven section channels (KNI-7 channels) monitor energy release in Russian RBMK-1000 reactors. KNI-7 channel lifetimes exceed the heat-releasing channel lifetimes in which they are installed. Therefore, KNI-7 channels are repositioned and require new correction factors. The calculation approach proposed accounts for the burnup history of direct charge detectors and heat-releasing channels by using values of the previously generated time-integrated detector current and energy production of heat-releasing channels. The computer program WIMS/DPZ was used to calculate the total correction factor as a product of two independent cofactors, and was found to perform satisfactorily. 4 refs., 3 figs.

Sokolov, A.P.; Saakov, E.S.; Garusov, Yu.V.; Chernikov, O.G.

1994-12-01

277

Burnup increase and Power Uprate - Operation history of KKL

The Leibstadt nuclear power plant in Switzerland? (KKL), a GE BWR\\/6 boiling water reactor with an up-rated thermal power of 3600 MW and a nominal net electrical output of 1145 W has been operated for more than 20 years. The core today consists of 648 modern 10x10 assemblies with part length rods which results in a power density of 32

G. Ledergerber; W. Kaufmann; A. Ritter; D. Greiner; Y. Parmar; R. Jacot-Guillarmod; J. Krouthen

2007-01-01

278

FRAP-T6 calculations of fuel-rod behavior during overpower transients. [PWR; BWR

The performance of the FRAP-T6 computer code in calculating fuel rod failure and fission gas release during overpower transient events was analyzed. Comparisons of the code's calculations with experiment data was used to determine the accuracy of the code in these two performance areas. First, the ability of the code to replicate observed failure trends as functions of power, ramp rate, hold time, burnup, pellet-cladding gap size, cladding thickness, and fuel density was examined. Then, the capability of the code's fission gas release model to duplicate experiment measurements of unfailed rods was tested at various burnups.

Chambers, R.; Resch, S.C.

1982-01-01

279

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 39 × 1026 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.

Tomoyuki Uwaba; Masahiro Ito; Kozo Katsuyama; Bruce J. Makenas; David W. Wootan; Jon Carmack

2011-05-01

280

Evaluation of Cross-Section Sensitivities in Computing Burnup Credit Fission Product Concentrations

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.

Gauld, I.C.

2005-08-12

281

A validated methodology for evaluating burnup credit in spent fuel casks

The concept of allowing reactivity credit for the transmuted state of spent fuel offers both economic and risk incentives. This paper presents a general overview of the technical work being performed in support of the US Department of Energy (DOE) program to resolve issues related to the implementation of burnup credit. An analysis methodology is presented along with information representing

M. C. Brady; T. L. Sanders

1991-01-01

282

Evaluation of Cross-Section Sensitivities in Computing Burnup Credit Fission Product Concentrations

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

I. C. Gauld; D. E. Mueller

2005-01-01

283

Reactor performance studies have been completed for conceptual plate designs and show that maintaining reactor performance while converting HFIR from high enriched to low enriched uranium (20 wt % 235U) fuel requires extending the end-of-life burnup value for HFIR fuel from the current nominal value of 2200 MWD to 2600 MWD. The current fuel fabrication procedure is discussed and changes

Trent Primm; David Chandler

2009-01-01

284

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)

Pusa, M.; Leppaenen, J. [VTT Technical Research Centre of Finland, P.O. Box 1000, FI-02044 VTT (Finland)

2012-07-01

285

The Dublin Core is a metadata element set intended to facilitate discovery of electronic resources. It was originally conceived for author-generated descriptions of Web resources, and the Dublin Core has attracted broad ranging international and interdisciplinary support. The cha...

286

High Burnup Dry Storage Cask Research and Development Project, Final Test Plan

EPRI is leading a project team to develop and implement the first five years of a Test Plan to collect data from a SNF dry storage system containing high burnup fuel.12 The Test Plan defined in this document outlines the data to be collected, and the storage system design, procedures, and licensing necessary to implement the Test Plan.13 The main goals of the proposed test are to provide confirmatory data14 for models, future SNF dry storage cask design, and to support license renewals and new licenses for ISFSIs. To provide data that is most relevant to high burnup fuel in dry storage, the design of the test storage system must mimic real conditions that high burnup SNF experiences during all stages of dry storage: loading, cask drying, inert gas backfilling, and transfer to the ISFSI for multi-year storage.15 Along with other optional modeling, SETs, and SSTs, the data collected in this Test Plan can be used to evaluate the integrity of dry storage systems and the high burnup fuel contained therein over many decades. It should be noted that the Test Plan described in this document discusses essential activities that go beyond the first five years of Test Plan implementation.16 The first five years of the Test Plan include activities up through loading the cask, initiating the data collection, and beginning the long-term storage period at the ISFSI. The Test Plan encompasses the overall project that includes activities that may not be completed until 15 or more years from now, including continued data collection, shipment of the Research Project Cask to a Fuel Examination Facility, opening the cask at the Fuel Examination Facility, and examining the high burnup fuel after the initial storage period.

none,

2014-02-27

287

The viscosity of the earth's core is probably the least well-known physical property of the earth. Miki [1952] gives an estimate, based on a theoretical calculation, that the dynamic viscosity lies between 10 - and 10 - poise. Malkus [1968] suggests the range 10 -' to 1 poise. Attenuation of S waves reflected from the core [Sato and Espinosa, 1967b;

Roger F. Gans

1972-01-01

288

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 the effects of extending fuel burnup on the subsequent handling, interim storage, and other operations (e.g., rod consolidation and shipping) associated with the back end of the fuel cycle. The results of this study will aid DOE and the nuclear industry in assessing the effects on waste management of extending the useful in-reactor life of nuclear fuel. The experience base with extended-burnup fuel is now substantial and projections for future use of extended-burnup fuel in domestic LWRs are positive. The basic performance and integrity of the fuel in the reactor has not been compromised by extending the burnup, and the potential limitations for further extending the burnup are not severe. 104 refs., 15 tabs.

Bailey, W.J.

1989-03-01

289

Evolution of First Cores in Rotating Molecular Cores

NASA Astrophysics Data System (ADS)

We investigate the effect of rotation on the star formation process quantitatively using axisymmetric numerical calculations. An adiabatic hydrostatic object (the so-called first core) forms in a contracting cloud core, after the central region becomes optically thick and continues to contract, driven by mass accretion onto it. The structure of a rotating first core is characterized by its total angular momentum Jcore and mass Mcore, both of which increase by accretion with time. We find that the first core evolves with a constant Jcore/M2core. Evolutionary paths of first cores can be classified into two types. In a slowly rotating core with Jcore/M2core<0.015G/(sqrt(2)ciso), where ciso and G represent the isothermal sound speed in the molecular cloud core and the gravitational constant, respectively, the core begins ``second collapse'' after the central density exceeds the H2 dissociation density. This is the same evolution as a standard scenario for a spherically symmetric, nonrotating core. On the other hand, a core with Jcore/M2core>0.015G/(sqrt(2)ciso) stops its contraction before the central density reaches the H2 dissociation density and does not begin the second collapse. These rapidly rotating first cores suffer from nonaxisymmetric instabilities, such as formation of massive spiral arms, deformation into a bar, or fragmentation. Although the rotating first cores have small average luminosities of Lcore=0.003-0.03(M?core/10-5 Msolar yr-1) Lsolar, assuming a constant mass accretion rate M?core. Their lifetimes last several thousand years or more, which is much longer than those expected for nonrotating clouds (~1000 yr). We expect that at least several percent of prestellar cores contain first cores as very low luminosity objects. Furthermore, we find a core with 0.012G/(sqrt(2)ciso)

Saigo, Kazuya; Tomisaka, Kohji

2006-07-01

290

McCARD for Neutronics Design and Analysis of Research Reactor Cores

NASA Astrophysics Data System (ADS)

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.

Shim, Hyung Jin; Park, Ho Jin; Kwon, Soonwoo; Seo, Geon Ho; Hyo Kim, Chang

2014-06-01

291

MCNP Simulation of Void Reactivity in a Simplified CANDU Core Sub-region

NASA Astrophysics Data System (ADS)

The Monte Carlo code MCNP with a continuous-energy ENDF/B-VI cross section library at the hot operating condition was used to determine the impact of the core environment on void reactivity in a sub-region of a simplified CANDU-6 core of 4 x 3 x 6 cell-size. The net (combined) impact of the adjuster rods, axial leakage and cell-to-cell radial leakage (due to fuel burnup variation in the core) was estimated to be between 1.44 ± 0.37 and 1.96 ± 0.39 mk (10-3k).

Rahnema, F.; Mosher, S.; Pitts, M.; Akhtar, P.; Serghiuta, D.

292

Degraded core modeling in MELCOR

A package of phenomenological models has been developed for the MELCOR code system to calculate the thermal response of structures in the core and lower plenum of an LWR during a severe accident. This package treats all important modes of heat transfer within the core, as well as oxidation, debris formation, and relocation of core and structural materials during melting, candling, and slumping. Comparison of MELCOR and MARCON calculations for the Browns Ferry BWR primary system shows many areas of agreement during the early stages of core heatup and oxidation, but very large differences at later times. Many of these differences are attributed to the effects of candling predicted by MELCOR and the lack of any mechanistic candling or debris relocation models in MARCON. The melting and slumping behavior calculated by MELCOR is in qualitative agreement with our current understanding of the processes involved.

Summers, R.M.

1986-01-01

293

The Challenges Associated with High Burnup and High Temperature for UO2 TRISO-Coated Particle Fuel

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.

David Petti; John Maki

2005-02-01

294

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.

Lanning, D.D.; Beyer, C.E.; Painter, C.L.

1997-12-01

295

Gamma-ray spectroscopy is an important nondestructive method for the qualification of irradiated nuclear fuels. Regarding research reactors, the main parameter required in the scope of such qualification is the average burnup of spent fuel elements. This work describes the measurement, using nondestructive gamma-ray spectroscopy, of the average burnup attained by Material Testing Reactor (MTR) fuel elements irradiated in the RP-10

Mariano Vela Mora; Alberto Gallardo Padilla; José Luis Castro Palomino; Luís Antônio Albiac Terremoto

2011-01-01

296

Electrolysis of Burnup-Simulated Uranium Nitride Fuels in LiCl-KCl Eutectic Melts

The electrochemical behavior of burnup-simulated uranium nitride fuels containing representative solid fission product elements, UN+Mo (Mo = 2.84 wt%), UN+Pd (Pd = 4.6 wt%) and (U, Nd)N (NdN = 8.0 wt%), was investigated in the molten LiCl-KCl eutectic salt with 0.54 wt% UCl3 in order to clarify the effects of fission products on the dissolution of actinide nitrides and the

Takumi SATOH; Takashi IWAI; Yasuo ARAI

2009-01-01

297

Advanced Corrosion-Resistant Zr Alloys for High Burnup and Generation IV Applications

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.

Arthur Motta; Yong Hwan Jeong; R.J. Comstock; G.S. Was; Y.S. Kim

2006-10-31

298

SENSITIVITY AND UNCERTAINTY ANALYSIS OF COMMERCIAL REACTOR CRITICALS FOR BURNUP CREDIT

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

Georgeta Radulescu; Don Mueller; John C Wagner

2009-01-01

299

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

Lars Olof JERNKVIST

2006-01-01

300

K-infinite trends with burnup, enrichment, and cooling time for BWR fuel assemblies

This report documents the work performed by ORNL for the Yucca Mountain project (YMP) M and O contractor, Framatome Cogema Fuels. The goal of this work was to obtain k{sub inf} values for infinite arrays of flooded boiling-water-reactor (BWR) fuel assemblies as a function of various burnup\\/enrichment and cooling-time combinations. These scenarios simulate expected limiting criticality loading conditions (for a

Broadhead

1998-01-01

301

Cladding metallurgy and fracture behavior during reactivity-initiated accidents at high burnup

High-burnup fuel failure during a reactivity-initiated accident has been the subject of safety-related concern. Because of wide variations in metallurgical and simulation test conditions, it has been difficult to understand the complex failure behavior from major tests in NSRR and CABRI reactors. In this paper, a failure model based on fracture toughness and microstructural characteristics is proposed in which fracture

H. M. Chung; T. F. Kassner

1996-01-01

302

SENSITIVITY COEFFICIENT GENERATION FOR A BURNUP CREDIT CASK MODEL USING TSUNAMI3D

The evolution of a complex criticality model for a burnup credit shipping cask to an accurate TSUNAMI-3D model for eigenvalue sensitivity coefficient generation is detailed in this paper. TSUNAMI-3D is a Monte Carlo-based eigenvalue sensitivity analysis sequence that was released with SCALE 5. In the criticality model, 32 fuel assemblies, each with 18 axial zones with differing depletion-dependent compositions, are

Donald E. Mueller; Bradley T. Rearden

303

K-Effective Trends with Burnup, Enrichment, and Pooling Time for BWR Fuel Assemblies

This report documents the work performed by ORNL for the Yucca Mountain Project (YMP) M and O contractor, Framatome Cogema Fuels. The goal of this work was to obtain k{sub inf} values for infinite arrays of flooded boiling-water-reactor (BWR) fuel assemblies as a function of various burnup\\/enrichment and cooling-time combinations. These scenarios simulate expected limiting criticality loading conditions (for a

Broadhead

1998-01-01

304

Development of an MCNP-tally based burnup code and validation through PWR benchmark exercises

The aim of this study is to evaluate the capabilities of a newly developed burnup code called BUCAL1. The code provides the full capabilities of the Monte Carlo code MCNP5, through the use of the MCNP tally information. BUCAL1 uses the fourth order Runge Kutta method with the predictor–corrector approach as the integration method to determine the fuel composition at

B. El Bakkari; T. El Bardouni; O. Merroun; Ch. El Younoussi; Y. Boulaich; E. Chakir

2009-01-01

305

Analysis of Burnup and Economic Potential of Alternative Fuel Materials in Thermal Reactors

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.

Oggianu, Stella Maris; No, Hee Cheon; Kazimi, Mujid S. [Massachusetts Institute of Technology (United States)

2003-09-15

306

Short-time creep and rupture tests on high burnup fuel rod cladding

NASA Astrophysics Data System (ADS)

Short-time creep and rupture tests were performed to assess the strain potential of cladding of high burnup fuel rods under conditions of dry storage. The tests comprised optimized Zry-4 cladding samples from fuel rods irradiated to burnups of up to 64 MWd/kg U and were carried out at temperatures of 573 and 643 K and at hoop stresses of about 400 and 600 MPa. The applied stresses were chosen to reach about 2% strain within an envisaged testing time of 3-4 days. The tests were followed by a low temperature phase at 423 K and 100 MPa to assess the long-term behaviour of the cladding ductility especially with regard to the higher hydrogen content in the cladding of the high burnup fuel. These tests showed that around 600 K, a uniform plastic strain of at least 2% is reached without cladding failure. The low temperature phase at 423 K for up to 5 days revealed no cladding failure under these conditions of reduced cladding ductility due to the increased hydrogen content.

Goll, W.; Spilker, H.; Toscano, E. H.

2001-03-01

307

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.

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

2011-01-01

308

Irradiation performance of PFBR MOX fuel after 112 GWd/t burn-up

NASA Astrophysics Data System (ADS)

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.

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

309

Rim structure formation and high burnup fuel behavior of large-grained UO 2 fuels

NASA Astrophysics Data System (ADS)

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.

Une, K.; Hirai, M.; Nogita, K.; Hosokawa, T.; Suzawa, Y.; Shimizu, S.; Etoh, Y.

2000-01-01

310

BURNUP OF FUEL IN WATER-MODERATED WATER-COOLED POWER REACTORS AND URANIUM WATER LATTICE EXPERIMENTS

The method and the results are reported of numerical calculations of ; fuel burning in ordinary water-cooled and -moderated reactors with a homogeneous ; core which use fresh slightly enriched fuel. The point of departure of the ; method is the notion of stationary conditions at which the process is maintained ; in the reactor as a result of a

S. M. Feinberg; E. S.. Antsiferov

1959-01-01

311

Dependence of Core and Extended Flux on Core Dominance Parameter for Radio Sources

NASA Astrophysics Data System (ADS)

Based on two extragalactic radio source samples, the core dominance parameter is calculated, and the correlations between the core/extended flux density and core dominance parameter are investigated. When the core dominance parameter is lower than unity, it is linearly correlated with the core flux density, but it is not correlated with the extended flux density. When the core dominance parameter is higher than unity, it is not correlated with the core flux density, but it is linearly correlated with the extended flux density. Therefore, there are different results from different samples. The results can be explained using a relativistic beaming model.

Nie, J. J.; Yang, J. H.

2015-01-01

312

Fuel Breeding and Core Behavior Analyses on In Core Fuel Management of Water Cooled Thorium Reactors

NASA Astrophysics Data System (ADS)

Thorium fuel cycle with recycled U-233 has been widely recognized having some contributions to improve the water-cooled breeder reactor program which has been shown by a feasible area of breeding and negative void reactivity which confirms that fissile of 233U contributes to better fuel breeding and effective for obtaining negative void reactivity coefficient as the main fissile material. The present study has the objective to estimate the effect of whole core configuration as well as burnup effects to the reactor core profile by adopting two dimensional model of fuel core management. About more than 40 months of cycle period has been employed for one cycle fuel irradiation of three batches fuel system for large water cooled thorium reactors. All position of fuel arrangement contributes to the total core conversion ratio which gives conversion ratio less than unity of at the BOC and it contributes to higher than unity (1.01) at the EOC after some irradiation process. Inner part and central part give the important part of breeding contribution with increasing burnup process, while criticality is reduced with increasing the irradiation time. Feasibility of breeding capability of water-cooled thorium reactors for whole core fuel arrangement has confirmed from the obtained conversion ratio which shows higher than unity. Whole core analysis on evaluating reactivity change which is caused by the change of voided condition has been employed for conservative assumption that 100% coolant and moderator are voided. It obtained always a negative void reactivity coefficient during reactor operation which shows relatively more negative void coefficient at BOC (fresh fuel composition), and it becomes less negative void coefficient with increasing the operation time. Negative value of void reactivity coefficient shows the reactor has good safety properties in relation to the reactivity profile which is the main parameter in term of criticality safety analysis. Therefore, this evaluation has confirmed that breeding condition and negative coefficient can be obtained simultaneously for water-cooled thorium reactor obtains based on the whole core fuel arrangement.

Permana, Sidik; Sekimoto, Hiroshi; Waris, Abdul; Subhki, Muhamad Nurul; Ismail

2010-12-01

313

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.

314

Viscosity of the earth's core.

NASA Technical Reports Server (NTRS)

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.

Gans, R. F.

1972-01-01

315

NASA Astrophysics Data System (ADS)

The performance of MOX fuel irradiated in the advanced thermal reactor, FUGEN, to a burnup of 47.5 GWd/t, was investigated by using a telescope, optical microscope, SEM and EPMA. Observations focused on elucidating the corrosion behavior of the cladding inner surface. A reaction layer was observed at burnups higher than about 35 GWd/t. The relationship between the thickness of the reaction layer and burnup was similar to that reported in the literature for conventional UO 2 fuel and other MOX fuels. The existence of a plutonium spot near the outer surface of the fuel pellet had no significant effect on the thickness of the reaction layer. A bonding layer was observed on the cladding inner surface. Its morphology and elemental distributions were not so different from those in BWR UO 2 fuel rods irradiated to high burnup, in which the fission gas release rate is high. In addition, the dependences of bonding layer formation on the burnup and linear heat rating were similar to results of UO 2 fuel rods. It was, thus, suggested that the bonding layer formation mechanism was similar in both UO 2 and MOX fuel rods.

Tanaka, Kosuke; Maeda, Koji; Sasaki, Shinji; Ikusawa, Yoshihisa; Abe, Tomoyuki

2006-10-01

316

The inert matrix fuels are a promising option to reduce-eliminate worldwide plutonium stockpiles by burning it in LWRs. These fuels, where plutonium is hosted in a U-free inert matrix phase, may reach high burning efficiency while preventing new plutonium build-up under irradiation. A specific investigation on CSZ and thoria inert matrices has been developed by ENEA since several years. In-pile testing on the ENEA-conceived innovative fuels is ongoing in the OECD Halden HBWR since June 2000 (IFA-652 experiment). The registered burnup at the end of 2005 is about 38 MWd.kgU{sub eq}{sup -1} vs. 45 MWd.kgU{sub eq}{sup -1} (40 MWd.kgUOX{sub eq}{sup -1}) target. Fuel pins are equipped with fuel temperature thermocouples, internal pressure transducers and fuel stack elongation sensors, with the task of studying thermal conductivity and its degradation with burnup, densification-swelling behaviour and the FGR. In this paper, the response at low burnup (< 7 MWd.kgU{sub eq}{sup -1}) of CSZ-based fuels loaded in IFA-652, is analysed by means of the TRANSURANUS code. To this purpose, a comprehensive modelling of the above mentioned un-irradiated fuels, mainly relying on the thermophysical characterisation performed at the JRC/ITU-Karlsruhe, has been implemented in a custom TRANSURANUS version (TU-IMF). A comparison of the code predictions vs. the experimental data, aimed at evaluating the early-stage under irradiation phenomena, particularly densification and relocation, has been performed. (authors)

Calabrese, R.; Vettraino, F. [ENEA, Nuclear Fission Division, via Martin di Monte Sole 4, 40129 Bologna (Italy); Tverberg, T. [OECD Halden Reactor Project, Institutt for energiteknikk, P.O. Box 175, N-1751 Halden (Norway)

2006-07-01

317

Analyses of Greek Research Reactor with mixed HEU-LEU Be reflected core

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.

Deen, J.R.; Snelgrove, J.L. [Argonne National Lab., IL (United States); Papastergiou, K. [National Center for Scientific Research, Athens (Greece)

1993-12-31

318

AGR 1 was the first in a series of experiments designed to test US TRISO fuel under high temperature gas-cooled reactor irradiation conditions. This experiment was irradiated in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) and is currently undergoing post irradiation examination (PIE) at INL and Oak Ridge National Laboratory. One component of the AGR 1 PIE is the experimental evaluation of the burnup of the fuel by two separate techniques. Gamma spectrometry was used to non destructively evaluate the burnup of all 72 of the TRISO fuel compacts that comprised the AGR 1 experiment. Two methods for evaluating burnup by gamma spectrometry were developed, one based on the Cs 137 activity and the other based on the ratio of Cs 134 and Cs 137 activities. Burnup values determined from both methods compared well with the values predicted from simulations. The highest measured burnup was 20.1 %FIMA for the direct method and 20.0 %FIMA for the ratio method (compared to 19.56% FIMA from simulations). An advantage of the ratio method is that the burnup of the cylindrical fuel compacts can determined in small (2.5 mm) axial increments and an axial burnup profile can be produced. Destructive chemical analysis by inductively coupled mass spectrometry (ICP MS) was then performed on selected compacts that were representative of the expected range of fuel burnups in the experiment to compare with the burnup values determined by gamma spectrometry. The compacts analyzed by mass spectrometry had a burnup range of 19.3 % FIMA to 10.7 % FIMA. The mass spectrometry evaluation of burnup for the four compacts agreed well with the gamma spectrometry burnup evaluations and the expected burnup from simulation. For all four compacts analyzed by mass spectrometry, the maximum range in the three experimentally determined values and the predicted value was 6% or less. The results confirm the accuracy of the nondestructive burnup evaluation from gamma spectrometry for TRISO fuel compacts across a burnup range of approximately 10 to 20 % FIMA and also validate the approach used in the physics simulation of the AGR 1 experiment.

Jason M. Harp; Paul A. Demkowicz; Phillip L. Winston; James W. Sterbentz

2014-10-01

319

NASA Technical Reports Server (NTRS)

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%.

Davison, H. W.; Fiero, I. B.

1971-01-01

320

Group structures for RBMK cell calculations

The Chernobyl accident has led to international interest in calculations on RBMK reactors. In such calculations it is natural to start by calculating the reactor physics properties of a single-channel cell. The void coefficient is very sensitive to the group structure used in this calculation, since it is determined by several phenomena acting in opposite directions, which makes accurate calculation of the net effect difficult. To study the adequacy of various group structures, the authors carried out calculations for 2% enriched fuel with a burnup of 10.3 MWd/kg U at void fractions of 0, 40, 80, and 100%. As candidate group structures, schemes with 22, 29, and 36 macrogroups were used. Reference calculations had 60 and 49 macrogroups, respectively, with the latter scheme avoiding group condensation between 4 and 0.625 eV, where some condensation was necessary in the former. The resulting void coefficients for K/sub infinity/ and k/sub eff/ according to the fundamental mode calculation are shown. The 22-group scheme is unsatisfactory. The 29-group results are acceptable, but the 36-group structure gives substantially better accuracy at moderate extra cost. Naturally, even the 60-group reference calculation contains errors arising from energy discretization, but it seems likely that with the 36-group structure, energy discretization will no longer be a dominant source of errors.

Wasastjerna, F.

1987-01-01

321

NASA Astrophysics Data System (ADS)

The goal of this project was to develop the best available non-destructive technique to determine burnup of the Advanced Test Reactor (ATR) fuels at Idaho National Laboratory, as well as to make a recommendation regarding the feasibility of implementing a permanent fuel scanning system at the ATR canal. The study determined that useful spectra for validation and fuel burnup predictions can be obtained in-situ at the ATR canal using three different detectors. In addition, the study established that calibration curves can be created to predict ATR fuel burnup onsite. The study also established that in order to design a rugged system that can stand the daily operations at the ATR canal a LaBr3 scintillator can be used effectively if deconvolution process is applied to increase the spectra resolution.

Navarro, J.; Ring, T. A.; Nigg, D. W.

2014-04-01

322

Thermal diffusivity of homogeneous SBR MOX fuel with a burn-up of 35 MWd/kgHM

NASA Astrophysics Data System (ADS)

The effect of burn-up on the thermal conductivity of homogeneous SBR MOX fuel is investigated and compared with standard UO 2 LWR fuel. 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. The fuel thermal conductivity was found to be in the same range as for UO 2 of similar burn-up. The annealing behaviour was characterized in order to identify the degradation due to the out-of-pile auto-irradiation.

Cozzo, C.; Staicu, D.; Pagliosa, G.; Papaioannou, D.; Rondinella, V. V.; Konings, R. J. M.; Walker, C. T.; Barker, M. A.; Hervé, P.

2010-05-01

323

NASA Astrophysics Data System (ADS)

Structural analyses of dry casks containing high-burnup fuel require cladding mechanical properties and failure limits to assess fuel behavior. Pre-storage drying-transfer operations and early stage storage subject cladding to higher temperatures and much higher pressure-induced tensile hoop stresses relative to in-reactor operation and pool storage. Under these conditions, radial hydrides may precipitate during slow cooling and provide an additional embrittlement mechanism as the cladding temperature decreases below the ductile-to-brittle transition temperature (DBTT). A test procedure was developed to simulate the effects of drying-storage temperature histories. Following drying-storage simulation, samples were subjected to ring-compression test (RCT) loading, which was used as a ductility screening test and to simulate pinch-type loading that may occur during cask transport. RCT samples with <2% offset strain prior to >50% wall cracking were assessed as brittle. Prior to testing high-burnup cladding, many tests were conducted with pre-hydrided Zircaloy-4 (Zry-4) and ZIRLO™ to determine target 400 °C hoop stresses for high-burnup rodlets. Zry-4 cladding segments, from a 67-GWd/MTU fuel rod, with 520-620 wppm hydrogen and ZIRLO™ cladding segments from a 70-GWd/MTU fuel rod, with 350-650 wppm hydrogen were defueled and tested. Following drying-storage simulation, the extent of radial-hydride precipitation was characterized by the radial-hydride continuity factor. It was found that the DBTT was dependent on: cladding material, irradiation conditions, and drying-storage histories (stress at maximum temperature). High-burnup ZIRLO™ exhibited higher susceptible to radial-hydride formation and embrittlement than high-burnup Zry-4. It was also observed that uniformly pre-hydrided, non-irradiated cladding was not a good surrogate for high-burnup cladding because of the high density of circumferential hydrides across the wall and the high metal-matrix ductility for pre-hydrided cladding.

Billone, M. C.; Burtseva, T. A.; Einziger, R. E.

2013-02-01

324

Multistate searching methods are a subfield of distributed artificial intelligence that aims to provide both principles for construction of complex systems involving multiple states and mechanisms for coordination of independent agents' actions. This paper proposes a multistate searching algorithm with reinforcement learning for the automatic core design of a boiling water reactor. The characteristics of this algorithm are that the coupling structure and the coupling operation suitable for the assigned problem are assumed and an optimal solution is obtained by mutual interference in multistate transitions using multiagents. Calculations in an actual plant confirmed that the proposed algorithm increased the convergence ability of the optimization process.

Kobayashi, Yoko; Aiyoshi, Eitaro

2005-07-15

325

Determination of high burn-up nuclear fuel elastic properties with acoustic microscopy

NASA Astrophysics Data System (ADS)

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.

Laux, D.; Baron, D.; Despaux, G.; Kellerbauer, A. I.; Kinoshita, M.

2012-01-01

326

Isotopic ratio measurements of Cs-135 to Cs-137 Were performed using both resonance ionization mass spectrometry (RIMS) and thermal ionization mass spectrometry (TIMS) to determine the chronological age of nuclear fuel burn-up samples. Initial measurements on a lake sediment sample are being performed at NIST for determination of cesium content in the sample. Atomization behavior of the graphite furnace source, the overall efficiency and selectivity were measured for different sample preparations. Single-resonance excitation 6s S-2(1/2) (F = 4) --> 6p P-2(3/2) (F' = 5) with an extended cavity diode laser followed by photoionization with the 488 nm line of an argon ion laser yielded optical selectivity for Cs-135 and Cs-137 of more than two orders of magnitude against stable Cs-133 and overall selectivity of 10(8). An overall efficiency of 5 x 10(-7) was measured for standard Cs-133 solutions and for the nuclear fuel burn-up samples.

Pibida, L.; Mcmahon, C. A.; Bushaw, Bruce A.

2004-04-15

327

EBSD and TEM Characterization of High Burn-up Mixed Oxide Fuel

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.

Melissa C Teague [Idaho National Laboratory (INL), Idaho Falls, ID (United States); Brian P. Gorman [Colorado School of Mines, Golden, CO (United States); Brandon D Miller [Idaho National Laboratory (INL), Idaho Falls, ID (United States); Jeffrey King [Colorado School of Mines, Golden, CO (United States)

2014-01-01

328

Thermal conductivity of homogeneous and heterogeneous MOX fuel with up to 44 MWd/kgHM burn-up

NASA Astrophysics Data System (ADS)

New thermal diffusivity data for homogeneous SBR and heterogeneous MIMAS and OCOM MOX fuels are reported. No significant difference between the thermal diffusivity of the homogeneous and heterogeneous fuels was found at the burn-up up to 44 MWd/kgHM. These measurements, combined with previously published results or correlation functions for irradiated UO 2 and MOX were compared and it was found that separate correlations for these two fuels are not justified. A correlation for the thermal conductivity of irradiated UO 2 and MOX as a function of burn-up and irradiation temperature is proposed.

Staicu, D.; Cozzo, C.; Pagliosa, G.; Papaioannou, D.; Bremier, S.; Rondinella, V. V.; Walker, C. T.; Sasahara, A.

2011-05-01

329

Modified Laser and Thermos cell calculations on microcomputers

In the course of designing and operating nuclear reactors, many fuel pin cell calculations are required to obtain homogenized cell cross sections as a function of burnup. In the interest of convenience and cost, it would be very desirable to be able to make such calculations on microcomputers. In addition, such a microcomputer code would be very helpful for educational course work in reactor computations. To establish the feasibility of making detailed cell calculations on a microcomputer, a mainframe cell code was compiled and run on a microcomputer. The computer code Laser, originally written in Fortran IV for the IBM-7090 class of mainframe computers, is a cylindrical, one-dimensional, multigroup lattice cell program that includes burnup. It is based on the MUFT code for epithermal and fast group calculations, and Thermos for the thermal calculations. There are 50 fast and epithermal groups and 35 thermal groups. Resonances are calculated assuming a homogeneous system and then corrected for self-shielding, Dancoff, and Doppler by self-shielding factors. The Laser code was converted to run on a microcomputer. In addition, the Thermos portion of Laser was extracted and compiled separately to have available a stand alone thermal code.

Shapiro, A.; Huria, H.C.

1987-01-01

330

NSDL National Science Digital Library

The gravity calculator calculates the gravitational force between two masses. Also included is a visualization of the typical measurement of gravitational force (weight) in different environments (stationary and free fall).

Brendan Cannell, Ronnie Johnson, The Shodor Education Foundation, Inc.

331

NSDL National Science Digital Library

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.

Birnbaum, Michael H.

332

In accordance with the need to determine whether cracking of the ceramic core disks which will be constructed and used in the High Temperature Test Facility (HTTF) for heatup and cooldown experiments, a set of calculation were performed using Abaqus to investigate the thermal stresses levels and likelihood for cracking. The calculations showed that using the material properties provided for the Greencast 94F ceramic, cracking is predicted to occur. However, this modeling does not predict the size or length of the actual cracks. It is quite likely that cracks will be narrow with rough walls which would impede the flow of coolant gases entering the cracks. Based on data recorded at Oregon State University using Greencast 94F samples that were heated and cooled at prescribed rates, it was concluded that the likelihood that the cracks would be detrimental to the experimental objectives is small.

Brian D. Hawkes; Richard Schultz

2012-07-01

333

Improved gas core propulsion model

A thermodynamic, radiation transport model of a gas core nuclear propulsion reactor has been developed in one-dimensional, spherical geometry, which satisfies local energy balance and allows for arbitrary variation of fuel/propellant ratio and flow rate as functions of radius. Initial cases calculated yield specific impulses of about 1150 sec, but very low thrusts ranging 5--10 kN.

Tanner, J.E.

1993-10-01

334

Modelling Characteristics of Ferromagnetic Cores with the Influence of Temperature

NASA Astrophysics Data System (ADS)

The paper is devoted to modelling characteristics of ferromagnetic cores with the use of SPICE software. Some disadvantages of the selected literature models of such cores are discussed. A modified model of ferromagnetic cores taking into account the influence of temperature on the magnetizing characteristics and the core losses is proposed. The form of the elaborated model is presented and discussed. The correctness of this model is verified by comparing the calculated and the measured characteristics of the selected ferromagnetic cores.

Górecki, K.; Rogalska, M.; Zar?bski, J.; Detka, K.

2014-04-01

335

NASA Technical Reports Server (NTRS)

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.

Voorhies, C. V.

1999-01-01

336

Gas core reactors for actinide transmutation and breeder applications

NASA Technical Reports Server (NTRS)

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.

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

1978-01-01

337

NSDL National Science Digital Library

A calculator that estimates the heating needs of a room, a combination of rooms, or an entire home. Enter the length and width of the area to be heated and select the climate and insulation factors from the pop-up boxes. A Java version of this calculator is also available.

338

After the end of a neutral-beam injection pulse into a low-density TFTR plasma, once the beam-injected deuterons have thermalized, the neutron emission is dominated by the 14-MeV neutron production from D-D triton burnup. Ordinary fission detectors can measure the 14-MeV emission rate, which can be extrapolated back in time to estimate the equilibrium triton burnup fraction. The fractional burnup determined by this method is in the range of 0.3 to 1.5% for TFTR discharges to date, and is consistent with classical confinement and slowing down. 10 refs., 3 figs.

Jassby, D.L.; Hendel, H.W.; Barnes, C.W.; Bosch, S.; Cecil, F.E.; McCune, D.C.; Nieschmidt, E.B.; Strachan, J.D.

1987-06-01

339

Benchmark data for validating irradiated fuel compositions used in criticality calculations

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.

Bierman, S.R.; Talbert, R.J.

1994-10-01

340

NSDL National Science Digital Library

Martindale Calculators is a Web-based tool collection that contains over 19,000 online calculators created by over "3,450" very "creative" individuals, businesses and Ã¢ÂÂtax supported entities world wide.Ã¢ÂÂ The collection is organized by the following topics: mathematics; statistics; science A-Z; chemistry; physics, astrophysics and astronomy; engineering A-Z; and electrical engineering, computer engineering, & computer science. Each section includes a wealth of websites to explore, all related to mathematical calculations, mostly course materials and articles. Another section lists online calculators relevant for various industries, such as aviation, cosmetics, insurance, and library science. The list is organized alphabetically and creatively stretches the meaning of Ã¢ÂÂcalculatorÃ¢ÂÂ to include such things as name translators and databases on animal breeds.

341

Start-up fuel and power flattening of sodium-cooled candle core

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.

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

342

Benchmarking the parallel FACETS core solver

NASA Astrophysics Data System (ADS)

The Framework Architecture for Core-Edge Transport Simulations (FACETS) is a SciDAC project targeting whole-device plasma simulations in tokamaks such as ITER. A key component in the multi-physics FACETS effort has been the development of a core transport solver (FACETS::core) that is both robust and runs in parallel. FACETS::core can interface to any of the flux calculators available through the Framework for Modernization and Componentization Fusion Modules (FMCFM), including GLF23 and MMM95. Electron and ion temperatures are advanced implicitly using the nonlinear fluxes from GLF23 (or other model). Here, we present results comparing the stability and accuracy of FACETS::core with the ASTRA transport code. Although FACET::core is slower than ASTRA on a per time step basis, the multigrid algorithm and PETSc/SNES solver applied by FACETS::core allow the latter to take orders of magnitude larger time steps, conferring to FACETS::core a 5-10x overall performance improvement over ASTRA. This, combined with the capability of FACETS::core to scale to tens of processors, contributes towards a wall clock time reduction of the core transport computation by a factor 200-500x.

Pletzer, Alexander; Hakim, Ammar; Miah, Mahmood; Cary, John; Kruger, Scott; Vadlamani, Srinath; Pankin, Alexei

2008-11-01

343

NASA Astrophysics Data System (ADS)

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.

Nur Asiah, A.; Su'ud, Zaki; Ferhat, A.; Sekimoto, H.

2010-06-01

344

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.

Nur Asiah, A.; Su'ud, Zaki [Nuclear Physics and Biophysics Research Division, Faculty of Mathematics and Natural Sciences, Institut Teknologi Bandung (Indonesia); Ferhat, A. [National Nuclear Energ Agency of Indonesia (BATAN) (Indonesia); Sekimoto, H. [Research Laboratory for Nuclear Reactors, Tokyo Institute of Technology (Japan)

2010-06-22

345

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.

Lee, C. H.; Zhong, Z.; Taiwo, T.A.; Yang, W.S.; Khalil, H.S.; Smith, M.A.; Nuclear Engineering Division

2006-10-13

346

Burn up calculations for the Iranian miniature reactor: A reliable and safe research reactor

Presenting neutronic calculations pertaining to the Iranian miniature research reactor is the main goal of this article. This is a key to maintaining safe and reliable core operation. The following reactor core neutronic parameters were calculated: clean cold core excess reactivity (?ex), control rod and shim worth, shut down margin (SDM), neutron flux distribution of the reactor core components, and

F. Faghihi; S. M. Mirvakili

2009-01-01

347

On the oxidation state of UO 2 nuclear fuel at a burn-up of around 100 MWd/kgHM

NASA Astrophysics Data System (ADS)

Results for the radial distribution of the oxygen potential and stoichiometry of a PWR fuel with an average pellet burn-up of 102 MWd/kgHM are presented. The local ?G¯(O2) of the fuel was measured using a miniature solid state galvanic cell, the local O/U ratio was calculated from the lattice parameter measured by micro-X-ray diffraction and the local O/M ratio was derived from the fuel composition determined by ICP-MS. During irradiation the O/U ratio of the fuel decreased from 2.005 to 1.991 ± 0.008. The average fuel O/M ratio was 1.973 compared with the stoichiometric value of 1.949. The amount of free oxygen in the fuel, represented by the difference between these two quantities, increased from the centre to periphery of the pellet. Similarly, the ?G¯(O2) of the fuel increased from -370 kJ mol -1 at r/ r0 = 0.1 to -293 kJ mol -1 at r/ r0 = 0.975. Thus, the ?G¯(O2) of the fuel had not been buffered by the oxidation of fission product Mo. About one-quarter of the free oxygen accumulated during the irradiation had been gettered by the Zircaloy cladding.

Walker, C. T.; Rondinella, V. V.; Papaioannou, D.; Van Winckel, S.; Goll, W.; Manzel, R.

2005-10-01

348

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.

Broadhead, B.L.

1991-08-01

349

One of my passions in life is to try and understand how we have developed the wonderful calculating ability we currently possess. Anyone with this hobby will undoubtedly start by looking back to see how the first PCs were developed, then progress back to older \\

Michael R. Williams

2004-01-01

350

NSDL National Science Digital Library

This interactive calculator produced by Teachers' Domain helps you determine the mercury levels in various types of fish, and enables you to make more informed choices about which fish are safe to eat and which should be avoided or eaten infrequently.

WGBH Educational Foundation

2010-12-23

351

NSDL National Science Digital Library

This web site, which is part of the NCTM Illuminations project, allows students to challenge themselves or opponents from anywhere in the world by playing games that are organized around content from the upper elementary and middle grades math curriculum. The games allow students to learn about fractions, factors, multiples, symmetry, as well as practice important skills like basic multiplication and calculating area.

2011-01-01

352

Rotating cardboard discs are used to read off total tree or topwood firewood volume (tons or cords) that can be expected from trees of d.b.h. 6 to 24 inches and tree height 10 to 90 feet. One side of the calculator is used for broadleaved species with deliquescent crowns and the other side for braodleaves with excurrent crowns.

Clark, A.; Curtis, A.B.; Darwin, W.N.

1981-01-01

353

SEM Characterization of the High Burn-up Microstructure of U-7Mo Alloy

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.

Dennis D. Keiser, Jr.; Jan-Fong Jue; Jian Gan; Brandon Miller; Adam Robinson; Pavel Medvedev; James Madden; Dan Wachs; M. Teague

2014-04-01

354

In-Situ Safeguards Verification of Low Burn-up Pressurized Water Reactor Spent Fuel Assemblies

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.

Ham, Y S; Sitaraman, S; Park, I; Kim, J; Ahn, G

2008-04-16

355

TEM Characterization of High Burn-up Microstructure of U-7Mo Alloy

As an essential part of global nuclear non-proliferation effort, the RERTR program is developing low enriched U-Mo fuels (< 20% U-235) for use in research and test reactors that currently employ highly enriched uranium fuels. One type of fuel being developed is a dispersion fuel plate comprised of U-7Mo particles dispersed in Al alloy matrix. Recent TEM characterizations of the ATR irradiated U-7Mo dispersion fuel plates include the samples with a local fission densities of 4.5, 5.2, 5.6 and 6.3 E+21 fissions/cm3 and irradiation temperatures of 101-136?C. The development of the irradiated microstructure of the U-7Mo fuel particles consists of fission gas bubble superlattice, large gas bubbles, solid fission product precipitates and their association to the large gas bubbles, grain subdivision to tens or hundreds of nanometer size, collapse of bubble superlattice, and amorphisation. This presentation will describe the observed microstructures specifically focusing on the U-7Mo fuel particles. The impact of the observed microstructure on the fuel performance and the comparison of the relevant features with that of the high burn-up UO2 fuels will be discussed.

Jian Gan; Brandon Miller; Dennis Keiser; Adam Robinson; James Madden; Pavel Medvedev; Daniel Wachs

2014-04-01

356

NSDL National Science Digital Library

This interactive applet helps students develop fluency and flexibility with numbers. At each of 6 difficulty levels the user is presented with 8 target numbers and a partial set of keys on a basic calculator (does not follow order of operations). The goal is to use the given keys to make as many of the target numbers as possible within the 3-minute time limit. Some levels include memory keys.

Mandy Barrow

2008-01-01

357

NASA Astrophysics Data System (ADS)

The goal of this study presented is to determine the best available nondestructive technique necessary to collect validation data as well as to determine burnup and cooling time of the fuel elements on-site at the Advanced Test Reactor (ATR) canal. This study makes a recommendation of the viability of implementing a permanent fuel scanning system at the ATR canal and leads to the full design of a permanent fuel scan system. The study consisted at first in determining if it was possible and which equipment was necessary to collect useful spectra from ATR fuel elements at the canal adjacent to the reactor. Once it was establish that useful spectra can be obtained at the ATR canal, the next step was to determine which detector and which configuration was better suited to predict burnup and cooling time of fuel elements nondestructively. Three different detectors of High Purity Germanium (HPGe), Lanthanum Bromide (LaBr3), and High Pressure Xenon (HPXe) in two system configurations of above and below the water pool were used during the study. The data collected and analyzed were used to create burnup and cooling time calibration prediction curves for ATR fuel. The next stage of the study was to determine which of the three detectors tested was better suited for the permanent system. From spectra taken and the calibration curves obtained, it was determined that although the HPGe detector yielded better results, a detector that could better withstand the harsh environment of the ATR canal was needed. The in-situ nature of the measurements required a rugged fuel scanning system, low in maintenance and easy to control system. Based on the ATR canal feasibility measurements and calibration results, it was determined that the LaBr3 detector was the best alternative for canal in-situ measurements; however, in order to enhance the quality of the spectra collected using this scintillator, a deconvolution method was developed. Following the development of the deconvolution method for ATR applications, the technique was tested using one-isotope, multi-isotope, and fuel simulated sources. Burnup calibrations were perfomed using convoluted and deconvoluted data. The calibrations results showed burnup prediction by this method improves using deconvolution. The final stage of the deconvolution method development was to perform an irradiation experiment in order to create a surrogate fuel source to test the deconvolution method using experimental data. A conceptual design of the fuel scan system is path forward using the rugged LaBr 3 detector in an above the water configuration and deconvolution algorithms.

Navarro, Jorge

358

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)

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

359

The goal of this study presented is to determine the best available non-destructive technique necessary to collect validation data as well as to determine burn-up and cooling time of the fuel elements onsite at the Advanced Test Reactor (ATR) canal. This study makes a recommendation of the viability of implementing a permanent fuel scanning system at the ATR canal and leads3 to the full design of a permanent fuel scan system. The study consisted at first in determining if it was possible and which equipment was necessary to collect useful spectra from ATR fuel elements at the canal adjacent to the reactor. Once it was establish that useful spectra can be obtained at the ATR canal the next step was to determine which detector and which configuration was better suited to predict burnup and cooling time of fuel elements non-destructively. Three different detectors of High Purity Germanium (HPGe), Lanthanum Bromide (LaBr3), and High Pressure Xenon (HPXe) in two system configurations of above and below the water pool were used during the study. The data collected and analyzed was used to create burnup and cooling time calibration prediction curves for ATR fuel. The next stage of the study was to determine which of the three detectors tested was better suited for the permanent system. From spectra taken and the calibration curves obtained, it was determined that although the HPGe detector yielded better results, a detector that could better withstand the harsh environment of the ATR canal was needed. The in-situ nature of the measurements required a rugged fuel scanning system, low in maintenance and easy to control system. Based on the ATR canal feasibility measurements and calibration results it was determined that the LaBr3 detector was the best alternative for canal in-situ measurements; however in order to enhance the quality of the spectra collected using this scintillator a deconvolution method was developed. Following the development of the deconvolution method for ATR applications the technique was tested using one-isotope, multi-isotope and fuel simulated sources. Burnup calibrations were perfomed using convoluted and deconvoluted data. The calibrations results showed burnup prediction by this method improves using deconvolution. The final stage of the deconvolution method development was to perform an irradiation experiment in order to create a surrogate fuel source to test the deconvolution method using experimental data. A conceptual design of the fuel scan system is path forward using the rugged LaBr3 detector in an above the water configuration and deconvolution algorithms.

Jorge Navarro

2013-12-01

360

Thermal property change of MOX and UO2 irradiated up to high burnup of 74 GWd/t

NASA Astrophysics Data System (ADS)

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.

Nakae, Nobuo; Akiyama, Hidetoshi; Miura, Hiromichi; Baba, Toshikazu; Kamimura, Katsuichiro; Kurematsu, Shigeru; Kosaka, Yuji; Yoshino, Aya; Kitagawa, Takaaki

2013-09-01

361

Neutron transport and diffusion theory space- and energy-dependent reactor kinetics calculations were performed in the frequency domain to determine the sensitivity of an ex-core neutron detector to in-core vibrations and coolant boiling in a PWR. Results of these calculations indicate that the ex-core detectors are sensitive to neutron sources, to vibrations, and to boiling occurring over large regions of the

F. J. Sweeney; J. P. A. Renier

1984-01-01

362

Sensitivity of Candu-Scwr Reactor Physics Calculations to Nuclear Data Files

NASA Astrophysics Data System (ADS)

A simplified MCNP model of a CANDU-SCWR lattice was used to test the sensitivity of the calculated reactivity to various nuclear data files involving issues of current interest. These tests were performed for cooled and voided conditions, with and without axial neutron leakage, for a uniform lattice of mid-life fuel and a mixed lattice with high-burnup and low-burnup fuel in alternate channels. Results were compared using different room-temperature data files for deuterium, various thermal-scattering-law data files for hydrogen bound in light water and deuterium bound in heavy water, and for pre-ENDF/B-VII and ENDF/B-VI.8 data for uranium. The reactivity differences observed were small (typically <1 mk) and increased with axial neutron leakage.

Kozier, K. S.; Dyck, G. R.

2006-04-01

363

NSDL National Science Digital Library

Jarmo Lammi has developed this simple, easy-to-use tool that provides information useful for teaching and research purposes. Users select a day, month, location (city or latitude and longitude) and time-of-day, and then submit their entry. The Calculator then generates the following information: latitude and longitude for the city/location, declination of the sun, height of sun at noon that day, daylength, and time of sunrise and sunset. This is a useful tool for ecological research and teaching.

Lammi, Jarmo J.

364

Stellar core collapse and supernova

Massive stars that end their stable evolution as their iron cores collapse to a neutron star or black hole long been considered good candidates for producing Type II supernovae. For many years the outward propagation of the shock wave produced by the bounce of these iron cores has been studied as a possible mechanism for the explosion. For the most part, the results of these studies have not been particularly encouraging, except, perhaps, in the case of very low mass iron cores or very soft nuclear equations of state. The shock stalls, overwhelmed by photodisintegration and neutrino losses, and the star does not explode. More recently, slow late time heating of the envelope of the incipient neutron star has been found to be capable of rejuvenating the stalled shock and producing an explosion after all. The present paper discusses this late time heating and presents results from numerical calculations of the evolution, core collapse, and subsequent explosion of a number of recent stellar models. For the first time they all, except perhaps the most massive, explode with reasonable choices of input physics. 39 refs., 17 figs., 1 tab.

Wilson, J.R.; Mayle, R.; Woosley, S.E.; Weaver, T.

1985-04-01

365

NSDL National Science Digital Library

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.

Kolb, Sandra

366

NASA Technical Reports Server (NTRS)

A banded transformer core formed by positioning a pair of mated, similar core halves on a supporting pedestal. The core halves are encircled with a strap, selectively applying tension whereby a compressive force is applied to the core edge for reducing the innate air gap. A dc magnetic field is employed in supporting the core halves during initial phases of the banding operation, while an ac magnetic field subsequently is employed for detecting dimension changes occurring in the air gaps as tension is applied to the strap.

Mclyman, C. W. T. (inventor)

1974-01-01

367

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.

John H. Cohen; Thomas E. Williams; Ali G. Kadaster; Bill V. Liddell

2002-11-01

368

Reactor whole core transport calculations without fuel assembly homogenization

The variational nodal method is generalized by dividing each spatial node into a number of triangular finite elements designated as subelements. The finite subelement trail functions allow for explicit geometry representations within each node, thus eliminating the need for nodal homogenization. The method is implemented within the Argonne National Laboratory code VARIANT and applied to two-dimensional multigroup problems. Eigenvalue and pin-power results are presented for a four-assembly OECD/NEA benchmark problem containing enriched U{sub 2} and MOX fuel pins. Our seven-group model combines spherical or simplified spherical harmonic approximations in angle with isoparametric linear or quadratic subelement basis functions, thus eliminating the need for fuel-coolant homogenization. Comparisons with reference seven-group Monte Carlo solutions indicate that in the absence of pin-cell homogenization, high-order angular approximations are required to obtain accurate eigenvalues, while the results are substantially less sensitive to the refinement of the finite subelement grids.

Nicholas Tsoulfanidis; Elmer Lewis; M.A. Smith; G. Palmiotti; T.A. Taiwo

2002-10-18

369

Evolution of the core physics concept for the Canadian supercritical water reactor

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.

Pencer, J.; Colton, A.; Wang, X.; Gaudet, M.; Hamilton, H.; Yetisir, M. [Atomic Energy of Canada, Ltd., Chalk River Laboratories, Chalk River, ON (Canada)

2013-07-01

370

Enhancement of the inherent self-protection of the fast sodium reactor cores with oxide fuel

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.

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

371

The Compactness of Presupernova Stellar Cores

NASA Astrophysics Data System (ADS)

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 ?. 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 ? and 25-30 M ?. 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.

Sukhbold, Tuguldur; Woosley, S. E.

2014-03-01

372

NASA Technical Reports Server (NTRS)

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.

Akins, James; Cobb, Billy; Hart, Steve; Leaptrotte, Jeff; Milhollin, James; Pernik, Mark

1989-01-01

373

Anisotropic Earth's Inner Core within a Dynamic Core Scenario

NASA Astrophysics Data System (ADS)

Recent global expansion of seismic data motivated a number of seismological studies of the Earth's inner core (EIC). An increasingly complex structure and anisotropy in EIC have been proposed to explain seismic data. In the meantime, new hypotheses of dynamic mechanisms have been put forward to interpret seismological results. In this study, the nature of anisotropy in EIC has been re-investigated by using PKP(BC-DF) core-sensitive differential travel-times and Fe-bcc/-hcp elastic constants calculated from first-principles. A Modified Transversely Isotropic Model (MTIM) has been introduced to account for a dynamic picture of EIC (e.g., eastward drift of material and heat flux variations at the CMB), where different chemical compositions could be stabilized at the polar/equatorial regions. Hemispherical patterns and anisotropic behaviour of EIC have been ascribed to the presence of denser polar regions (Si-poor Fe alloys) and lighter equatorial zones (Si-rich Fe alloys). A conglomerate-like EIC structure containing different material domains is then needed to address the complex anisotropy behaviour of the solid part of the Earth's core. Results have been discussed using both the seismic data from South Sandwich Islands (SSI) recorded in Alaska and the more recently collected travel-time residuals from the northern hemisphere to Antarctica.

Mattesini, M.; Belonoshko, A. B.; Tkalcic, H.; Buforn, E.; Ahuja, R.

2011-12-01

374

The major chemical trends in the binding energies of intrinsic and extrinsic core excitons are predicted for zinc-blende semiconductors using an empirical tight-binding theory and localized empirical core-hole potentials. A transition from a shallow Wannier exciton to a deep Frenkel exciton is predicted for an exciton at a core-exciton absorption edge, depending on the chemical structure of the excited atom

Harold P. Hjalmarson; Helmut Büttner; John D. Dow

1981-01-01

375

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)

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

376

Radioprotection calculations for the TRADE experiment

The TRADE project is based on the coupling of, in a sub-critical configuration, of a 115 MeV, 2 mA proton cyclotron with a TRIGA research reactor at the ENEA Casaccia centre (Rome). Detailed radioprotection calculations using the FLUKA and EA-MC Monte Carlo codes were performed during the feasibility study. The study concentrated on dose rates due to beam losses in normal operating conditions and in the calculation of activation in the most sensitive components of the experiment. Results show that a shielding of 1.4 m of barytes concrete around the beam line will be sufficient to maintain the effective doses below the level of 10 Mu Sv/h, provided that the beam losses are at the level of 10 nA/m. The activation level around the beam line and in the water will be negligible, while the spallation target will reach an activation level comparable to the one of a fuel element at maximum burnup.

Zanini, L; Herrera-Martínez, A; Kadi, Y; Rubbia, Carlo; Burgio, N; Carta, M; Santagata, A; Cinotti, L

2002-01-01

377

PRIZMA predictions of in-core detection indications in the VVER-1000 reactor

NASA Astrophysics Data System (ADS)

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.

Kandiev, Yadgar Z.; Kashayeva, Elena A.; Malyshin, Gennady N.; Modestov, Dmitry G.; Khatuntsev, Kirill E.

2014-06-01

378

Irradiated boiling water reactor (BWR) fuel behavior under reactivity-initiated accident (RIA) conditions was investigated in the Nuclear Safety Research Reactor (NSRR) of the Japan Atomic Energy Research Institute. Short test fuel rods, refabricated from a commercial 7 x 7 type BWR fuel rod at a burnup of 26 GWd/ tonne U, were pulse irradiated in the NSRR under simulated cooled startup RIA conditions of the BWRs. Thermal energy from 230 J/g fuel (55 cal/g fuel) to 410 J/g fuel (98 cal/g fuel) was promptly subjected to the test fuel rods by pulse irradiation within [approximately] 10 ms. The peak fuel enthalpies are believed to be the same as the prompt energy depositions. The test fuel rods demonstrated characteristic behavior of the irradiated fuel rods under the accident conditions, such as enhanced pellet cladding mechanical interaction (PCMI) and fission gas release. However, all the fuel rods survived the accident conditions with considerable margins. Simulations by the FRAP-T6 code and fresh fuel rod tests under the same RIA conditions highlighted the burnup effects on the accident fuel performance. The tests and the simulation suggested that the BWR fuel would possibly fail by a cladding burst due to fission gas release during the cladding temperature escalation rather than the PCMI under the cold startup RIA conditions of a severe power burst.

Nakamura, Takehiko; Yoshinaga, Makio (Japan Atomic Energy Research Inst., Ibaraki (Japan). Dept. of Reactor Safety Research); Sobajima, Makoto (Ministry of International Trade and Industry, Tokyo (Japan)); Ishijima, Kiyomi; Fujishiro, Toshio (Japan Atomic Energy Research Inst., Ibaraki (Japan). Dept. of Reactor Safety Research)

1994-10-01

379

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.

Kramer, K J; Latkowski, J F; Abbott, R P; Boyd, J K; Powers, J J; Seifried, J E

2008-10-24

380

KSI's Cross Insulated Core Transformer Technology

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.

Uhmeyer, Uwe [Kaiser Systems, Inc, 126 Sohier Road, Beverly, MA 01915 (United States)

2009-08-04

381

This experiment safety assurance package (ESAP) is a revision of the last MOX ESAP issued in February 2001(Khericha 2001). The purpose of this revision is to identify the changes in the loading pattern and to provide a basis to continue irradiation up to ~42 GWd/MT burnup (+ 2.5% as predicted by MCNP (Monte Carlo N-Particle) transport code before the preliminary postirradiation examination (PIE) results for 40 GWd/MT burnup are available. Note that the safety analysis performed for the last ESAP is still applicable and no additional analysis is required (Khericha 2001). In July 2001, it was decided to reconfigure the test assembly using the loading pattern for Phase IV, Part 3, at the end of Phase IV, Part 1, as the loading pattern for Phase IV, Parts 2 and 3. Three capsule assemblies will be irradiated until the highest burnup capsule assembly accumulates: ~50 GWd/MT burnup, based on the MCNP code predictions. The last ESAP suggests that at the end of Phase IV, Part 1, we remove the two highest burnup capsule assemblies (@ ~40 GWd/MT burnup) and send them to ORNL for PIE. Then, irradiate the test assembly using the loading pattern for Phase IV, Part 2, until the highest burnup capsule reaches ~40 GWd/MT burnup per MCNP-predicted values.

Khericha, Soli T

2002-06-01

382

This experiment safety assurance package (ESAP) is a revision of the last MOX ESAP issued in February 2001(Khericha 2001). The purpose of this revision is to identify the changes in the loading pattern and to provide a basis to continue irradiation up to {approx}42 GWd/MT burnup (+ 2.5%) as predicted by MCNP (Monte Carlo N-Particle) transport code before the preliminary postirradiation examination (PIE) results for 40 GWd/MT burnup are available. Note that the safety analysis performed for the last ESAP is still applicable and no additional analysis is required (Khericha 2001). In July 2001, it was decided to reconfigure the test assembly using the loading pattern for Phase IV, Part 3, at the end of Phase IV, Part 1, as the loading pattern for Phase IV, Parts 2 and 3. Three capsule assemblies will be irradiated until the highest burnup capsule assembly accumulates: {approx}50 GWd/MT burnup, based on the MCNP code predictions. The last ESAP suggests that at the end of Phase IV, Part 1, we remove the two highest burnup capsule assemblies ({at} {approx}40 GWd/MT burnup) and send them to ORNL for PIE. Then, irradiate the test assembly using the loading pattern for Phase IV, Part 2, until the highest burnup capsule reaches {approx}40 GWd/MT burnup per MCNP-predicted values.

Khericha, S.T.

2002-06-30

383

First Principles Study of Core-Shell Semiconductor Nanocrystals

NASA Astrophysics Data System (ADS)

Core-shell nanocrystals composed of two different semiconductors have recently attracted considerable attention. These structures provide enhanced functionality and possess more degrees of freedom than single-component semiconductor nanocrystals and quantum dots. I present the results of ab initio density functional calculations for the structures, electronic densities of states, and optical absorption gaps of core-shell nanocrystals composed of group II-VI semiconductors, such as CdSe, CdTe, ZnSe, and ZnTe. The outer surfaces of the nanocrystals are passivated using partially charged hydrogen atoms. The calculations are performed for "traditional" core-shell nanocrystals, in which a core a narrow gap semiconductor is covered with a shell of a wide gap material, and "inverted" core-shell nanocrystals, in which a wide-gap core is enclosed in a narrow-gap shell.

Vasiliev, Igor

2011-03-01

384

How Do Calculators Calculate? Helmut Knaust

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

Knaust, Helmut

385

The existence of the martian core, which has been accepted for many decades, is interesting for several reasons. First, its size and composition tell us about Mars as a whole --- its constituents and provenance. Second, its antiquity tells us about early conditions on Mars; we believe that the core formed early, and this requires that Mars had a hot

David J. Stevenson

2001-01-01

386

The detection of strongly magnetized ancient crust on Mars is one of the most surprising outcomes of recent Mars exploration, and provides important insight about the history and nature of the martian core. The iron-rich core probably formed during the hot accretion of Mars ~4.5 billion years ago and subsequently cooled at a rate dictated by the overlying mantle. A

David J. Stevenson

2001-01-01

387

The detection of strongly magnetized ancient crust on Mars is one of the most surprising outcomes of recent Mars exploration, and provides important insight about the history and nature of the martian core. The iron-rich core probably formed during the hot accretion of Mars ?4.5 billion years ago and subsequently cooled at a rate dictated by the overlying mantle. A

David J. Stevenson

2001-01-01

388

The core of ann-person game is the set of feasible outcomes that cannot be improved upon by any coalition of players. A convex game is defined as one that is based on a convex set function. In this paper it is shown that the core of a convex game is not empty and that it has an especially regular structure.

Lloyd S. Shapley

1971-01-01

389

ERIC Educational Resources Information Center

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)

Kopaska-Merkel, David C.

1995-01-01

390

Reading Antarctica's Rock Cores

NSDL National Science Digital Library

In this activity, students learn about the tools and methods paleoclimatologists use to reconstruct past climates. In constructing sediment cores themselves, students will achieve a very good understanding of the sedimentological interpretation of past climates that scientists can draw from cores.

Dahlman, Luann; Andrill

391

NSDL National Science Digital Library

The NSDL Math Common Core collection provides quick and easy access to high-quality math resources that have been related to one or more standard statements within the Math Common Core. These resources are selected from the larger NSDL collection and other trusted providers, and organized by grade level and domain area.

2010-08-10

392

ERIC Educational Resources Information Center

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,…

Michigan State Univ., East Lansing. Inst. for International Studies in Education.

393

ERIC Educational Resources Information Center

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…

Krim, Jessica; Brody, Michael

2008-01-01

394

The detection of strongly magnetized ancient crust on Mars is one of the most surprising outcomes of recent Mars exploration, and provides important insight about the history and nature of the martian core. The iron-rich core probably formed during the hot accretion of Mars approximately 4.5 billion years ago and subsequently cooled at a rate dictated by the overlying mantle. A core dynamo operated much like Earth's current dynamo, but was probably limited in duration to several hundred million years. The early demise of the dynamo could have arisen through a change in the cooling rate of the mantle, or even a switch in convective style that led to mantle heating. Presently, Mars probably has a liquid, conductive outer core and might have a solid inner core like Earth. PMID:11449282

Stevenson, D J

2001-07-12

395

NASA Technical Reports Server (NTRS)

Variations in rotation and orientation of the Moon are sensitive to solid-body tidal dissipation, dissipation due to relative motion at the fluid-core/solid-mantle boundary, and tidal Love number k2 [1,2]. There is weaker sensitivity to flattening of the core-mantle boundary (CMB) [2,3,4] and fluid core moment of inertia [1]. Accurate Lunar Laser Ranging (LLR) measurements of the distance from observatories on the Earth to four retroreflector arrays on the Moon are sensitive to lunar rotation and orientation variations and tidal displacements. Past solutions using the LLR data have given results for dissipation due to solid-body tides and fluid core [1] plus Love number [1-5]. Detection of CMB flattening, which in the past has been marginal but improving [3,4,5], now seems significant. Direct detection of the core moment has not yet been achieved.

Williams, J. G.; Boggs, D. H.; Ratcliff, J. T.

2004-01-01

396

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

397

A Technique to Determine Billet Core Charge Weight for P/M Fuel Tubes

The core length in an extruded tube depends on the weight of powder in the billet core. In the past, the amount of aluminum powder needed to give a specified core length was determined empirically. This report gives a technique for calculating the weight of aluminum powder for the P/M core. An equation has been derived which can be used to determine the amount of aluminum needed for P/M billet core charge weights. Good agreement was obtained when compared to Mark 22 tube extrusion data. From the calculated charge weight, the elastomeric bag can be designed and made to compact the U3O8-Al core.

Peacock, H.B.

2001-07-02

398

Core materials development for the fuel cycle R&D program

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.

Toloczko, M [Pacific Northwest National Laboratory (PNNL); Maloy, S [Los Alamos National Laboratory (LANL); Cole, James I. [Idaho National Laboratory (INL); Byun, Thak Sang [ORNL

2011-01-01

399

Core materials development for the fuel cycle R&D program

NASA Astrophysics Data System (ADS)

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.

Maloy, S. A.; Toloczko, M.; Cole, J.; Byun, T. S.

2011-08-01

400

Core Materials Development for the Fuel Cycle R&D Program

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.

S. A. Maloy; M. Toloczko; J. Cole; T. S. Byun

2011-08-01

401

A NEW METHOD TO QUANTIFY CORE TEMPERATURE INSTABILITY IN RODENTS.

Methods to quantify instability of autonomic systems such as temperature regulation should be important in toxicant and drug safety studies. Stability of core temperature (Tc) in laboratory rodents is susceptible to a variety of stimuli. Calculating the temperature differential o...

402

Core Disruptive Accident Analysis using ASTERIA-FBR

NASA Astrophysics Data System (ADS)

JNES is developing a core disruptive accident analysis code, ASTERIA-FBR, which tightly couples the thermal-hydraulics and the neutronics to simulate the core behavior during core disruptive accidents of fast breeder reactors (FBRs). ASTERIA-FBR consists of the three-dimensional thermal-hydraulics calculation module: CONCORD, the fuel pin behavior calculation module: FEMAXI-FBR, and the space-time neutronics module: Dynamic-GMVP or PARTISN/RKIN. This paper describes a comparison between characteristics of GMVP and PARTISN and summarizes the challenging issues on applying Dynamic-GMVP to the calculation against unprotected loss-of-flow (ULOF) event which is a typical initiator of core disruptive accident of FBR. The statistical error included in the calculation results may affect the super-prompt criticality during ULOF event and thus the amount of released energy.

Ishizu, Tomoko; Endo, Hiroshi; Yamamoto, Toshihisa; Tatewaki, Isao

2014-06-01

403

Chemical Models of Star-Forming Cores

NASA Astrophysics Data System (ADS)

We review chemical models of low-mass star forming cores including our own work. Chemistry in molecular clouds are not in equilibrium. Molecular abundances in star forming cores change not only with physical conditions in cores but also with time. In prestellar cores, temperature stays almost constant ˜ 10 K, while the gas density increases as the core collapses. Three chemical phenomena are observed in this cold phase: molecular depletion, chemical fractionation, and deuterium enrichment. They are reproduced by chemical models combined with isothermal gravitational collapse. The collapse timescale of prestellar cores depends on the initial ratios of thermal, turbulent and magnetic pressure to gravitational energy. Since the chemical timescales, such as adsorption timescale of gas particle onto grains, are comparable to the collapse timescale, molecular abundances in cores should vary depending on the collapse timescale. Observations found that molecular abundances in some cores deviate from those in other cores, in spite of their similar central densities; it could originate in the pressure to gravity ratio in the cores. As the core contraction proceeds, compressional heating eventually overwhelms radiative cooling, and the core starts to warm up. Temperature of the infalling gas rises, as it approaches the central region. Grain-surface reactions of adsorbed molecules occur in this warm-up phase, as well as in prestellar phase. Hydrogenation is efficient at T ? 20 K, whereas radicals can migrate on grain surface and react with each other to form complex organic molecules (COMs) at T ? 30 K. Grain-surface species are sublimated to the gas phase and re-start gas-phase reactions; e.g. unsaturated carbon chains are formed from sublimated methane. Our model calculation predicts that COMs increases as the warm region extends outwards and the abundances of unsaturated carbon chains depend on the gas density in the CH4 sublimation zone. Recent detection of COMs in prestellar cores may indicate that a fraction of COMs formed in the vicinity of a protostar could be distributed to ambient clouds by outflows. COMs and carbon chains in protostellar phase inherit the high D/H ratio of their mother molecules, which originate mostly in cold prestellar phase.

Aikawa, Y.

2013-10-01

404

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.

G. S. Chang; M. A. Lillo; R. G. Ambrosek

2008-06-01

405

Comparisons of RELAP4\\/MOD6 with core blowdown data

The independent verification of the RELAP4\\/MOD6, Update 3 computer code for a pressurized water reactor core during the blowdown phase of a loss-of-coolant accident is described. Calculations of RELAP4 core component models were compared with data from Semiscale and Thermal-Hydraulic Test Facility blowdown experiments. Measured boundary conditions were applied to the component models. The sensitivity of the calculated results to

1978-01-01

406

NASA Astrophysics Data System (ADS)

We have performed smoothed particle radiation magnetohydrodynamic simulations of the collapse of rotating, magnetized 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 both the first hydrostatic core and stellar core, and the breakout of the stellar outflow from the remnant of the first core. We investigate the roles of magnetic fields and thermal feedback on the outflow launching process, finding that both magnetic and thermal forces contribute to the launching of the stellar outflow. We also follow the stellar cores until they grow to masses of up to 20 Jupiter-masses, and determine their properties. We find that at this early stage, before fusion begins, the stellar cores have radii of ?3 R? with radial entropy profiles that increase outward (i.e. are convectively stable) and minimum entropies per baryon of s/kB ? 14 in their interiors. The structure of the stellar cores is found to be insensitive to variations in the initial magnetic field strength. With reasonably strong initial magnetic fields, accretion on to the stellar cores occurs through inspiralling magnetized pseudo-discs with negligible radiative losses, as opposed to first cores which effectively radiate away the energy liberated in the accretion shocks at their surfaces. We find that magnetic field strengths of >10 kG can be implanted in stellar cores at birth.

Bate, Matthew R.; Tricco, Terrence S.; Price, Daniel J.

2014-01-01

407

NASA Technical Reports Server (NTRS)

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.

Miller, R.H.; Morrison, David (Technical Monitor)

1994-01-01

408

A three-group model (PLUTON) is described, which predicts the power density distribution, plutonium buildup, and burnup profiles across the fuel pellet radius as a function of in-pile time and parameters characterizing the type of reactor system with respect to fuel temperature and changes of density during the irradiation period. The PLUTON model is a part of two fuel performance codes

Sergei Lemehov; Jinichi Nakamura; Motoe Suzuki

2001-01-01

409

This research examines the practice of equating the reactivity of spent fuel to that of fresh fuel for the purpose of performing burnup credit criticality safety analyses for PWR spent fuel pool (SFP) storage conditions. The investigation consists of comparing k{sub inf} estimates based on reactivity equivalent fresh fuel enrichment (REFFE) to k{sub inf} estimates using the actual spent fuel

J. C. Wagner; C. V. Parks

2000-01-01

410

Assessment of deep burnup concept based on graphite moderated gas-cooled thermal reactor

A systematic assessment of the General Atomics (GA) proposed one-pass and two-pass deep-burn concepts based on the modular helium-cooled reactor design (DB-MHR) using non-uranium fuel has been performed. Sensitivity studies are done to investigate the impact of core design parameters and concept on the transmutation performance (maximum of 60% destruction). The repository loading benefits arising from the DB-MHR and LWR Inert Matrix Fuel (IMF) concepts are also estimated and compared ({approx}2.0 and 1.6, respectively). (authors)

Kim, T. K.; Taiwo, T. A.; Yang, W. S.; Hill, R. N. [Argonne National Laboratory, 9700 South Cass Avenue, Argonne, IL 60439 (United States); Venneri, F. [General Atomics, P. O. Box 85608, San Diego, CA (United States)

2006-07-01

411

Core radii and common-envelope evolution

NASA Astrophysics Data System (ADS)

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.

Hall, Philip D.; Tout, Christopher A.

2014-11-01

412

NASA Astrophysics Data System (ADS)

A commercial PWR fuel sample of local burn-up of about 200 GWd/t was annealed in a Knudsen Cell Mass Spectrometer system (KCMS) with a temperature rate of 10 K/min up to 2750 K at which temperature the sample was completely vaporised. At high temperature the vapour above the sample contains mainly actinide and lanthanide oxides, the vapour pressure of refractory metals remaining below the detection limit. The local isotopic composition was determined by mass spectrometry and indicated a high content in higher actinides. From comparison of the partial vapour pressure with data for the pure oxides, obtained from experiments or literature, it is demonstrated that the actinide and lanthanide oxides dissolved in UO 2 matrix obey Henry's law.

Hiernaut, J.-P.; Gotcu, P.; Colle, J.-Y.; Konings, R. J. M.

2008-09-01

413

NSDL National Science Digital Library

This is a PDF interview, PowerPoint slide set, and webpage biography of a core manager, detailing the importance of a lab manager to oversee the complex workings of DNA sequencing machines for an entire company.

2012-05-02

414

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.

415

Core assembly storage structure

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.

Jones, Jr., Charles E. (Northridge, CA); Brunings, Jay E. (Chatsworth, CA)

1988-01-01

416

Iron diffusion from first principles calculations

NASA Astrophysics Data System (ADS)

The cores of Earth and other terrestrial planets are made up largely of iron1 and it is therefore very important to understand iron's physical properties. Chemical diffusion is one such property and is central to many processes, such as crystal growth, and viscosity. Debate still surrounds the explanation for the seismologically observed anisotropy of the inner core2, and hypotheses include convection3, anisotropic growth4 and dendritic growth5, all of which depend on diffusion. In addition to this, the main deformation mechanism at the inner-outer core boundary is believed to be diffusion creep6. It is clear, therefore, that to gain a comprehensive understanding of the core, a thorough understanding of diffusion is necessary. The extremely high pressures and temperatures of the Earth's core make experiments at these conditions a challenge. Low-temperature and low-pressure experimental data must be extrapolated across a very wide gap to reach the relevant conditions, resulting in very poorly constrained values for diffusivity and viscosity. In addition to these dangers of extrapolation, preliminary results show that magnetisation plays a major role in the activation energies for diffusion at low pressures therefore creating a break down in homologous scaling to high pressures. First principles calculations provide a means of investigating diffusivity at core conditions, have already been shown to be in very good agreement with experiments7, and will certainly provide a better estimate for diffusivity than extrapolation. Here, we present first principles simulations of self-diffusion in solid iron for the FCC, BCC and HCP structures at core conditions in addition to low-temperature and low-pressure calculations relevant to experimental data. 1. Birch, F. Density and composition of mantle and core. Journal of Geophysical Research 69, 4377-4388 (1964). 2. Irving, J. C. E. & Deuss, A. Hemispherical structure in inner core velocity anisotropy. Journal of Geophysical Research 116, B04307 (2011). 3. Buffett, B. A. Onset and orientation of convection in the inner core. Geophysical Journal International 179, 711-719 (2009). 4. Bergman, M. Measurements of electric anisotropy due to solidification texturing and the implications for the Earth's inner core. Nature 389, 60-63 (1997). 5. Deguen, R. & Cardin, P. Thermochemical convection in Earth's inner core. Geophysical Journal International 187, 1101-1118 (2011). 6. Reaman, D. M., Daehn, G. S. & Panero, W. R. Predictive mechanism for anisotropy development in the Earth's inner core. Earth and Planetary Science Letters 312, 437-442 (2011). 7. Ammann, M. W., Brodholt, J. P., Wookey, J. & Dobson, D. P. First-principles constraints on diffusion in lower-mantle minerals and a weak D'' layer. Nature 465, 462-5 (2010).

Wann, E.; Ammann, M. W.; Vocadlo, L.; Wood, I. G.; Lord, O. T.; Brodholt, J. P.; Dobson, D. P.

2013-12-01

417

Parallel plasma fluid turbulence calculations

The study of plasma turbulence and transport is a complex problem of critical importance for fusion-relevant plasmas. To this day, the fluid treatment of plasma dynamics is the best approach to realistic physics at the high resolution required for certain experimentally relevant calculations. Core and edge turbulence in a magnetic fusion device have been modeled using state-of-the-art, nonlinear, three-dimensional, initial-value fluid and gyrofluid codes. Parallel implementation of these models on diverse platforms--vector parallel (National Energy Research Supercomputer Center`s CRAY Y-MP C90), massively parallel (Intel Paragon XP/S 35), and serial parallel (clusters of high-performance workstations using the Parallel Virtual Machine protocol)--offers a variety of paths to high resolution and significant improvements in real-time efficiency, each with its own advantages. The largest and most efficient calculations have been performed at the 200 Mword memory limit on the C90 in dedicated mode, where an overlap of 12 to 13 out of a maximum of 16 processors has been achieved with a gyrofluid model of core fluctuations. The richness of the physics captured by these calculations is commensurate with the increased resolution and efficiency and is limited only by the ingenuity brought to the analysis of the massive amounts of data generated.

Leboeuf, J.N.; Carreras, B.A.; Charlton, L.A.; Drake, J.B.; Lynch, V.E.; Newman, D.E.; Sidikman, K.L.; Spong, D.A.

1994-12-31

418

NASA Technical Reports Server (NTRS)

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.

Collins, David; Brooks, Marshall; Chen, Paul; Dwelle, Paul; Fischer, Ben

1989-01-01

419

The Evolution of ONeMg Cores with MESA

NASA Astrophysics Data System (ADS)

We present calculations of the evolution of degenerate cores composed primarily of oxygen, neon, and magnesium which are undergoing compression. We make use of the state-of-the-art MESA stellar evolution code, with updated weak reaction rates from Martinez-Pinedo et al. (2014). We perform a detailed parameter study of the effects a number of quantities, including the accretion rate, magnesium mass fraction, and initial core temperature. We discuss the final fate of these ONeMg cores, focusing on cores formed as a result of the merger of two carbon-oxygen white dwarfs.

Schwab, Josiah; Quataert, Eliot; Bildsten, Lars

2015-01-01

420

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.

2008-09-09