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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

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

3

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

4

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

5

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

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

High-burnup core design using minor actinide-containing metal fuel

A neutronic design study of metal fuel fast reactor (FR) cores is conducted on the basis of an innovative fuel design concept to achieve an extremely high burnup and realize an efficient fuel cycle system. Since it is expected that the burnup reactivity swing will become extremely large in an unprecedented high burnup core, minor actinides (MAs) from light water reactors (LWRs) are added to fresh fuel to improve the core internal conversion. Core neutronic analysis revealed that high burnups of about 200 MWd/kg for a small-scale core and about 300 MWd/kg for a large-scale core can be attained while suppressing the burnup reactivity swing to almost the same level as that of conventional cores with normal burnup. An actinide burnup analysis has shown that the MA consumption ratio is improved to about 60% and that the accumulated MAs originating from LWRs can be efficiently consumed by the high-burnup metal fuel FR. (authors)

Ohta, Hirokazu; Ogata, Takanari [Central Research Institute of Electric Power Industry, 2-11-1, Iwado Kita. Komae-shi, Tokyo 201-8511 (Japan); Obara, T. [Tokyo Institute of Technology, 2-12-1, Ookayama, Meguro-ku, Tokyo 152-8550 (Japan)

2013-07-01

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

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

12

Effect of error propagation of nuclide number densities on Monte Carlo burn-up calculations

As a result of improvements in computer technology, the continuous energy Monte Carlo burn-up calculation has received attention as a good candidate for an assembly calculation method. However, the results of Monte Carlo calculations contain the statistical errors. The results of Monte Carlo burn-up calculations, in particular, include propagated statistical errors through the variance of the nuclide number densities. Therefore,

Masayuki Tohjoh; Tomohiro Endo; Masato Watanabe; Akio Yamamoto

2006-01-01

13

Reactor critical benchmark calculations for burnup credit applications

In the criticality safety analyses for the development and certification of spent fuel casks, the current approach requires the assumption of fresh fuel'' isotopics. It has been shown that the removal of the fresh fuel'' assumption and the use of spent fuel isotopics ( burnup credit'') greatly increases the payload of spent fuel casks by reducing the reactivity of the fuel. Regulatory approval of burnup credit and the requirements of ANSI/ANS 8.1 specify that calculational methods for away-from-reactor criticality safety analyses be validated against experimental measurements. Criticality analyses for low-enriched lattices of fuel pins using the fresh fuel isotopics'' assumption have been widely benchmarked against applicable critical experiments. However, the same computational methods have not been benchmarked against criticals containing spent fuel because of the non-existence of spent fuel critical experiments. Commercial reactors offer an excellent and inexhaustible source of critical configurations against which criticality analyses can be benchmarked for spent fuel configurations. This document provides brief descriptions of the benchmarks and the computational methods for the criticality analyses. 8 refs., 1 fig., 1 tab.

Renier, J.P.; Parks, C.V.

1990-01-01

14

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

15

Burn-up Calculation of Fusion-Fission Hybrid Reactor Using Thorium Cycle

A burn-up calculation system has been developed to estimate performance of blanket in a fusion-fission hybrid reactor which is a fusion reactor with a blanket region containing nuclear fuel. In this system, neutron flux is calculated by MCNP4B and then burn-up calculation is performed by ORIGEN2. The cross-section library for ORIGEN2 is made from the calculated neutron flux and evaluated

S. Shido; Y. Yamamoto; M. Matsunaka; K. Kondo; I. Murata

16

Approach to equilibrium fuelling scheme of 500 MWe PFBR based on 3-D core burnup modeling

Approach to equilibrium fuelling scheme of 500MWe prototype fast breeder reactor (PFBR) has been predicted using detailed 3-D core burnup modeling. Equilibrium is reached after two cycles of 180 effective full power days (efpd) each. One-third core is refueled every time in a repeatable scatter load scheme after every 3 cycles. Considering the constraints of linear heat rating (LHR) on

K. Devan; A. Riyas; P. Mohanakrishnan

2011-01-01

17

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

18

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

19

Burnup calculations for the HOMER-15 and SAFE-300 reactors

NASA Astrophysics Data System (ADS)

The Heatpipe Power System (HPS) is a near-term low-cost space fission power system. As the U-235 fuel of the HPS is burned, higher actinides and fission products will be produced. This will cause changes in system reactivity, radioactivity, and decay power. One potential concern is that gaseous fission products may exert excessive pressure on the fuel pin cladding. To evaluate these issues, simulations were run in MONTEBURNS. MONTEBURNS is an automated tool that links the Monte Carlo transport code MCNP with the radioactive decay and burnup code ORIGEN2. This paper describes the results of these simulations, as well as how those results compare with the current experimental database of irradiated materials. .

Amiri, Benjamin W.; Poston, David I.

2002-01-01

20

Burnup calculation of fusion–fission hybrid energy system with thorium cycle

A fusion–fission hybrid reactor which is a fusion reactor with a blanket region containing nuclear fuel. In our group, a calculation system for analysis of fusion–fission hybrid reactor has been developed and various transport and burnup calculations were carried out for hybrid systems with U–Pu fuel cycle and ITER model so far. It was confirmed that such system is feasible

M. Matsunaka; S. Shido; K. Kondo; H. Miyamaru; I. Murata

2007-01-01

21

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

22

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

23

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

24

CANDLE: The New Burnup Strategy

The new burnup strategy CANDLE (Constant Axial shape of Neutron flux, nuclide densities and power shape During Life of Energy production) is proposed. With this burnup strategy, distributions of fuel nuclide densities, neutron flux, and power density move with the same constant speed and without any change in their shapes. The excess reactivity is constant during the burnup. Therefore, any control mechanisms for the burnup are not required. Calculation procedures are presented to find these shapes and the speed of the burning region with the neutron multiplication factor of a reactor employing this burnup strategy.To demonstrate the CANDLE burnup strategy, it is applied to a fast reactor with excellent neutron economy. Only the initially built reactor requires some fissile material such as plutonium or enriched uranium for the nuclear ignition region of its core, but only natural uranium or depleted uranium is required for the other region. Succeeding reactors require only natural or depleted uranium since the burning region of the previous reactor can be utilized for the ignition region. The life of a reactor can be made longer by elongating the core height. The drift speed of the burning region for the presented fast reactor design is {approx}4 cm/yr, which is a preferable value for designing a long-life reactor. The burnup of spent fuel is {approx}40%. It is equivalent to 40% utilization of natural uranium without reprocessing and enrichment.

Sekimoto, Hiroshi; Ryu, Kouichi; Yoshimura, Yoshikane [Tokyo Institute of Technology (Japan)

2001-11-15

25

Burnup calculations and chemical analysis of irradiated fuel samples studied in LWR-PROTEUS phase II

The isotopic compositions of 5 UO{sub 2} samples irradiated in a Swiss PWR power plant, which were investigated in the LWR-PROTEUS Phase II programme, were calculated using the CASMO-4 and BOXER assembly codes. The burnups of the samples range from 50 to 90 MWd/kg. The results for a large number of actinide and fission product nuclides were compared to those of chemical analyses performed using a combination of chromatographic separation and mass spectrometry. A good agreement of calculated and measured concentrations is found for many of the nuclides investigated with both codes. The concentrations of the Pu isotopes are mostly predicted within {+-}10%, the two codes giving quite different results, except for {sup 242}Pu. Relatively significant deviations are found for some isotopes of Cs and Sm, and large discrepancies are observed for Eu and Gd. The overall quality of the predictions by the two codes is comparable, and the deviations from the experimental data do not generally increase with burnup. (authors)

Grimm, P.; Guenther-Leopold, I. [Paul Scherrer Inst., CH-5232 Villigen PSI (Switzerland); Berger, H. D. [AREVA NP GmbH, FEEP, Bunsenstrasse 43, D-91058 Erlangen (Germany)

2006-07-01

26

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

27

Environment-based pin-power reconstruction method for homogeneous core calculations

Core calculation schemes are usually based on a classical two-step approach associated with assembly and core calculations. During the first step, infinite lattice assemblies calculations relying on a fundamental mode approach are used to generate cross-sections libraries for PWRs core calculations. This fundamental mode hypothesis may be questioned when dealing with loading patterns involving several types of assemblies (UOX, MOX), burnable poisons, control rods and burn-up gradients. This paper proposes a calculation method able to take into account the heterogeneous environment of the assemblies when using homogeneous core calculations and an appropriate pin-power reconstruction. This methodology is applied to MOX assemblies, computed within an environment of UOX assemblies. The new environment-based pin-power reconstruction is then used on various clusters of 3x3 assemblies showing burn-up gradients and UOX/MOX interfaces, and compared to reference calculations performed with APOLLO-2. The results show that UOX/MOX interfaces are much better calculated with the environment-based calculation scheme when compared to the usual pin-power reconstruction method. The power peak is always better located and calculated with the environment-based pin-power reconstruction method on every cluster configuration studied. This study shows that taking into account the environment in transport calculations can significantly improve the pin-power reconstruction so far as it is consistent with the core loading pattern. (authors)

Leroyer, H.; Brosselard, C.; Girardi, E. [EDF R and D/SINETICS, 1 av du General de Gaulle, F92141 Claman Cedex (France)

2012-07-01

28

NASA Astrophysics Data System (ADS)

In this work, the Monte Carlo burn-up code SERPENT-2 has been extended and employed to study the material isotopic evolution of the Molten Salt Fast Reactor (MSFR). This promising GEN-IV nuclear reactor concept features peculiar characteristics such as the on-line fuel reprocessing, which prevents the use of commonly available burn-up codes. Besides, the presence of circulating nuclear fuel and radioactive streams from the core to the reprocessing plant requires a precise knowledge of the fuel isotopic composition during the plant operation. The developed extension of SERPENT-2 directly takes into account the effects of on-line fuel reprocessing on burn-up calculations and features a reactivity control algorithm. It is here assessed against a dedicated version of the deterministic ERANOS-based EQL3D procedure (PSI-Switzerland) and adopted to analyze the MSFR fuel salt isotopic evolution. Particular attention is devoted to study the effects of reprocessing time constants and efficiencies on the conversion ratio and the molar concentration of elements relevant for solubility issues (e.g., trivalent actinides and lanthanides). Quantities of interest for fuel handling and safety issues are investigated, including decay heat and activities of hazardous isotopes (neutron and high energy gamma emitters) in the core and in the reprocessing stream. The radiotoxicity generation is also analyzed for the MSFR nominal conditions. The production of helium and the depletion in tungsten content due to nuclear reactions are calculated for the nickel-based alloy selected as reactor structural material of the MSFR. These preliminary evaluations can be helpful in studying the radiation damage of both the primary salt container and the axial reflectors.

Aufiero, M.; Cammi, A.; Fiorina, C.; Leppänen, J.; Luzzi, L.; Ricotti, M. E.

2013-10-01

29

Technique for Sensitivity Analysis of Space- and Energy-Dependent Burn-Up Calculations.

National Technical Information Service (NTIS)

A practical method is presented for sensitivity analysis of the very complex, space-energy dependent burn-up equations, in which the neutron and nuclide fields are coupled nonlinearly. The adjoint burn-up equations that are given are in a form which can b...

M. L. Williams, J. R. White

1979-01-01

30

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

31

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

32

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

33

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

34

NASA Astrophysics Data System (ADS)

The development of tools for nuclear data uncertainty propagation in lattice calculations are presented. The Total Monte Carlo method and the Generalized Perturbation Theory method are used with the code DRAGON to allow propagation of nuclear data uncertainties in transport calculations. Both methods begin the propagation of uncertainties at the most elementary level of the transport calculation - the Evaluated Nuclear Data File. The developed tools are applied to provide estimates for response uncertainties of a PWR cell as a function of burnup.

Sabouri, P.; Bidaud, A.; Dabiran, S.; Lecarpentier, D.; Ferragut, F.

2014-04-01

35

KARATE - a code for VVER-440 core calculation

A modular calculation system has been elaborated at the KFKI Atomic Energy Research Institute for VVER-440 cores. The purpose of KARATE is the calculation of neutron physical and thermal-hydraulic processes in the core at normal, startup, and slow transient conditions. KARATE is under validation and verification (V&V) against mathematical, experimental, and operational data.

Gado, J.; Hegedus, Cs.J.; Hegyi, Gy.; Kereszturi, A.; Makai, M.; Maraczi, Cs.; Telbisz, M.

1994-12-31

36

Monte Carlo Benchmark Calculations for 400MWth PBMR Core

Benchmark calculations for the core of the 400MWth Pebble-bed Modular Reactor (PBMR), being developed in South Africa, were carried out by using MCNP5 code as a part of establishing Monte Carlo computation system for HTGR core analysis. After the detailed MCNP modeling of the whole facility, two neutronic benchmark problems, for fresh fuel and cold conditions(Case F-1), and first core

Hong-Chul KIM; Soon Young KIM; Jong Kyung KIM; Jae Man NOH

37

Core Hardenability Calculations for Carburizing Steels

Analytical expressions are presented which allow the calculation of an ideal critical diameter (D1) and a Jominy end-quench hardenability curve for a steel from its chemical composition and prior austenite grain size. The\\u000a expressions are based on alloy hardenability factors in the literature and on the previously unpublished “hardness drop” method\\u000a of determiningD, from end-quench hardenability curves. Relationships defining Jominy

J. M. Tartaglia; G. T. Eldis

1984-01-01

38

Core Hardenability Calculations for Carburizing Steels

NASA Astrophysics Data System (ADS)

Analytical expressions are presented which allow the calculation of an ideal critical diameter (D1) and a Jominy end-quench hardenability curve for a steel from its chemical composition and prior austenite grain size. The expressions are based on alloy hardenability factors in the literature and on the previously unpublished “hardness drop” method of determining D, from end-quench hardenability curves. Relationships defining Jominy curve shape as a function of D I are developed. These differ from similar relationships previously published by recognizing that, for steels of low to medium hardenability, the microstructure contains significant amounts of non-martensitic transformation products even at the prescribed first position of hardness measurement on the end-quench hardenability bar, 1.59 mm (1/16 inch) from the quenched end. The analytical expressions presented are particularly well suited for the calculation of D I and end-quench hardenability curves for boron-free carburizing steels containing 0.15 to 0.25 pct carbon.

Tartaglia, J. M.; Eldis, G. T.

1984-06-01

39

Core Physics and Kinetics Calculations for the Fissioning Plasma Core Reactor

NASA Technical Reports Server (NTRS)

Highly efficient, compact nuclear reactors would provide high specific impulse spacecraft propulsion. This analysis and numerical simulation effort has focused on the technical feasibility issues related to the nuclear design characteristics of a novel reactor design. The Fissioning Plasma Core Reactor (FPCR) is a shockwave-driven gaseous-core nuclear reactor, which uses Magneto Hydrodynamic effects to generate electric power to be used for propulsion. The nuclear design of the system depends on two major calculations: core physics calculations and kinetics calculations. Presently, core physics calculations have concentrated on the use of the MCNP4C code. However, initial results from other codes such as COMBINE/VENTURE and SCALE4a. are also shown. Several significant modifications were made to the ISR-developed QCALC1 kinetics analysis code. These modifications include testing the state of the core materials, an improvement to the calculation of the material properties of the core, the addition of an adiabatic core temperature model and improvement of the first order reactivity correction model. The accuracy of these modifications has been verified, and the accuracy of the point-core kinetics model used by the QCALC1 code has also been validated. Previously calculated kinetics results for the FPCR were described in the ISR report, "QCALC1: A code for FPCR Kinetics Model Feasibility Analysis" dated June 1, 2002.

Butler, C.; Albright, D.

2007-01-01

40

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

41

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

42

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

43

A formalism is developed for obtaining abinitio effective core potentials from numerical Hartree–Fock wavefunctions and such potentials are presented for C, N, O, F, Cl, Fe, Br, and I. The effective core potentials enable one to eliminate the core electrons and the associated orthogonality constraints from electronic structure calculations on atoms and molecules. The effective core potentials are angular momentum

Luis R. Kahn; Paul Baybutt; Donald G. Truhlar

1976-01-01

44

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

45

Value of burnup credit beyond actinides

DOE has submitted a topical report to the NRC justifying burnup credit based only on actinide isotopes (U-234, U-235, U-236, U-238, Pu-238, Pu-239, Pu-240, Pu-241, Pu-242, and Am-241). When this topical report is approved, it will allow a great deal of the commercial spent nuclear fuel to be transported in significantly higher capacity casks. A cost savings estimate for shipping fuel in 32 assembly (burnup credit) casks as opposed to 24 assembly (non-burnup credit) casks was previously presented. Since that time, more detailed calculations have been performed using the methodology presented in the Actinide-Only Burnup Credit Topical Report. Loading curves for derated casks have been generated using actinide-only burnup credit and are presented in this paper. The estimates of cost savings due to burnup credit for shipping fuel utilizing 32, 30, 28, and 24 assembly casks where only the 24 assembly cask does not burnup credit have been created and are discussed. 4 refs., 2 figs.

Lancaster, D.; Fuentes, E.; Kang, Chi

1997-12-01

46

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

47

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

48

Potential curves and spectroscopic constants for electronic states of the Li2 molecule dissociating into 2s + 2s, 2s + 2p, 2s + 3s, 2p + 2p, 2s + 3p, and 2s + 3d atomic configurations (49 states) are obtained from CI calculations with electronic core potentials, including core polarization effects.

R. Poteau; F. Spiegelmann

1995-01-01

49

Calculation methods for core distortions and mechanical behavior

This paper describes ABADAN, a general purpose, nonlinear, multi-dimensional finite element structural analyses computer code developed for the express purpose of solving large nonlinear problems as typified by the Liquid Metal Fast Breeder Reactor (LMFBR) Core Restraint System design problem. All of the structural modeling features inherent in a general purpose finite element code and required to adequately model an LMFBR core restraint system are demonstrated. Typical results for a radial row and a sixty degree sector model of FFTF are presented. The sixty degree sector results are interpreted in terms of the design criteria that the core restraint system must satisfy. Extensions and adaptations of these modeling techniques to different core restraint design concepts can be readily achieved. 27 figures.

Sutherland, W.H.

1984-09-01

50

Continuum-discretized coupled-channels calculations with core excitation

NASA Astrophysics Data System (ADS)

The effect of core excitation in the elastic scattering and breakup of a two-body halo nucleus on a stable target nucleus is studied. The structure of the weakly bound projectile is described in the weak-coupling limit, assuming a particle-rotor model. The eigenfunctions and the associated eigenvalues are obtained by diagonalizing this Hamiltonian in a square-integrable basis (pseudostates). For the radial coordinate between the particle and the core, a transformed harmonic oscillator (THO) basis is used. For the reaction dynamics, an extension of the continuum-discretized coupled-channels (CDCC) method, which takes into account dynamic core excitation and de-excitation due to the presence of noncentral parts in the core-target interaction, is adapted to be used along with a pseudostates (PS) basis.

de Diego, R.; Arias, J. M.; Lay, J. A.; Moro, A. M.

2014-06-01

51

The CANDLE burnup strategy is a new reactor burnup concept, where the distributions of fuel nuclide densities, neutron flux, and power density move with the same constant speed along the core axis from bottom to top or from top to bottom of the core and without any change in their shapes. It can be applied easily to the block-type high temperature gas cooled reactor using an appropriate burnable poison mixed with uranium oxide fuel. In the present study, the burnup distribution and the temperature distribution in the core are investigated and their effects on the CANDLE burnup core characteristics are studied. In this study, the natural gadolinium is used as the burnable poison. With the fuel enrichment of 15%, the natural gadolinium concentration of 3.0% and the fuel pin pitch of 6.6 cm, the CANDLE burnup is realized with the burning region moving speed of 29 cm/year and the axial half width of power density distribution of 1.5 m for uniform group constant case at 900 K. When the effect of nuclide change by burnup is considered, the burning region speed becomes 25 cm/year and the axial half-width of power density distribution becomes 1.25 m. When the temperature distributions effect is considered, the effects on the core characteristics are smaller than the burnup distribution effect. The maximum fuel temperature of the parallel flow case is higher than the counter flow case. (authors)

Yasunori Ohoka; Ismile; Hiroshi Sekimoto [Tokyo Institute of Technology, Research Laboratory for Nuclear Reactors, O-okayama, Meguro-ku, Tokyo 152-8550 (Japan)

2004-07-01

52

Calculations of the accretion and evolution of giant planets The effects of solid cores

NASA Technical Reports Server (NTRS)

The present calculation of giant planet evolution proceeds under the hypothesis that the solid cores formed by small particle accretion later attracted their gaseous envelopes from the solar nebula gravitationally. Evolutionary calculations are presented for the beginning of gaseous envelope formation around the core mass; the growth of core and envelope to a critical core mass; the derivation of high luminosity from the envelope mass due to gravitational contraction, and the termination of both core and envelope accretion on a time-scale of 5 Gyr.

Bodenheimer, P.; Pollack, J. B.

1986-01-01

53

Core design calculations for MEU fuel in FSV

Results of a study on the feasibility of converting the Fort St. Vrain reactor core from the present 6-year, High Enriched Uranium (HEU) fuel cycle using uranium enriched to 93% U-235 and thorium to a 6-year, Medium Enriched Uranium (MEU) fuel cycle using uranium enriched to 20% U-235 and thorium, are described. The study shows that a transition from the

M. E. Fehrenbach; A. M. Baxter

1979-01-01

54

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

55

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

56

Whole-Core Heterogeneous Transport Calculations and Their Comparison with Diffusion Results

Recently the method of characteristics (MOC) has been considered as an effective methodology in lattice calculations. This method gives accurate solutions in complex geometries and strong absorber problems. With increasingly more heterogeneous reactor cores such as a mixed-oxide (MOX) fuel-loaded core or a burnable absorber-loaded core, the limitations due to homogenization and diffusion theory are evident, and the need for

Nam Zin Cho; Gil Soo Lee; Ser Gi Hong; Chang Keun Jo; Kyung Taek Lee

2000-01-01

57

Ab initio free energy calculations on the polymorphs of iron at core conditions

In order to predict the stable polymorph of iron under core conditions, calculations have been performed on all the candidate phases proposed for inner core conditions, namely, body-centred cubic (bcc), body-centred tetragonal (bct), hexagonal close-packed (hcp), double-hexagonal close-packed (dhcp) and an orthorhombically distorted hcp polymorph. Our simulations are ab initio free energy electronic structure calculations, based upon density functional theory,

Lidunka Vo?adlo; John Brodholt; Dario Alfè; Michael J. Gillan; Geoffrey D. Price

2000-01-01

58

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

59

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

60

RMC - A Monte Carlo Code for Reactor Core Analysis

NASA Astrophysics Data System (ADS)

A new Monte Carlo transport code RMC has been being developed by Department of Engineering Physics, Tsinghua University, Beijing as a tool for reactor core analysis on high-performance computing platforms. To meet the requirements of reactor analysis, RMC now has such functions as criticality calculation, fixed-source calculation, burnup calculation and kinetics simulations. Some techniques for geometry treatment, new burnup algorithm, source convergence acceleration, massive tally and parallel calculation, and temperature dependent cross sections processing are researched and implemented in RMC to improve the effciency. Validation results of criticality calculation, burnup calculation, source convergence acceleration, tallies performance and parallel performance shown in this paper prove the capabilities of RMC in dealing with reactor analysis problems with good performances.

Wang, Kan; Li, Zeguang; She, Ding; Liang, Jin'gang; Xu, Qi; Qiu, Yishu; Yu, Jiankai; Sun, Jialong; Fan, Xiao; Yu, Ganglin

2014-06-01

61

Local Burn-Up Effects in the NBSR Fuel Element

This study addresses the over-prediction of local power when the burn-up distribution in each half-element of the NBSR is assumed to be uniform. A single-element model was utilized to quantify the impact of axial and plate-wise burn-up on the power distribution within the NBSR fuel elements for both high-enriched uranium (HEU) and low-enriched uranium (LEU) fuel. To validate this approach, key parameters in the single-element model were compared to parameters from an equilibrium core model, including neutron energy spectrum, power distribution, and integral U-235 vector. The power distribution changes significantly when incorporating local burn-up effects and has lower power peaking relative to the uniform burn-up case. In the uniform burn-up case, the axial relative power peaking is over-predicted by as much as 59% in the HEU single-element and 46% in the LEU single-element with uniform burn-up. In the uniform burn-up case, the plate-wise power peaking is over-predicted by as much as 23% in the HEU single-element and 18% in the LEU single-element. The degree of over-prediction increases as a function of burn-up cycle, with the greatest over-prediction at the end of Cycle 8. The thermal flux peak is always in the mid-plane gap; this causes the local cumulative burn-up near the mid-plane gap to be significantly higher than the fuel element average. Uniform burn-up distribution throughout a half-element also causes a bias in fuel element reactivity worth, due primarily to the neutronic importance of the fissile inventory in the mid-plane gap region.

Brown N. R.; Hanson A.; Diamond, D.

2013-01-31

62

Dependence of control rod worth on fuel burnup

One important parameter in the design and the analysis of a nuclear reactor core is the reactivity worth of the control rods, i.e. their efficiency to absorb excess reactivity. The control rod worth is affected by parameters such as the fuel burnup in the rod vicinity, the Xe concentration in the core, the operational time of the rod and its

P. Savva; M. Varvayanni; N. Catsaros

2011-01-01

63

The ORR Whole-Core LEU Fuel Demonstration

The ORR Whole-Core LEU Fuel Demonstration, conducted as part of the US Reduced Enrichment Research and Test Reactor Program, has been successfully completed. Using commercially-fabricated U{sub 3}Si{sub 2}-Al 20%-enriched fuel elements (4.8 g U/cc) and fuel followers (3.5 g U/cc), the 30-MW Oak Ridge Research Reactor was safely converted from an all-HEU core, through a series of HEU/LEU mixed transition cores, to an all-LEU core. There were no fuel element failures and average discharge burnups were measured to be as high as 50% for the standard elements and 75% for the fuel followers. Experimental results for burnup-dependent critical configurations, cycle-averaged fuel element powers, and fuel-element-averaged {sup 235}U burnups validated predictions based on three-dimensional depletion calculations. Calculated values for plutonium production and isotopic mass ratios as functions of {sup 235}U burnup support the corresponding measured quantities. In general, calculations for reaction rate distributions, control rod worths, prompt neutron decay constants, and isothermal temperature coefficients were found to agree with corresponding measured values. Experimentally determined critical configurations for fresh HEU and LEU cores radially reflected with water and with beryllium are well-predicted by both Monte Carlo and diffusion calculations. 17 refs.

Bretscher, M.M.; Snelgrove, J.L.

1990-01-01

64

Melting of iron under Earth's core conditions from diffusion Monte Carlo free energy calculations.

The temperature of Earth's core is a parameter of critical importance to model the thermal structure of Earth. Since the core is mainly made of iron, with a solid liquid boundary (the inner core boundary) at 1220 km from the center of the Earth, the melting temperature of iron at the pressure of the ICB provides constraints on the temperature of the core. These constraints are based either on extrapolations to ICB pressure of experimental measurements, or on theoretical calculations which employed various flavors of quantum mechanics, most notably density functional theory. Significant disagreement between estimates obtained with different methods calls for calculations based on more accurate techniques. Here we used quantum Monte Carlo techniques to compute the free energies of solid and liquid iron at ICB conditions. We obtained an iron melting temperature at 330 GPa of 6900+/-400 K. PMID:19792692

Sola, Ester; Alfè, Dario

2009-08-14

65

One primary concern for design of safety systems for reactors is the time response of external detectors to changes in the core. This paper describes a way to estimate the time delay between the core power production and the external detector response using Monte Carlo calculations and suggests a technique to measure the time delay. The Monte Carlo code KENO-NR was used to determine the time delay between the core power production and the external detector response for a conceptual design of the Advanced Neutron Source (ANS) reactor. The Monte Carlo estimated time delay was determined to be about 10 ms for this conceptual design of the ANS reactor.

Valentine, T.E.; Mihalczo, J.T.

1996-08-01

66

Ab initio effective core potentials (ECP's) have been generated to replace the innermost core electron for third-row (K--Au), fourth-row (Rb--Ag), and fifth-row (Cs--Au) atoms. The outermost core orbitals: corresponding to the ns/sup 2/np/sup 6/ configuration for the three rows here: are not replaced by the ECP but are treated on an equal footing with the nd, (n+1)s and (n+1)p valence orbitals. These ECP's have been derived for use in molecular calculations where these outer core orbitals need to be treated explicitly rather than to be replaced by an ECP. The ECP's for the forth and fifth rows also incorporate the mass--velocity and Darwin relativistic effects into the potentials. Analytic fits to the potentials are presented for use in multicenter integral evaluation. Gaussian orbital valence basis sets are developed for the (3s, 3p, 3d, 4s, 4p), (4s, 4p, 4d, 5s, 5p), and (5s, 5p, 5d, 6s, 6p) ortibals of the three respective rows.

Hay, P.J.; Wadt, W.R.

1985-01-01

67

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

68

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

69

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

70

VENUS-2 MOX Core Benchmark: Results of ORNL Calculations Using HELIOS-1.4 - Revised Report

The Task Force on Reactor-Based Plutonium Disposition (TFRPD) was formed by the Organization for Economic Cooperation and Development/Nuclear Energy Agency (OECD/NEA) to study reactor physics, fuel performance, and fuel cycle issues related to the disposition of weapons-grade (WG) plutonium as mixed-oxide (MOX) reactor fuel. To advance the goals of the TFRPD, 10 countries and 12 institutions participated in a major TFRPD activity: a blind benchmark study to compare code calculations to experimental data for the VENUS-2 MOX core at SCK-CEN in Mol, Belgium. At Oak Ridge National Laboratory, the HELIOS-1.4 code system was used to perform the comprehensive study of pin-cell and MOX core calculations for the VENUS-2 MOX core benchmark study.

Ellis, RJ

2001-06-01

71

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

72

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

73

Melting of Iron under Earth's Core Conditions from Diffusion Monte Carlo Free Energy Calculations

Melting of Iron under Earth's Core Conditions from Diffusion Monte Carlo Free Energy Calculations Ester Sola1 and Dario Alfe`1,2 1 Thomas Young Centre@UCL, and Department of Earth Sciences, UCL, Gower. Here we used quantum Monte Carlo techniques to compute the free energies of solid and liquid iron

AlfÃ¨, Dario

74

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

75

Burnup credit validation of SCALE-4 using light-water-reactor criticals

The ANSI/ANS 8.1 criticality safety standard recommends validation and benchmarking of the calculational methods used in evaluating criticality safety limits for away-from-reactor applications. The lack of critical experiments with burned light-water reactor (LWR) fuel in racks or in casks necessitates the validation of burnup credit methods by comparison with LWR core criticals. These benchmarks are relevant because they test a methodology`s ability to predict spent fuel isotopic and to evaluate the reactivity effects of heterogeneities and strong absorbers. Data are available to perform analyses at precise state points. As part of the Burnup Credit Analysis Verification (BCAV) Task, the U.S. Department of Energy has sponsored analysis of selected reactor core critical configurations from commercial pressurized-water-reactors (PWRs). The initial analysis methodology used the SCALE-4 code system to analyze a set of reactor critical configurations from Virginia Power`s Slurry and North Anna reactors. However, the analysis procedure was complex and included the calculation of lumped fission products. The methodology has since been revised to simplify both the data requirements and the calculational procedure for the criticality analyst. This revised methodology is validated here by a comparison with three reactor critical configurations from Tennessee Valley Authority`s Sequoyah Unit 2 Cycle 3 and two from Virginia Power`s Slurry Unit 1 Cycle 2.

Bowman, S.M.; Hermann, O.W. [Oak Ridge National Lab., TN (United States); Brady, M.C. [Sandia National Labs., Albuquerque, NM (United States)

1993-03-01

76

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. The change of excess reactivity along burnup is theoretically zero, and the core characteristics, such as power feedback coefficients and power peaking factor, are not changed along burnup. Application of this burnup strategy to neutron rich fast reactors makes excellent performances. Only natural or depleted uranium is required for the replacing fuels. About 40 % of natural or depleted uranium undergoes fission without the conventional reprocessing and enrichment. If the LWR produced energy of X Joules, the CANDLE reactor can produce about 50 X Joules from the depleted uranium left at the enrichment facility for the LWR fuel. If we can say LWRs have produced energy sufficient for full 20 years, we can produce the energy for 1000 years by using the CANDLE reactors with depleted uranium. We need not mine any uranium ore, and do not need reprocessing facility. The burnup of spent fuel becomes 10 times. Therefore, the spent fuel amount per produced energy is also reduced to one-tenth. The equilibrium core contains a lot of instable materials such as higher actinides and fission products, the enough amounts of which can not be obtained easily. The construction of the initial core is a difficult problem. However, by using enriched uranium substituted for actinides in the equilibrium core, we can construct the initial core whose power profile is similar to the equilibrium one and will reach the equilibrium state without any big change during transient. At present we do not have any material standing for such a high burnup. However, the CANDLE burnup can be realized by employing simple reprocessing, which separates actinides and fission products and replaces the cladding by new one. (author)

Hiroshi, Sekimoto [Research Laboratory for Nuclear Reactors, Tokyo Institute of Technology, O-okayama, Meguro-ku, Tokyo (Japan)

2006-07-01

77

Improvements in cost performance of computer hardware are extending applicability of cell-heterogeneous transport calculations that are currently applied in lattice cell or assembly geometry. In this paper, a cell-heterogeneous whole-core transport calculation by the method of characteristics (MOC) is applied to calculations for analyses of pressurized water reactor (PWR) initial cores. Calculation accuracy by the MOC has been verified within

Masahiro Tatsumi; Tatsuya Kimoto; Akio Yamamoto

2000-01-01

78

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

79

VENUS-2 MOX Core Benchmark: Results of ORNL Calculations Using HELIOS-1.4

The Task Force on Reactor-Based Plutonium Disposition, now an Expert Group, was set up through the Organization for Economic Cooperation and Development/Nuclear Energy Agency to facilitate technical assessments of burning weapons-grade plutonium mixed-oxide (MOX) fuel in U.S. pressurized-water reactors and Russian VVER nuclear reactors. More than ten countries participated to advance the work of the Task Force in a major initiative, which was a blind benchmark study to compare code benchmark calculations against experimental data for the VENUS-2 MOX core at SCK-CEN in Mol, Belgium. At the Oak Ridge National Laboratory, the HELIOS-1.4 code was used to perform a comprehensive study of pin-cell and core calculations for the VENUS-2 benchmark.

Ellis, RJ

2001-02-02

80

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

The nonuniformity of the temperature distribution in an LMFBR leads to differential thermal expansion of the walls of an assembly hexcan. These thermal expansion differentials cause the hexcan to distort or bow. Consequentially, the assembly experiences a spatial displacement, which results in a change in reactivity for the core. A computational model to calculate the reactivity feedback due to material displacements induced by assembly bowing effects has been developed.

Not Available

1983-01-01

81

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

82

After the severe core damage of Fukushima Dai-Ichi Nuclear Power Station, radioactive material leaked from the reactor buildings. As part of monitoring of radioactivity in the site, measurements of radioactivity in soils at three fixed points have been performed for {sup 134}Cs and {sup 137}Cs with gamma-ray spectrometry and for Pu, Pu, and {sup 240}Pu with {alpha}-ray spectrometry. Correlations of radioactivity ratios of {sup 134}Cs to {sup 137}Cs, and {sup 238}Pu to the sum of {sup 239}Pu and {sup 240}Pu with fuel burnup were studied by using theoretical burnup calculations and measurements on isotopic inventories, and compared with the Cs and Pu radioactivity rations in the soils. The comparison indicated that the burnup of the fuel sourcing the radioactivity was from 18 to 38 GWd/t, which corresponded to that of the fuel in the highest power and, therefore, the highest decay heat in operating high-burnup fueled BWR cores. (authors)

Yamamoto, T.; Suzuki, M.; Ando, Y. [Japan Nuclear Energy Safety Organization, Toranomon Towers Office, 14F, 4-1-28, Toranomon, Minato-ku, Tokyo 105-0001 (Japan)

2012-07-01

83

Burnup credit validation of SCALE-4 using light water reactor criticals

The ANSI/ANS 8.1 criticality safety standard recommends validation and benchmarking of the calculational methods used in evaluating criticality safety limits for away-from-reactor applications. The lack of critical experiments with burned light-water-reactor (LWR) fuel in racks or in casks necessitates the validation of burnup credit methods by comparison to LWR core criticals. These are relevant benchmarks because they test a methodology's ability to predict spent fuel isotopics and to evaluate the reactivity effects of heterogeneities and strong absorbers. Data are available to perform analyses at precise state points. The US Department of Energy Burnup Credit Program has sponsored analysis of selected reactor core critical configurations from commercial pressurized-water-reactors (PWRs) in order to validate an appropriate analysis methodology. The initial methodology used the SCALE-4 code system to analyze a set of reactor critical configurations from Virginia Power's Surry and North Anna reactors. The methodology has since been revised to simplify both the data requirements and the calculational procedure for the criticality analyst. This revised methodology is validated here by comparison to three reactor critical configurations from Tennessee Valley Authority's Sequoyah Unit 2 Cycle 3 and two from Virginia Power's Surry Unit 1 Cycle 2.

Bowman, S.M.; Hermann, O.W. (Oak Ridge National Lab., TN (United States)); Brady, M.C. (Sandia National Labs., Albuquerque, NM (United States))

1993-01-01

84

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

85

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 estimate in criticality calculations. Because of the numerous economical benefits including all available negative reactivity (i.e., full burnup credit) could provide [2], it is advantageous to work toward full burnup credit. Unfortunately, comparatively little work has been done to include non-major actinides and other fission products (FP) in burnup credit analyses due in part to insufficient experimental data for validation of codes and nuclear data. The Burnup Credit Criticality Experiment (BUCCX) at Sandia National Laboratory was a set of experiments with {sup 103}Rh that have relevance for burnup credit [3]. This work uses TSUNAMI-3D to investigate and adjust a BUCCX model to match isotope-specific, energy-dependent k{sub eff} sensitivity profiles to those of a representative high-capacity cask model (GBC-32) [4] for each FP of interest. The isotopes considered are {sup 149}Sm, {sup 143}Nd, {sup 103}Rh, {sup 133}Cs, {sup 155}Gd, {sup 152}Sm, {sup 99}Tc, {sup 145}Nd, {sup 153}Eu, {sup 147}Sm, {sup 109}Ag, {sup 95}Mo, {sup 150}Sm, {sup 101}Ru, and {sup 151}Eu. The goal is to understand the biases and bias uncertainties inherent in nuclear data, and ultimately, to apply these in support of full burnup credit.

Mueller, Don [ORNL; Roberts, Jeremy A [ORNL

2008-01-01

86

Full Core 3-D Simulation of a Partial MOX LWR Core

A comparative analysis and comparison of results obtained between 2-D lattice calculations and 3-D full core nodal calculations, in the frame of MOX fuel design, was conducted. This study revealed a set of advantages and disadvantages, with respect to each method, which can be used to guide the level of accuracy desired for future fuel and fuel cycle calculations. For the purpose of isotopic generation for fuel cycle analyses, the approach of using a 2-D lattice code (i.e., fuel assembly in infinite lattice) gave reasonable predictions of uranium and plutonium isotope concentrations at the predicted 3-D core simulation batch average discharge burnup. However, it was found that the 2-D lattice calculation can under-predict the power of pins located along a shared edge between MOX and UO2 by as much as 20%. In this analysis, this error did not occur in the peak pin. However, this was a coincidence and does not rule out the possibility that the peak pin could occur in a lattice position with high calculation uncertainty in future un-optimized studies. Another important consideration in realistic fuel design is the prediction of the peak axial burnup and neutron fluence. The use of 3-D core simulation gave peak burnup conditions, at the pellet level, to be approximately 1.4 times greater than what can be predicted using back-of-the-envelope assumptions of average specific power and irradiation time.

S. Bays; W. Skerjanc; M. Pope

2009-05-01

87

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

88

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

89

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

90

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

91

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

92

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

93

Whole-core neutron transport calculations without fuel-coolant homogenization

The variational nodal method implemented in the VARIANT code is generalized to perform full core transport calculations without spatial homogenization of cross sections at either the fuel-pin cell or fuel assembly level. The node size is chosen to correspond to one fuel-pin cell in the radial plane. Each node is divided into triangular finite subelements, with the interior spatial flux distribution represented by piecewise linear trial functions. The step change in the cross sections at the fuel-coolant interface can thus be represented explicitly in global calculations while retaining the fill spherical harmonics capability of VARIANT. The resulting method is applied to a two-dimensional seven-group representation of a LWR containing MOX fuel assemblies. Comparisons are made of the accuracy of various space-angle approximations and of the corresponding CPU times.

Smith, M. A.; Tsoulfanidis, N.; Lewis, E. E.; Palmiotti, G.; Taiwo, T. A.

2000-02-10

94

A consistent set of ab initio effective core potentials (ECP) has been generated for the main group elements from Na to Bi using the procedure originally developed by Kahn. The ECP's are derived from all-electron numerical Hartree--Fock atomic wave functions and fit to analytical representations for use in molecular calculations. For Rb to Bi the ECP's are generated from the relativistic Hartree--Fock atomic wave functions of Cowan which incorporate the Darwin and mass--velocity terms. Energy-optimized valence basis sets of (3s3p) primitive Gaussians are presented for use with the ECP's. Comparisons between all-electron and valence-electron ECP calculations are presented for NaF, NaCl, Cl/sub 2/, Cl/sub 2//sup -/, Br/sub 2/, Br/sub 2//sup -/, and Xe/sub 2//sup +/. The results show that the average errors introduced by the ECP's are generally only a few percent.

Wadt, W.R.; Hay, P.J.

1985-01-01

95

NASA Technical Reports Server (NTRS)

To determine the feasibility of coupling the output of a single-mode optical fiber into a single-mode rib waveguide in a temperature varying environment, a theoretical calculation of the coupling efficiency between the two was investigated. Due to the complex geometry of the rib guide, there is no analytical solution to the wave equation for the guided modes, thus, approximation and/or numerical techniques must be utilized to determine the field patterns of the guide. In this study, three solution methods were used for both the fiber and guide fields; the effective-index method (EIM), Marcatili's approximation, and a Fourier method. These methods were utilized independently to calculate the electric field profile of each component at two temperatures, 20 C and 300 C, representing a nominal and high temperature. Using the electric field profile calculated from each method, the theoretical coupling efficiency between an elliptical-core optical fiber and a rib waveguide was calculated using the overlap integral and the results were compared. It was determined that a high coupling efficiency can be achieved when the two components are aligned. The coupling efficiency was more sensitive to alignment offsets in the y direction than the x, due to the elliptical modal field profile of both components. Changes in the coupling efficiency over temperature were found to be minimal.

Tuma, Margaret L.; Weisshaar, Andreas; Li, Jian; Beheim, Glenn

1995-01-01

96

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

97

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

98

Criticality validation for burnup credit using recycle Pu criticals

A set of 23 additional critical experiments were analyzed to provide additional input to the criticality validation portion of spent fuel cask analysis. The results of this analyses were combined with the previously analyzed criticals to determine the upper safety limit on k{sub eff}. The combined set of criticals can be used used for criticality validation for burnup credit, and are better suited for the range of isotopics in spent nuclear fuels. A trend observed in the analysis was that the calculated k{sub eff} deviates from the criticals in the positive direction, implying that increased burnup results in increased safety margin. 6 refs., 2 figs., 1 tab.

Fuentes, E.; Lancaster, D.

1997-04-01

99

Destructive methods were used for the burnup determination of a PWR nuclear fuel irradiated to a high burnup in power reactors, and of a dry processed fuel fabricated from a spent PWR fuel and irradiated in the Hanaro research reactor. The total burnup was determined from a measurement of the Nd and Cs isotope burnup monitors. The methods included U,

Jung Suk KIM; Young Shin JEON; Soon Dal PARK; Sun Ho HAN; Jong Goo KIM

2007-01-01

100

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

101

NASA Astrophysics Data System (ADS)

Composites including VO2-based thermochromic nanoparticles are able to combine high luminous transmittance Tlum with a significant modulation of the solar energy transmittance ?Tsol at a "critical" temperature in the vicinity of room temperature. Thus nanothermochromics is of much interest for energy efficient fenestration and offers advantages over thermochromic VO2-based thin films. This paper presents calculations based on effective medium theory applied to dilute suspensions of core-shell nanoparticles and demonstrates that, in particular, moderately thin-walled hollow spherical VO2 nanoshells can give significantly higher values of ?Tsol than solid nanoparticles at the expense of a somewhat lowered Tlum. This paper is a sequel to a recent publication [S.-Y. Li, G. A. Niklasson, and C. G. Granqvist, J. Appl. Phys. 108, 063525 (2010)].

Li, S.-Y.; Niklasson, G. A.; Granqvist, C. G.

2011-06-01

102

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

103

Experiment and calculation on seismic behavior of RC composite core walls with concealed steel truss

To improve the seismic performance of reinforced concrete core walls, reinforced concrete composite core walls with concealed\\u000a steel truss were proposed and systemically investigated. Two 1\\/6 scale core wall specimens, including a normal reinforced\\u000a concrete core wall and a reinforced concrete composite core wall with concealed steel truss, were designed. The experimental\\u000a study on seismic performance under cyclic loading was

Wanlin Cao; Weihua Chang; Changjun Zhao; Jianwei Zhang

2009-01-01

104

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

105

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

106

NASA Astrophysics Data System (ADS)

Use of U and U-Th fuels in CANDU type of reactors (CANDU-6 and ACR-700) on the once-through nuclear fuel cycle is investigated. Based on the unit-cell approximation with the homogeneous-bundle/core model, utilizing the MONTEBURNS code, burnup computations are performed; discharge burnups are determined and expressed as functions of 235U and Th fractions, when applicable. Natural Uranium Requirement (and Saving) and Nuclear Resource Utilization are calculated for varying fuel compositions. Results are analyzed to observe the effects of 235U and Th fractions, thus to reach conclusions about use of Th in CANDU-6 and ACR-700 on the once-through cycle.

Türkmen, Mehmet; Zabuno?lu, Okan H.

2012-10-01

107

Melting of Iron under Earth's Core Conditions from Diffusion MonteCarlo Free Energy Calculations

The temperature of Earth's core is a parameter of critical importance to model the thermal structure of Earth. Since the core is mainly made of iron, with a solid liquid boundary (the inner core boundary) at 1220 km from the center of the Earth, the melting temperature of iron at the pressure of the ICB provides constraints on the temperature

Ester Sola; Dario Alfè

2009-01-01

108

Investigation of Burnup Credit Issues in BWR Fuel

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

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

1999-09-20

109

The unitary correlation operator method (UCOM) and the similarity renormalization group theory (SRG) are compared and discussed in the framework of no-core Monte Carlo shell model (MCSM) calculations for $^{3}$H and $^{4}$He. The treatment of spurious center-of-mass motion by Lawson's prescription is performed in the MCSM calculations. These results with both transformed interactions show good suppression of spurious center-of-mass motion with proper Lawson's prescription parameter $\\beta_{\\rm c.m.}$ values. The UCOM potentials obtains faster convergence of total energy for the ground state than that of SRG potentials in the MCSM calculations, which differs from the cases in the no-core shell model calculations (NCSM). This differences are discussed and analyzed in terms of the truncation scheme in the MCSM and NCSM, as well as the properties of potentials of SRG and UCOM.

Lang Liu

2014-11-22

110

The unitary correlation operator method (UCOM) and the similarity renormalization group theory (SRG) are compared and discussed in the framework of no-core Monte Carlo shell model (MCSM) calculations for $^{3}$H and $^{4}$He. The treatment of spurious center-of-mass motion by Lawson's prescription is performed in the MCSM calculations. These results with both transformed interactions show good suppression of spurious center-of-mass motion with proper Lawson's prescription parameter $\\beta_{\\rm c.m.}$ values. The UCOM potentials obtains faster convergence of total energy for the ground state than that of SRG potentials in the MCSM calculations, which differs from the cases in the no-core shell model calculations (NCSM). This differences are discussed and analyzed in terms of the truncation scheme in the MCSM and NCSM, as well as the properties of potentials of SRG and UCOM.

Liu, Lang

2014-01-01

111

Burnup credit feasibility for BWR spent fuel shipments

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

Broadhead, B.L.

1990-01-01

112

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

113

Overview of spent fuel burnup measurements

The purpose of this paper is to provide a summary of the principal factors that influence burnup measurement accuracy, and the manner in which different combinations of these factors in any particular application influence the achievable accuracy in that application. The principal focus of the paper is on the passive measurement of various burnup indicators. This paper also provides a general background for four subsequent papers which discuss burnup measurement in two particular applications: for burnup credit in spent fuel shipping casks; and for safeguards purposes. This paper provides a basis for comparing these two applications in terms that directly relate to the measurement process.

McLeod, N.B. (Johnson (E.R.) Associates, Inc., Oakton, VA (United States))

1991-01-01

114

PHYSICAL REVIEW B 89, 035120 (2014) Electronic stopping power from first-principles calculations electronic stopping power Se of energetic ions in graphitic targets from first principles. By treating core into the dependence of the electronic stopping power Se on projectile velocity have been obtained with the explicit

Krasheninnikov, Arkady V.

115

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

116

Calculation of the elastic properties of a triangular cell core for lightweight composite mirrors

NASA Astrophysics Data System (ADS)

The use of composite materials in the fabrication of optical telescope mirrors offers many advantages over conventional methods, including lightweight, portability and the potential for lower manufacturing costs. In the construction of the substrate for these mirrors, sandwich construction offers the advantage of even lower weight and higher stiffness. Generally, an aluminum or Nomex honeycomb core is used in composite applications requiring sandwich construction. However, the use of a composite core offers the potential for increased stiffness and strength, low thermal distortion compatible with that of the facesheets, the absence of galvanic corrosion and the ability to readily modify the core properties. In order to design, analyze and optimize these mirrors, knowledge of the mechanical properties of the core is essential. In this paper, the mechanical properties of a composite triangular cell core (often referred to as isogrid) are determined using finite element analysis of a representative unit cell. The core studied offers many advantages over conventional cores including increased thermal and dimensional stability, as well as low weight. Results are provided for the engineering elastic moduli of cores made of high stiffness composite material as a function of the ply layup and cell size. Finally, in order to illustrate the use of these properties in a typical application, a 1.4-m diameter composite mirror is analyzed using the finite element method, and the resulting stiffness and natural frequencies are presented.

Penado, F. Ernesto; Clark, James H., III; Walton, Joshua P.; Romeo, Robert C.; Martin, Robert N.

2007-09-01

117

When atoms are brought together to form molecules or solids the change in the kinetic energy of the core electrons can be an order of magnitude larger than the change in total energy. In spite of this, pseudopotential methods, which neglect the redistribution of the core electrons, give results very close to the fully self-consistent results. We explain this apparent

U. von Barth; C. D. Gelatt

1980-01-01

118

Calculation of load effect produced by ferrite core attached to wire above a ground plane

Ferrite cores are commonly attached to a cable\\/wire to reduce the electromagnetic noise emission from digital information equipment. In this paper, an equivalent circuit for the load effect produced by a ferrite core attached to a wire above a ground plane was considered. A practical method for determining the equivalent circuit parameters was presented, and the resultant load effect was

A. Z. Samir; O. Fujiwara

1999-01-01

119

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

120

The REBUS Experimental Programme for Burn-Up Credit

An international programme called REBUS (Reactivity tests for a direct evaluation of the Burn-Up credit on Selected irradiated LWR fuel bundles) for the investigation of the burn-up credit has been initiated by the Belgian Nuclear Research Center SCK.CEN and Belgonucleaire with the support of USNRC, EdF from France, VGB, representing German nuclear utilities and NUPEC, representing the Japanese industry. The programme aims to establish a neutronic benchmark for reactor physics codes. This benchmark will qualify the codes to perform calculations of the burn-up credit. The benchmark exercise investigates the following fuel types with associated burn-up: - Reference 3.3% enriched UO{sub 2} fuel; - Fresh commercial PWR UO{sub 2} fuel; - Irradiated commercial PWR UO{sub 2} fuel (51 GWd/tM); - Fresh PWR MOX fuel; - Irradiated PWR MOX fuel (20 GWd/tM). Reactivity effects are measured in the critical facility VENUS. Fission rate and flux distributions in the experimental bundles will be determined. The accumulated burn-up of all rods is measured non-destructively in a relative way by gross gamma-scanning, while some rods are examined by gamma-spectrometry for an absolute determination of the burn-up. Some rods will be analyzed destructively with respect to accumulated burn-up, actinides content and TOP-19 fission products (i.e. those non-gaseous fission products that have most implications on the reactivity). Additionally some irradiated rods have undergone a profilometry and length determination. The experimental implementation of the programme has started in 2000 with major changes in the VENUS critical facility. Gamma scans, profilometry, length determination and gamma-spectrometry measurements on the MOX fuel have been performed. In the course of October 2001 the first fresh fuel configuration will be investigated. In the same period the commercial irradiated fuel will arrive at the SCK.CEN hot cells and will be re-fabricated into fuel rodlets of 1 meter length. (authors)

D'hondt, Pierre; Van der Meer, Klaas; Baeten, Peter [SCK.CEN, Boeretang 200, 2400 Mol (Belgium); Marloye, Daniel; Lance, Benoit; Basselier, Jacques [Belgonucleaire (Belgium)

2002-07-01

121

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

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

122

Core and valence electrons in atoms and ions: configuration interaction calculations

Summary The partitioning of ground-state atoms or ions into inner spherical cores with radius ?b and outer valence regions extending from ?b to infinity is explored with the help of the expression \\u000a

Normand Desmarais; Sándor Fliszár

1996-01-01

123

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.

124

Effects of non-uniform core flow on peak cladding temperature: MOXY\\/SCORE sensitivity calculations

The MOXY\\/SCORE computer program is used to evaluate the potential effect on peak cladding temperature of selective cooling that may result from a nonuniform mass flux at the core boundaries during the blowdown phase of the LOFT L2-4 test. The results of this study indicate that the effect of the flow nonuniformity at the core boundaries will be neutralized by

1979-01-01

125

Model biases in high-burnup fast reactor simulations

A new code system called the Advanced Reactor Modeling Interface (ARMI) has been developed that loosely couples multiscale, multiphysics nuclear reactor simulations to provide rapid, user-friendly, high-fidelity full systems analysis. Incorporating neutronic, thermal-hydraulic, safety/transient, fuel performance, core mechanical, and economic analyses, ARMI provides 'one-click' assessments of many multi-disciplined performance metrics and constraints that historically require iterations between many diverse experts. The capabilities of ARMI are implemented in this study to quantify neutronic biases of various modeling approximations typically made in fast reactor analysis at an equilibrium condition, after many repetitive shuffles. Sensitivities at equilibrium that result in very high discharge burnup are considered ( and >20% FIMA), as motivated by the development of the Traveling Wave Reactor. Model approximations discussed include homogenization, neutronic and depletion mesh resolution, thermal-hydraulic coupling, explicit control rod insertion, burnup-dependent cross sections, fission product model, burn chain truncation, and dynamic fuel performance. The sensitivities of these approximations on equilibrium discharge burnup, k{sub eff}, power density, delayed neutron fraction, and coolant temperature coefficient are discussed. (authors)

Touran, N.; Cheatham, J.; Petroski, R. [TerraPower LLC, 11235 S.E. 6th St, Bellevue, WA 98004 (United States)

2012-07-01

126

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 captur...

Nabi, Jameel-Un

2014-01-01

127

An Analytical Solution For Multi-Core Energy Calculation With Consideration Of Leakage And

an analytical method to calculate the energy consumption efficiently and effectively for a given voltage problem in energy efficiency design is calculating the energy consumption for a design alternative is constant. Under this assumption, the calculation of energy consumption for a given voltage schedule

Quan, Gang

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

ATR WG-MOX Fuel Pellet Burnup Measurement by Monte Carlo - Mass Spectrometric Method

This paper presents a new method for calculating the burnup of nuclear reactor fuel, the MCWO-MS method, and describes its application to an experiment currently in progress to assess the suitability for use in light-water reactors of Mixed-OXide (MOX) fuel that contains plutonium derived from excess nuclear weapons material. To demonstrate that the available experience base with Reactor-Grade Mixed uranium-plutonium OXide (RGMOX) can be applied to Weapons-Grade (WG)-MOX in light water reactors, and to support potential licensing of MOX fuel made from weapons-grade plutonium and depleted uranium for use in United States reactors, an experiment containing WG-MOX fuel is being irradiated in the Advanced Test Reactor (ATR) at the Idaho National Engineering and Environmental Laboratory. Fuel burnup is an important parameter needed for fuel performance evaluation. For the irradiated MOX fuel’s Post-Irradiation Examination, the 148Nd method is used to measure the burnup. The fission product 148Nd is an ideal burnup indicator, when appropriate correction factors are applied. In the ATR test environment, the spectrum-dependent and burnup-dependent correction factors (see Section 5 for detailed discussion) can be substantial in high fuel burnup. The validated Monte Carlo depletion tool (MCWO) used in this study can provide a burnup-dependent correction factor for the reactor parameters, such as capture-to-fission ratios, isotopic concentrations and compositions, fission power, and spectrum in a straightforward fashion. Furthermore, the correlation curve generated by MCWO can be coupled with the 239Pu/Pu ratio measured by a Mass Spectrometer (in the new MCWO-MS method) to obtain a best-estimate MOX fuel burnup. A Monte Carlo - MCWO method can eliminate the generation of few-group cross sections. The MCWO depletion tool can analyze the detailed spatial and spectral self-shielding effects in UO2, WG-MOX, and reactor-grade mixed oxide (RG-MOX) fuel pins. The MCWO-MS tool only needs the MS-measured 239Pu/Pu ratio, without the measured isotope 148Nd concentration data, to determine the burnup accurately. MCWO-MS not only provided linear heat generation rate, Pu isotopic composition versus burnup, and burnup distributions within the WG-MOX fuel capsules, but also correctly pointed out the inconsistency in the large difference in burnups obtained by the 148Nd method.

Chang, Gray Sen I

2002-10-01

130

Ab initio effective core potentials (ECP's) have been generated to replace the Coulomb, exchange, and core-orthogonality effects of the chemically inert core electron in the transition metal atoms Sc to Hg. For the second and third transition series relative ECP's have been generated which also incorporate the mass--velocity and Darwin relativistic effects into the potential. The ab initio ECP's should facilitate valence electron calculations on molecules containing transition-metal atoms with accuracies approaching all-electron calculations at a fraction of the computational cost. Analytic fits to the potentials are presented for use in multicenter integral evaluation. Gaussian orbital valence basis sets are developed for the (3d,4s,4p), (4d,5s,5p), and (5d,6s,6p) orbitals of the first, second, and third transition series atoms, respectively. All-electron and valence-electron atomic excitation energies are also compared for the low-lying states of Sc--Hg, and the valence-electron calculations are found to reproduce the all-electron excitation energies (typically within a few tenths of an eV).

Hay, P.J.; Wadt, W.R.

1985-01-01

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

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

133

We describe the implementation of a parallel, in-core, integral-direct Hartree-Fock and density functional theory code for the efficient calculation of Hartree-Fock wave functions and density functional theory. The algorithm is based on a parallel master-slave algorithm, and the two-electron integrals calculated by a slave are stored in available local memory. To ensure the greatest computational savings, the master node keeps track of all integral batches stored on the different slaves. The code can reuse undifferentiated two-electron integrals both in the wave function optimization and in the evaluation of second-, third-, and fourth-order molecular properties. Superlinear scaling is achieved in a series of test examples, with speedups of up to 55 achieved for calculations run on medium-sized molecules on 16 processors with respect to the time used on a single processor. PMID:16365846

Fossgård, Eirik; Ruud, Kenneth

2006-02-01

134

This paper describes a new technique for determining the pin power in heterogeneous core calculations. It is based on a domain decomposition with overlapping sub-domains and a component mode synthesis technique for the global flux determination. Local basis functions are used to span a discrete space that allows fundamental global mode approximation through a Galerkin technique. Two approaches are given to obtain these local basis functions: in the first one (Component Mode Synthesis method), the first few spatial eigenfunctions are computed on each sub-domain, using periodic boundary conditions. In the second one (Factorized Component Mode Synthesis method), only the fundamental mode is computed, and we use a factorization principle for the flux in order to replace the higher order Eigenmodes. These different local spatial functions are extended to the global domain by defining them as zero outside the sub-domain. These methods are well-fitted for heterogeneous core calculations because the spatial interface modes are taken into account in the domain decomposition. Although these methods could be applied to higher order angular approximations - particularly easily to a SPN approximation - the numerical results we provide are obtained using a diffusion model. We show the methods' accuracy for reactor cores loaded with UOX and MOX assemblies, for which standard reconstruction techniques are known to perform poorly. Furthermore, we show that our methods are highly and easily parallelizable. (authors)

Guerin, P.; Baudron, A. M.; Lautard, J. J. [Commissariat a l'Energie Atomique, DEN/DANS/DM2S/SERMA/LENR, CEA Saclay, 91191 Gif sur Yvette (France)

2006-07-01

135

Intranodal burnup gradient correction in Analytic Function Expansion Nodal (AFEN) method

A new burnup correction model is proposed to take into account the effects of the intranodal burnup- and feedback-induced cross section variations on the neutron diffusion nodal calculation. This model homogenizes equivalently the node with cross section gradients into the homogeneous node, in accordance with the equivalence theory. This is accomplished by solving two single-node problems with the same current boundary conditions by expanding their flux solutions in fourth-order polynomials: one calculation for the node with cross section gradients and the other for the node with constant flux-weighted cross sections obtained from the former calculation. The new model was implemented in the AFEN method and its ability to treat the intranodal cross section gradients was tested on a benchmark problem. The results show that the new model eliminates almost all the errors in the nodal unknowns which are induced by the intranodal burnup gradients.

Noh, Jae Man; Cho, Nam Zin [Korea Advanced Institute of Science and Technology, Ysong-gu, Taejon (Korea, Republic of)

1995-12-31

136

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

137

NASA Astrophysics Data System (ADS)

Using a combination of atomistic tight-binding and valence force field model, the relative impact of quantum confinement and strain field on structural and optical properties in InAs/InP core/shell nanocrystals within a constant volume is systematically elucidated and quantified as a function of the aspect ratios (?) (Rcore/Rshell). The sensitivity of the structural optimization and aspect ratios (?) with the single-particle spectra, single-particle gaps, excitonic gaps, and ground e-h overlaps is successfully realized. Good agreement with experimental results is presented. These calculations provide an understanding of the interplay between structural characteristics and changes in electronic properties. Finally, a study of the desired InAs/InP core/shell nanocrystal provides a theoretical basis for tuning the spectral range to cover the novel applications.

Sukkabot, Worasak

2014-09-01

138

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

139

NASA Astrophysics Data System (ADS)

Herein, we report finite-element calculations of the effective (relative) permittivity of composite materials consisting of inclusions and inclusion arrays with a core-shell structure embedded in a surrounding host. The material making up the core of the two-dimensional structures, or cross sections of infinite three-dimensional objects (parallel, infinitely long, and identical cylinders) where the properties and characteristics are invariant along the perpendicular cross sectional plane, is assumed to have a negative real part of the permittivity, while the coating material (annular shell) is considered to be lossless. While strictly valid only in a dc situation, our analysis can be extended to treat electric fields that oscillate with time, provided that the wavelengths and attenuation lengths associated with the fields are much larger than the microstructure dimension in order that the homogeneous (effective-medium) representation of the composite structure makes sense. While one may identify features of the electrostatic resonance (ER) which are common to core-shell structures characterized by permittivities with real parts of opposite signs, it appears that the predicted ER positions are sensitive to the shell thickness and can be tuned through varying this geometric parameter. For example, we observe that the ER is broadened and shifted as the loss and the shell thickness are increased, respectively. We also argue that such core shell may also be valuable in controlling ER characteristics via polarization in an external electric field. In addition, by considering calculations of the electric field distribution, we find that the ER results in very strong and local-field enhancements into small parts of the shell perimeter. Our findings open up possibilities for the development of hybrid structures that could exploit the ER features for a particular application.

Mejdoubi, Abdelilah; Brosseau, Christian

2007-11-01

140

Core and valence electrons in atoms and ions: configuration interaction calculations

Summary. The partitioning of ground-state atoms or ions into inner spherical cores with radius $r_{\\\\rm b}^{\\\\phantom\\\\ast}$ and outer valence regions extending from $r_{\\\\rm b}^{\\\\phantom\\\\ast}$ to infinity is explored with the help of the expression $E^{\\\\rm v}=\\\\frac{1}{3}(T^{\\\\rm v}+2V^{\\\\rm v})$ for the valence-region energy (where $T^{\\\\rm v}\\\\!$ and $V^{\\\\rm v}$ are, respectively, the kinetic and potential energies of the ‘valence electrons’ $N^{\\\\rm

Normand Desmarais; Sándor Fliszár

1996-01-01

141

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

142

Quantum chemical valence-only electronic structure calculations have been performed for 22 states of NaK using full-valence configuration interaction, effective core potentials to provide the core-valence orthogonality constraints, and independent-particle core polarization functions (W. J. Stevens, D. D. Konowalow, and L. B. Ratcliff, J. Chem. Phys. 80, 1215-1224 (1984); L. B. Ratcliff, D. D. Konowalow, and W. J. Stevens, J. Mol.

Lyn Braxton Ratcliff

1989-01-01

143

Ab initio shell-model calculation for ^{18}O in a restricted no-core model space

We perform an ab initio shell-model calculation for ^{18}O in a restricted no-core model space, microscopically deriving a two-body effective interaction and introducing a minimal refinement of one-body energies in the spsd or spsdpf model space. Low-lying energy levels, except for the experimental 0_{2}^{+} and 2_{3}^{+} states, are better described in the spsdpf space than in the spsd space. The structure of low-lying energy levels is discussed with an emphasis on many-particle many-hole states beyond the four-particle two-hole configuration.

S. Fujii; B. R. Barrett

2009-02-12

144

NASA Astrophysics Data System (ADS)

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

145

On the explanation and calculation of anomalous reflood hydrodynamics in large PWR cores

Reflood hydrodynamics from large-scale (1:20) test facilities in Japan have yielded apparently anomalous behavior relative to FLECHT tests. Namely, even at reflooding rates below one inch per second, very large liquid volume fractions (10-15%) exist above the quench fronts shortly after flood begins; thus cladding temperature excursions are terminated early in the reflood phase. This paper discusses an explanation for this behavior: liquid films on the core's unheated rods. The experimental findings are shown to be correctly simulated with a new four-field (vapor, films, droplets) version of the best-estimate TRAC-PF1 computer code, TRAC-FF. These experimental and analytical findings have important implications for PWR large-break LOCA licensing.

Rodriguez, S.E.

1985-01-01

146

The US Department of Energy, Consumers Power Company, Exxon Nuclear Company, and General Public Utilities Nuclear Corporation have participated since 1979 in a cooperative Extended Burnup Demonstration Program. Under the program, standard ENC-fabricated reload fuel in the Big Rock Point and Oyster Creek reactor cores has been irradiated to discharge burnups at or beyond 35,000 MWD/MTU, one to two cycles beyond its originally projected exposure life. The program provides for examination of the fuel at poolside before and after each extended burnup cycle as well as for limited destructive hot cell examination. The 1984 progress report covers work performed under the EBD program between April 1983 and March 1984. Major milestones reached during the period include completion of a hot cell examination on four high burnup rods from Big Rock Point and of a poolside on the Oyster Creek EBD fuel at discharge. The hot cell examination of four rods at burnups to 37.2 GWD/MTU confirmed poolside measurements on the same fuel, showing the urania and gadolinia-bearing fuel rods to be in excellent condition. No major cladding degradation, pellet restructuring, or pellet-clad interaction was found in any of the samples examined. The Oyster Creek fuel, examined at an assembly average exposure of 34.5 GWD/MTU, showed good performance with regard to both diametral creepdown and clad oxide accumulation.

Exarhos, C.A.

1985-06-20

147

NASA Astrophysics Data System (ADS)

Since the identification of f-orbital contribution to the bonding in PaO+, investigations into Pa cations have hoped to characterize as many of the electronic states possible.1 Electronic states of the Pan+ (n=0-4) ions have been investigated using multi-reference spin-orbit configuration interaction (MR-SOCI). Initial investigations using Dunning style correlation consistent double-{?} basis sets are re-examined with a larger triple-{?} basis, with the hope of supporting the order of electronic states. Calculations using Hartree-Fock and CI calculations on the neutral atom did not produce the known order of states. A case study was deemed necessary on similar electron configurations present in the low energy states of Pa2+ more specifically those generated from the 5f26d1 and 5f16d2 configurations. Comparison in the Pa2+ ion is complicated by the lack of experimental results, but the states are presumed to be similar sequence as those in the neutral atom, with the addition of two electrons in the 7s shell. In evaluating the impact of inclusion of the outer core, calculations including valence-outer core correlation were completed for the 5d, 6s, and 6p shells of the Pa2+ ion. The magnitude of these individual shell correlation calculations will allow for identification of the energy level shifts associated with even and odd configurations, better describing the energy order in both the Pa2+ ion case study and for the neutral Pa atom. Upon completion of this aspect of the Pa neutral atom study, the knowledge of the energy levels in the Pan+ (n=0-4) family of ions will be greatly expanded, and may yield a model for future studies of atomic actinide systems. Gibson {et al.} Organometallics 2007, 26, 3947-3956.

Mrozik, Michael K.; Pitzer, Russell M.; Bursten, Bruce E.

2010-06-01

148

Electric-field-gradient calculations for systems with large extended-core-state contributions

Electric-field-gradient (EFG) calculations for TiO2 in the rutile structure using the standard full-potential linearized-augmented-plane-wave (LAPW) method have shown that the contribution of the Ti 3p semicore states is comparable to that of the valence electrons in contrast to other systems studied so far, where the latter dominate. This makes the treatment of the 3p and 4p states of Ti an

P. Blaha; D. J. Singh; P. I. Sorantin; K. Schwarz

1992-01-01

149

No-Core MCSM calculation for $^{10}$Be and $^{12}$Be low-lying spectra

The low-lying excited states of $^{10}$Be and $^{12}$Be are investigated within a no-core Monte Carlo Shell Model (MCSM) framework employing a realistic potential obtained via the Unitary Correlation Operator Method. The excitation energies of the 2$^+_1$ and 2$^+_2$ states and the B(E2;$\\,2^+_{1,2}\\rightarrow$ 0$^+_{g.s.}$) for $^{10}$Be in the MCSM with a standard treatment of spurious center-of-mass motion show good agreement with experimental data. Some properties of low-lying states of $^{10}$Be are studied in terms of quadrupole moments, E2 transitions and single-particle occupation numbers. The E2 transition probability of $^{10}$C, the mirror nucleus of $^{10}$Be, is also presented with a good agreement to experiment. The triaxial deformation of $^{10}$Be and $^{10}$C is discussed in terms of the B(E2) values. The removal of the spurious center-of-mass motion affects differently on various states: for instance, negligible effects on the 2$^+_1$ and 2$^+_2$ levels of $^{10}$Be, while significant and favorable shift for the 1$^-_1$ level. It is suggested that the description of $^{12}$Be needs a larger model space as well as some other higher excited states of $^{10}$Be, as an indicator that these are dominated by intruder configurations.

Lang Liu; Takaharu Otsuka; Noritaka Shimizu; Yutaka Utsuno; Robert Roth

2011-05-16

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

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

152

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

153

Dependence of transuranic content in spent fuel on fuel burnup

As the increasing demand for nuclear energy results in larger spent fuel volume, implementation of longer fuel cycles incorporating higher burnup are becoming common. Understanding the effect of higher burnup on the spent ...

Reese, Drew A. (Drew Amelia)

2007-01-01

154

We report hybrid density functional theory-molecular mechanics (DFT/MM) calculations performed for glycine in water solution at different pH values. In this paper, we discuss several aspects of the quantum mechanics-molecular mechanics (QM/MM) simulations where the dynamics and spectral binding energy shifts are computed sequentially, and where the latter are evaluated over a set of configurations generated by molecular or Car-Parrinello dynamics simulations. In the used model, core ionization takes place in glycine as a quantum mechanical (QM) system modeled with DFT, and the solution is described with expedient force fields in a large molecular mechanical (MM) volume of water molecules. The contribution to the core electronic binding energy from all interactions within and between the two (DFT and MM) parts is accounted for, except charge transfer and dispersion. While the obtained results were found to be in qualitative agreement with experiment, their precision must be qualified with respect to the problem of counter ions, charge transfer and optimal division of QM and MM parts of the system. Results are compared to those of a recent study [Ottoson et al., J. Am. Chem. Soc., 2011, 133, 3120]. PMID:23160171

Niskanen, Johannes; Arul Murugan, N; Rinkevicius, Zilvinas; Vahtras, Olav; Li, Cui; Monti, Susanna; Carravetta, Vincenzo; Agren, Hans

2013-01-01

155

From ab initio calculations to multiscale design of Si/C core-shell particles for Li-ion anodes.

The design of novel Si-enhanced nanocomposite electrodes that will successfully mitigate mechanical and chemical degradation is becoming increasingly important for next generation Li-ion batteries. Recently Si/C hollow core-shell nanoparticles were proposed as a promising anode architecture, which can successfully sustain thousands of cycles with high Coulombic efficiency. As the structural integrity and functionality of these heterogeneous Si materials depend on the strength and fracture energy of the active materials, an in-depth understanding of the interface and their intrinsic mechanical properties, such as fracture strength and debonding, becomes critical for the successful design of such and similar composites. Here, we first perform ab initio simulations to calculate these properties for lithiated a-Si/a-C interface structures and combine these results with linear elasticity expressions to model conditions that will avert fracture and debonding in these heterostructures. We find that the a-Si/a-C interface retains good adhesion even at high stages of lithiation. For average lithiated structures, we predict that the strong Si-C bonding averts fracture at the interface; instead, the structure ruptures within lithiated a-Si. From the calculated values and linear elastic fracture mechanics, we then construct a continuum level diagram, which outlines the safe regimes of operation in terms of the core and shell thickness and the state of charge. We believe that this multiscale approach can serve as a foundation for developing quantitative failure models and for subsequent development of flaw-tolerant anode architectures. PMID:24611810

Stournara, Maria E; Qi, Yue; Shenoy, Vivek B

2014-04-01

156

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

157

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

158

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

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

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

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 alpha-Al2O3 is found to originate from the presence of a core hole. Such an effect is less significant in MgO and

Isao Tanaka; Hirohiko Adachi

1996-01-01

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

Weapons-grade MOX PWR benchmark calculations

A simplified model of a Westinghouse pressurized water reactor (PWR) assembly has been proposed as a weapons-grade mixed-oxide (MOX) benchmark. The bundle design consists of a uniform plutonium loading that would be appropriate for a full MOX core. The benchmark consists of several state point calculations at zero burnup, a multicycle depletion to 45 MWd/kg, and several state point calculations at end of life. Calculations of the PWR MOX benchmark have been performed with the KENO Monte Carlo code and the VENTURE diffusion theory code using cross sections created using the SCALE system and with the HELIOS system. The benchmark has been proposed as a light water reactor MOX benchmark with initial results being submitted by the participants by September 1998. The complete specifications for the benchmark are available at http://www.engr.utk.edu/org/ans/benchmark/ansmoxbm.html.

Gehin, J.C. [Oak Ridge National Lab., TN (United States)

1998-12-31

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

Parametric Study of the Effect of Burnable Poison Rods for PWR Burnup Credit

The Interim Staff Guidance on burnup credit (ISG-8) issued by the United States Nuclear Regulatory Commission's (U.S. NRC) Spent Fuel Project Office recommends restricting the use of burnup credit to assemblies that have not used burnable absorbers. This recommended restriction eliminates a large portion of the currently discharged spent fuel assemblies from cask loading, and thus severely limits the practical usefulness of burnup credit. In the absence of readily available information on burnable poison rod (BPR) design specifications and usage in U.S. pressurized-water-reactors (PWRs), and the subsequent reactivity effect of BPR exposure on discharged spent nuclear fuel (SNF), NRC staff has indicated a need for additional information in these areas. In response, this report presents a parametric study of the effect of BPR exposure on the reactivity of SNF for various BPR designs, fuel enrichments, and exposure conditions, and documents BPR design specifications. Trends in the reactivity effects of BPRs are established with infinite pin-cell and assembly array calculations with the SCALE and HELIOS code packages, respectively. Subsequently, the reactivity effects of BPRs for typical initial enrichment and burnup combinations are quantified based on three-dimensional (3-D) KENO V.a Monte Carlo calculations with a realistic rail-type cask designed for burnup credit. The calculations demonstrate that the positive reactivity effect due to BPR exposure increases nearly linearly with burnup and is dependent on the number, poison loading, and design of the BPRs and the initial fuel enrichment. Expected typical reactivity increases, based on one-cycle BPR exposure, were found to be less than 1% {Delta}k. Based on the presented analysis, guidance is offered on an appropriate approach for calculating bounding SNF isotopic data for assemblies exposed to BPRs. Although the analyses do not address the issue of validation of depletion methods for assembly designs with BPRs, they do demonstrate that the effect of BPRs is generally well behaved and that independent codes and cross-section libraries predict similar results. The report concludes with a discussion of the issues for consideration and recommendations for inclusion of SNF assemblies exposed to BPRs in criticality safety analyses using burnup credit for dry cask storage and transport.

Wagner, J.C.

2001-09-28

166

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

167

Overview of effects of burnup credit on cask design

A number of opportunities exist to increase the productivity of the next generation of spent fuel shipping casks. One of the opportunities being evaluated by Sandia National Laboratories (SNL) under the sponsorship of the Department of Energy's (DOE's) Cask System Development Program at Idaho is the implementation of burnup credit in the design of spent fuel casks. Burnup credit is defined as accounting for the effects of fuel burnup in the criticality safety analysis of a cask. Burnup is fissile nuclide reduction and fission and activation product buildup, that result in reduced reactivity of fuel after it has been used for producing power in a nuclear reactor. An important step in implementing burnup credit is to establish an understanding as to how burnup credit affects cask design. The purpose of this paper is to provide an overview of the effects of burnup credit on spent fuel cask design. This will be accomplished by addressing the following questions: What functions does a cask perform. How does burnup credit potentially affect a cask. How does burnup credit affect basket design. What design or functional requirements are likely to be affected by burnup credit. 1 fig., 4 tabs.

Allen, G.C.

1988-11-02

168

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

169

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

170

A Monte Carlo burnup code linking MCNP and REBUS.

The REBUS-3 burnup code, used in the ANL RERTR Program, is a very general code that uses diffusion theory (DIF3D) to obtain the fluxes required for reactor burnup analyses. Diffusion theory works well for most reactors. However, to include the effects of exact geometry and strong absorbers that are difficult to model using diffusion theory, a Monte Carlo method is required. MCNP, a general-purpose, generalized-geometry, time-dependent, Monte Carlo transport code, is the most widely used Monte Carlo code. This paper presents a linking of the MCNP code and the REBUS burnup code to perform these difficult burnup analyses. The linked code will permit the use of the full capabilities of REBUS which include non-equilibrium and equilibrium burnup analyses. Results of burnup analyses using this new linked code are also presented.

Hanan, N. A.

1998-10-19

171

Modelling the high burnup UO 2 structure in LWR fuel

The concept of a burnup threshold for the formation of the high burnup UO2 structure (HBS) is supported by experimental data, which also reveal that a transition zone exists between the normal UO2 structure and the fully developed HBS. From the analysis of radial xenon profiles measured by EPMA a threshold burnup is obtained in the range 60–75 GW d\\/t

K. Lassmann; C. T. Walker; J. van de Laar; F. Lindström

1995-01-01

172

S and 4 Reactor: Operating Lifetime and Estimates of Temperature and Burnup Reactivity Coefficients

The S and 4 reactor has a sectored, Mo-14%Re solid core for avoidance of single point failures in reactor cooling and Closed Brayton Cycle (CBC) energy conversion. The reactor is loaded with UN fuel, cooled with a He-Xe gas mixture at {approx}1200 K and operates at steady thermal power of 550 kW. Following a launch abort accident, the axial and radial BeO reflectors easily disassemble upon impact so that the bare reactor is subcriticial when submerged in wet sand or seawater and the core voids are filled with seawater. Spectral Shift Absorber (SSA) additives have been shown to increase the UN fuel enrichment and significantly reduce the total mass of the reactor. This paper investigates the effects of SSA additions on the temperature and burnup reactivity coefficients and the operational lifetime of the S and 4 reactor. SSAs slightly decrease the temperature reactivity feedback coefficient, but significantly increase the operating lifetime by decreasing the burnup reactivity coefficient. With no SSAs, fuel enrichment is only 58.5 wt% and the estimated operating lifetime is the shortest (7.6 years) with the highest temperature and burnup reactivity feedback coefficients (-0.2709 cent /K and -1.3470 $/atom%). With europium-151 and gadolinium-155 additions, the enrichment (91.5 and 94 wt%) and operating lifetime (9.9 and 9.8 years) of the S and 4 reactor are the highest while the temperature and burnup reactivity coefficients (-0.2382 and -0.2447 cent /K; -0.9073 and 0.8502 $/atom%) are the lowest.

King, Jeffrey C.; El-Genk, Mohamed S. [Institute for Space and Nuclear Power Studies, University of New Mexico, Albuquerque, NM 87131 (United States); Chemical and Nuclear Engineering Dept., University of New Mexico, Albuquerque, NM 87131 (United States)

2006-01-20

173

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

174

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

175

ATR WG-MOX Fuel Pellet Burnup Measurement by Monte Carlo - Mass Spectrometric Method

This paper presents a new method for calculating the burnup of nuclear reactor fuel, the MCWO-MS method, and describes its application to an experiment currently in progress to assess the suitability for use in light-water reactors of Mixed-OXide (MOX) fuel that contains plutonium derived from excess nuclear weapons material. To demonstrate that the available experience base with Reactor-Grade Mixed uranium-plutonium

G. S. Chang; Gray Sen I

2002-01-01

176

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

177

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

178

NASA Astrophysics Data System (ADS)

We performed Finite Difference Time Domain (FDTD) calculation to investigate the enhancement of optical properties such as light scattering and absorption of Au-hybridized TiO2 core-shell structures which can lead to the improvement of photocatalytic and solar cell performance. The results showed that by hybridization of Au as core and TiO2 as shell provides the significant enhancement of light scattering and absorption. Furthermore, the tuning of scattering resonance wavelength may be achieved by varying the diameter of Au core. Our result suggests that hybridization Au and TiO2, with proper introduction of interband states in TiO2, can increase and color-tune the photocatalytic efficiency and solar cell performance of TiO2 nanostructures.

Lee, Jubok; Lee, Sun-Hee; Kim, Min Su; Shin, Hyungjung; Kim, Jeongyong

2014-09-01

179

Impact of High Burnup on PWR Spent Fuel Characteristics

Reducing the burden of management of spent nuclear fuel is important to the future of nuclear energy. The impact of higher pressurized water reactor (PWR) fuel burnup is examined in this paper from the perspective of its impact on spent-fuel radioactivity, decay heat, and plutonium content. The necessary fresh fuel enrichments to achieve high burnup in PWRs with the same

Xu Zhiwen; Mujid S. Kazimi; Michael J. Driscoll

2005-01-01

180

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

181

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

182

Burnup credit applications in a high-capacity truck cask

General Atomics (GA) has designed two legal weight truck (LWT) casks, the GA-4 and GA-9, to carry four pressurized-water-reactor (PWR) and nine boiling-water-reactor (BWR) fuel assemblies, respectively. GA plans to submit applications for certification to the US Nuclear Regulatory Commission (NRC) for the two casks in mid-1993. GA will include burnup credit analysis in the Safety Analysis Report for Packaging (SARP) for the GA-4 Cask. By including burnup credit in the criticality safety analysis for PWR fuels with initial enrichments above 3% U-235, public and occupation risks are reduced and cost savings are realized. The GA approach to burnup credit analysis incorporates the information produced in the US Department of Energy Burnup Credit Program. This paper describes the application of burnup credit to the criticality control design of the GA-4 Cask.

Boshoven, J.K.

1992-09-01

183

Needs of reliable nuclear data and covariance matrices for Burnup Credit in JEFF-3 library

NASA Astrophysics Data System (ADS)

Burnup Credit (BUC) is the concept which consists in taking into account credit for the reduction of nuclear spent fuel reactivity due to its burnup. In the case of PWR-MOx spent fuel, studies pointed out that the contribution of the 15 most absorbing, stable and non-volatile fission products selected to the credit is as important as the one of the actinides. In order to get a "best estimate" value of the keff, biases of their inventory calculation and individual reactivity worth should be considered in criticality safety studies. This paper enhances the most penalizing bias towards criticality and highlights possible improvements of nuclear data for the 15 FPs of PWRMOx BUC. Concerning the fuel inventory, trends in function of the burnup can be derived from experimental validation of the DARWIN-2.3 package (using the JEFF-3.1.1/SHEM library). Thanks to the BUC oscillation programme of separated FPs in the MINERVE reactor and fully validated scheme PIMS, calculation over experiment ratios can be accurately transposed to tendencies on the FPs integral cross sections.

Chambon, A.; Santamarina, A.; Riffard, C.; Lavaud, F.; Lecarpentier, D.

2013-03-01

184

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

185

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

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

2011-01-01

186

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 (>300dpa

S. A. Maloy; M. Toloczko; James I. Cole; Thak Sang Byun

2011-01-01

187

Methodology for the Weapons-Grade MOX Fuel Burnup Analysis in the Advanced Test Reactor

A UNIX BASH (Bourne Again SHell) script CMO has been written and validated at the Idaho National Laboratory (INL) to couple the Monte Carlo transport code MCNP with the depletion and buildup code ORIGEN-2 (CMO). The new Monte Carlo burnup analysis methodology in this paper consists of MCNP coupling through CMO with ORIGEN-2, and is therefore called the MCWO. MCWO is a fully automated tool that links the Monte Carlo transport code MCNP with the radioactive decay and burnup code ORIGEN-2. MCWO is capable of handling a large number of fuel burnup and material loading specifications, Advanced Test Reactor (ATR) lobe powers, and irradiation time intervals. MCWO processes user input that specifies the system geometry, initial material compositions, feed/removal specifications, and other code-specific parameters. Calculated results from MCNP, ORIGEN-2, and data process module calculations are output in succession as MCWO executes. The principal function of MCWO is to transfer one-group cross-section and flux values from MCNP to ORIGEN-2, and then transfer the resulting material compositions (after irradiation and/or decay) from ORIGEN-2 back to MCNP in a repeated, cyclic fashion. The basic requirements of MCWO are a working MCNP input file and some additional input parameters; all interaction with ORIGEN-2 as well as other calculations are performed by CMO. This paper presents the MCWO-calculated results for the Reduced Enrichment Research and Test Reactor (RERTR) experiments RERTR-1 and RERTR-2 as well as the Weapons-Grade Mixed Oxide (WG-MOX) fuel testing in ATR. Calculations performed for the WG-MOX test irradiation, which is managed by the Oak Ridge National Laboratory (ORNL), supports the DOE Fissile Materials Disposition Program (FMDP). The MCWO-calculated results are compared with measured data.

G. S. Chang

2005-08-01

188

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

189

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

190

NASA Astrophysics Data System (ADS)

We investigate the X-ray photoelectron spectroscopy (XPS) binding energies of As 3d in Si for various defects in neutral and charged states by first-principles calculation. It is found that the complexes of a substitutional As and a vacancy in charged and neutral states explain the experimentally observed unknown peak very well.

Kishi, Hiroki; Miyazawa, Miki; Matsushima, Naoki; Yamauchi, Jun

2014-02-01

191

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

192

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

193

Characterization of High Burnup Fuel for Safety Related Fuel Testing

Fuel Assemblies designed and fabricated by Westinghouse Electric Sweden (WSE) to reach high burnup have been operated in the Leibstadt nuclear power plant (KKL) for seven cycles attaining an assembly average burnup above 60 MWd\\/kgU. The irradiation conditions in KKL featured linear heat generation rates ranging from 250 W\\/cm early in life down to 100W\\/cm in the last cycle and

Guido LEDERGERBER; Sousan ABOLHASSANI; Magnus LIMBÄCK; Roger J. LUNDMARK; Kurt-Åke MAGNUSSON

2006-01-01

194

Modelling the high burnup UO 2 structure in LWR fuel

NASA Astrophysics Data System (ADS)

The concept of a burnup threshold for the formation of the high burnup UO 2 structure (HBS) is supported by experimental data, which also reveal that a transition zone exists between the normal UO 2 structure and the fully developed HBS. From the analysis of radial xenon profiles measured by EPMA a threshold burnup is obtained in the range 60-75 GW d/t U. The lower value is considered to be the threshold for the onset of the HBS and the higher value the threshold for the fully developed HBS. Xenon depletion in the transition zone and the fully developed HBS can be described by a simple model. At local burnups above 120 GW d/t U the xenon generated is in equilibrium with the xenon lost to the fission gas pores and the concentration does not fall below 0.25 wt%. The TRANSURANUS burnup model TUBRNP predicts reasonably well the penetration of the HBS and the associated xenon depletion up to a cross section average burnup of approximately 70 GW d/t U.

Lassmann, K.; Walker, C. T.; van de Laar, J.; Lindström, F.

1995-10-01

195

Burn-up and neutron economy of accelerator-driven reactor

It is desirable to have only a small reactivity change in the large burn-up of a solid fuel fast reactor, so that the number of replacements or shuffling of the fuel can be reduced, and plant factor accordingly increased. Also, this reduces the number of control rods needed for the change in burn-up reactivity. In subcritical operation, power controlled by beam power is suggested, but this practice is not as economical as the use of control rods and makes more careful operation of the accelerator is required due to changes in the wake field. In subcritical operation, even a slightly subcritical one, the safety problems associated with a hard neutron spectrum can be alleviated. Neutron leakage from a flattened core, which is needed for operation of the critical fast reactor can be lessen by using the non flat core which has good neutron economy. For generating nuclear energy, it is essential to have a high neutron economy, although breeding the fuel is not welcomed in the present political climate, as is needed for transmuting long lived fission products. In contrast to the breeder, the accelerator driven reactor can separate the energy production from fuel production and processing. Thus, it is suited for non-proliferation of nuclear material by prohibiting the processing and production of fuel in the unrestricted area so this can be only done in international controlled areas which are restricted and remote.

Takahashi, H.; Yang, W.; An, Y.; Yamazaki, Y.

1997-07-01

196

The KUGEL computer code is designed to perform thermal/hydraulic analysis and coated-fuel particle performance calculations for axisymmetric pebble bed reactor (PBR) cores. This computer code was developed as part of a Department of Energy (DOE)-funded study designed to verify the published core performance data on PBRs. The KUGEL code is designed to interface directly with the 2DB code, a two-dimensional neutron diffusion code, to obtain distributions of thermal power, fission rate, fuel burnup, and fast neutron fluence, which are needed for thermal/hydraulic and fuel performance calculations. The code is variably dimensioned so that problem size can be easily varied. An interpolation routine allows variable mesh size to be used between the 2DB output and the two-dimensional thermal/hydraulic calculations.

Shamasundar, B.I.; Fehrenbach, M.E.

1981-05-01

197

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

198

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

199

Modeling Cladding-Coolant Heat Transfer of High-Burnup Fuel During RIA

This paper describes a model for the cladding-coolant heat transfer of high burnup fuel during a Reactivity Initiated Accident (RIA) which is implemented in the fuel performance code FRAPTRAN 1.2. The minimum stable film boiling temperature, affected by the subcooling and the clad oxidation, is modeled by a modified Henry correlation. This accounts for the effects of thermal properties of the cladding surface on the transient temperature drop during liquid-solid contact. The transition boiling regime is described as the interpolation of the heat flux between two anchor points on the boiling curve: the Critical Heat Flux (CHF) and minimum stable film boiling. The CHF correlation is based on the Zuber hydrodynamic model multiplied by a subcooling factor. Frederking correlation is chosen to model the film boiling regime. The heat conduction through the oxide layer of the cladding surface of high burnup fuel is calculated by solving heat conduction equations with thermal properties of zirconia taken from MATPRO. This model is validated in the FRAPTRAN code for test cases of both high burnup and fresh test fuel rods including the burnup level (0--56 MW d/kg), peak fuel enthalpy deposit (70--190 cal/g), degree of subcooling (0--80 deg. C), and extent of oxidation (0--25 micron). The modified code demonstrates the capability of differentiating between the departure from nucleate boiling (DNB) and none-DNB cases. The predicted peak cladding temperature (PCT) and duration of DNB achieves generally good agreement with the experimental data. It is found that the cladding surface oxidation of high burnup fuel causes an early rewetting of cladding or suppresses DNB due to two factors: 1) Thick zirconia layer may delay the heat conducted to the surface while keeping the surface heat transfer in the most effective nucleate boiling regime. 2) The transient liquid-solid contact resulting from vapor breaking down would cause a lower interface temperature for an oxidized surface, essentially raises the minimum stable film boiling temperature. (authors)

Wenfeng Liu; Kazimi, Mujid S. [Massachusetts Institute of Technology, 77 Massachusetts avenue Cambridge, MA 02139-4307 (United States)

2006-07-01

200

We present a new pseudospectral approach for incorporating many-body, nonlocal exact exchange interactions to understand the formation of electron gases in core-shell nanowires. Our approach is efficiently implemented in the open-source software package PAMELA (Pseudospectral Analysis Method with Exchange & Local Approximations) that can calculate electronic energies, densities, wavefunctions, and band-bending diagrams within a self-consistent Schrodinger-Poisson formalism. The implementation of both local and nonlocal electronic effects using pseudospectral methods is key to PAMELA's efficiency, resulting in significantly reduced computational effort compared to finite-element methods. In contrast to the new nonlocal exchange formalism implemented in this work, we find that the simple, conventional Schrodinger-Poisson approaches commonly used in the literature (1) considerably overestimate the number of occupied electron levels, (2) overdelocalize electrons in nanowires, and (3) significantly underestima...

Long, Andrew W; 10.1063/1.4754603

2012-01-01

201

The tracing algorithm that is implemented in the geometrical module of Monte-Carlo transport code MCU is applied to calculate the volume fractions of original materials by spatial cells of the mesh that overlays problem geometry. In this way the 3D combinatorial geometry presentation of the problem geometry, used by MCU code, is transformed to the user defined 2D or 3D bit-mapped ones. Next, these data are used in the volume fraction (VF) method to approximate problem geometry by introducing additional mixtures for spatial cells, where a few original materials are included. We have found that in solving realistic 2D and 3D core problems a sufficiently fast convergence of the VF method takes place if the spatial mesh is refined. Virtually, the proposed variant of implementation of the VF method seems as a suitable geometry interface between Monte-Carlo and S{sub n} transport codes. (authors)

Gurevich, M. I.; Oleynik, D. S. [RRC Kurchatov Inst., Kurchatov Sq., 1, 123182, Moscow (Russian Federation); Russkov, A. A.; Voloschenko, A. M. [Keldysh Inst. of Applied Mathematics, Miusskaya Sq., 4, 125047, Moscow (Russian Federation)

2006-07-01

202

Based on the result of micro-gamma scanning of a fuel pin irradiated to high burnup in a commercial PWR, the radial distribution of chemical forms of fission products (FPs) in LWR fuel pins was theoretically predicted by a thermochemical computer code SOLGASMIX-PV. The absolute amounts of fission products generated in the fuel was calculated by ORIGEN-2 code, and the radial

Kouki MORIYAMA; Hirotaka FURUYA

1997-01-01

203

The HELIOS reactor-physics computer program system was used to simulate the burnup of UO{sub 2} fuel in three VVER reactors. The manner in which HELIOS was used in these simulations is described. Predictions of concentrations for actinides up to {sup 244}Cm and for isotopes of neodymium were compared with laboratory-measured values. Reasonable agreement between calculated and measured values was seen for experimental samples from a fuel rod in the interior of an assembly.

Murphy, B.D.; Kravchenko, J.; Lazarenko, A.; Pavlovitchev, A.; Sidorenko, V.; Chetverikov, A.

2000-03-01

204

Comparison of experimentally determined spent-fuel compositions with ORIGEN 2 calculations

The specific experimental measurements of interest here involve the determination of parameters related to the actinide and fission product composition of samples from five elements taken from fuel assemblies discharged from the Turkey Point Unit 3 PWR. Two fuel assemblies were obtained for the purposes of nondestructive and destructive assay. These assemblies were initially fueled with 448 kg of UO/sub 2/ enriched to 2.556 wt % /sup 235/U and then irradiated for 851 full-power days. Five elements were then removed, and assay samples were taken from each element near the core midplane. The relevant parameters measured were /sup 148/Nd//sup 238/U, /sup 239/Pu//sup 238/U, and the isotopic compositions of U, Pu, Kr, and Xe. Fuel depletion calculations were performed using the updated ORIGEN2 PWR model. The burnup of the fuel was determined by adjusting the ORIGEN2 fuel burnup to match the experimentally determined /sup 148/Nd//sup 238/U ratio for each fuel element. The resulting burnup was then used to calculate the other experimentally determined parameters listed above. The agreement between ORIGEN2 and the experimental results is very good, with the average error for five samples being < 4% for most parameters. Based on this comparison, it appears that the ORIGEN2 computer code is capable of accurately calculating the composition of irradiated fuel from a modern PWR. However, well-characterized experimental measurements should continue to be obtained for validation purposes because the calculated values of many nuclides, particularly the minor actinides, still have significant uncertainties.

Croff, A.G.

1981-01-01

205

Burnup measurements with the Los Alamos fork detector

The fork detector system can determine the burnup of spent-fuel assemblies. It is a transportable instrument that can be mounted permanently in a spent-fuel pond near a loading area for shipping casks, or be attached to the storage pond bridge for measurements on partially raised spent-fuel assemblies. The accuracy of the predicted burnup has been demonstrated to be as good as 2% from measurements on assemblies in the United States and other countries. Instruments have also been developed at other facilities throughout the world using the same or different techniques, but with similar accuracies. 14 refs., 2 figs., 2 tabs.

Bosler, G.E.; Rinard, P.M.

1991-01-01

206

Isotopic and criticality validation for actinide-only burnup credit

The techniques used for actinide-only burnup credit isotopic validation and criticality validation are presented and discussed. Trending analyses have been incorporated into both methodologies, requiring biases and uncertainties to be treated as a function of the trending parameters. The isotopic validation is demonstrated using the SAS2H module of SCALE 4.2, with the 27BURNUPLIB cross section library; correction factors are presented for each of the actinides in the burnup credit methodology. For the criticality validation, the demonstration is performed with the CSAS module of SCALE 4.2 and the 27BURNUPLIB, resulting in a validated upper safety limit.

Fuentes, E.; Lancaster, D.; Rahimi, M.

1997-07-01

207

Whole-core LEU fuel demonstration in the ORR

A whole-core demonstration of LEU fuel in the ORR is expected to begin during November 1985. Fuel elements will contain U/sub 3/Si/sub 2/ at 4.8 Mg U/m/sup 3/ and shim rod fuel followers will contain U/sub 3/Si/sub 2/ at 3.5 Mg U/m/sup 3/. Fuel fabrication is underway at B and W, CERCA, and NUKEM, with shipments scheduled to commence in October. The primary objectives of the demonstration are to provide data for validation of LEU and mixed-core fuel cycle calculations and to provide a large-scale demonstration of the acceptable performance of production-line U/sub 3/Si/sub 2/ fuel elements. It is planned to approach the full LEU core through a series of mixed cores. Measurements to be made include flux distribution, reactivity swing, control rod worths, cycle length, fuel discharge burnup, gamma heating rates, ..beta../sub eff/l, and isothermal temperature coefficient. Measurements will also be made on fresh LEU and fresh HEU critical configurations. Preliminary safety approval has been received and the final safety assessment is being reviewed.

Snelgrove, J.L.; Bretscher, M.M.; Cornella, R.J.; Hobbs, R.W.

1985-01-01

208

Study on Ultra-Long Life,Small U-Zr Metallic Fuelled Core With Burnable Poison

A conceptual design for a 50 MWe sodium cooled, U-Pu-Zr metallic fuelled, fast reactor core, which aims at a core lifetime of 30 years, has been performed [1]. As for the compensation for a large burn-up reactivity through 30 years, an axially movable reflector, which is located around the core, carries the major part of it and a burnable poison does the rest. This concept has achieved not only a long core lifetime but also a high discharged burn-up. On this study, a conceptual design for a small fast reactor loading U-Zr metallic fuelled core instead of U-Pu-Zr fuelled core has been conducted, based on the original core arrangement of 4S reactor [2]. Within the range of this study including safety requirements, adopting the burnable poison would be effective to construct a core concept that achieves both a long lifetime and a high discharged burn-up. (authors)

Kenji Tsuji; Hiromitsu Inagaki; Akira Nishikawa [Chubu Electric Power Company Inc., 1, Higashi-shincho Higashi-ku, Nagoya-shi, ACH 461-8680 (Japan); Hisato Matsumiya; Yoshiaki Sakashita; Yasuyuki Moriki; Mitsuaki Yamaoka; Norihiko Handa [Toshiba Corporation (Japan)

2002-07-01

209

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 the validation of the method against available experimental data. The experimental data that are applicable to burnup credit include chemical assay data for the validation of the isotopic prediction models, fresh fuel critical experiments for the validation of criticality calculations for various cask geometries, and reactor restart critical data to validate criticality calculations with spent fuel. The methodology has been specifically developed to be simple and generally applicable, therefore giving rise to uncertainties or sensitivities which are identified and quantified in terms of a percent bias in k{sub eff}. Implementation issues affecting licensing requirements and operational procedures are discussed briefly. 24 refs., 3 tabs.

Brady, M.C. (Oak Ridge National Lab., TN (USA)); Sanders, T.L. (Sandia National Labs., Albuquerque, NM (USA))

1991-01-01

210

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

211

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

212

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

213

NASA Astrophysics Data System (ADS)

The eddy current testing method is widely used to evaluate conductive pieces. This method requires an adequate mathematical model which is able to describe the complicated interactions between the source and induced currents, primary and secondary, the fields and the flaws in materials. This paper describes a model which predicts the apparent changes in the impedance of an absolute ferrite-cored probe in axially symmetric non destructive testing (NDT) configurations. Originally, this model is based on coupled electromagnetic quantities principle. To include the contribution of the magnetic environment, the state variable chosen is the current because the magnetic magnetization is replaced by the fictional equivalent currents. The obtained modelling results are validated by comparison to finite element computations. Once validated, the suggested model is not only applied to calculation of probe's impedance in the presence of a defect inside the load but it is also applied to determine geometrical and physical characteristics of the eddy current non destructive testing (ECNDT) device. This half-numerical technique with a very weak time of simulation can be used for the design of new probes and offers a simple solution to the inversion problem. The model is implemented within a software tool (CECM: coupling electromagnetic circuits method) developed in MATLAB environment.

Zerguini, S.; Maouche, B.; Latreche, M.; Feliachi, M.

2009-12-01

214

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

215

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

216

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

217

Next Generation CANDU Core Physics Innovations

NG CANDU is the 'Next Generation' CANDU{sup R} reactor, aimed at producing electrical power at a capital cost significantly less than that of the current reactor designs. A key element of cost reduction is the use of H{sub 2}O as coolant and Slightly Enriched Uranium fuel in a tight D{sub 2}O-moderated lattice. The innovations in the CANDU core physics result in substantial improvements in economics as well as significant enhancements in reactor licensability, controllability, and waste reduction. The full-core coolant-void reactivity in NG CANDU is about -3 mk. Power coefficient is substantially negative. Fuel burnup is about three times the current natural-uranium burnup. (authors)

Chan, P.S.W.; Hopwood, J.M.; Love, J.W. [Atomic Energy of Canada Ltd., Ontario (Canada)

2002-07-01

218

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

219

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

220

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

221

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

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

Thermochemical behavior of the core melt in the VVER-type reactor at severe accident is discussed. Experimental information\\u000a gained made it possible to construct a thermodynamic model of the O-U-Zr-Fe system. The model describes the immiscibility\\u000a of the oxide and metal phases of the core melt and makes it possible to estimate their densities. Parameters of the model\\u000a were obtained by

V. D. Ozrin; O. V. Tarasov; V. F. Strizhov; A. S. Filippov

2010-01-01

224

Thermochemical behavior of the core melt in the VVER-type reactor at severe accident is discussed. Experimental information gained made it possible to construct a thermodynamic model of the O-U-Zr-Fe system. The model describes the immiscibility of the oxide and metal phases of the core melt and makes it possible to estimate their densities. Parameters of the model were obtained by

V. D. Ozrin; O. V. Tarasov; V. F. Strizhov; A. S. Filippov

2010-01-01

225

Modeling of WWER-440 fuel pin behavior at extended burn-up

Currently, there is an ongoing effort to increase fuel discharge burn-up of all LWRs fuel including WWERs as much as possible in order to decrease power production cost. Therefore, burn-up is expected to be increased from 60 to 70MWd\\/kgU. The change in the fuel radial power distribution as a function of fuel burn-up can affect the radial fuel temperature distribution

Moustafa S El-Koliel; Attya A Abou-Zaid; A. A El-Kafas

2004-01-01

226

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

227

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

228

National Technical Information Service (NTIS)

The Core-Quasiparticle Coupling Model (CQCM) for odd-mass nuclei, which is based on dynamical field theory and the Bardeen-Cooper-Schrieffer (BCS) method, has been applied to two problems. In the first, a study of Pauli exchange effects for the odd partic...

P. B. Semmes

1985-01-01

229

Core-level excitations are generated by absorption of high-energy radiation such as X-rays. To describe these energetically high-lying excited states theoretically, we have implemented a variant of the algebraic-diagrammatic construction scheme of second-order ADC(2) by applying the core-valence separation (CVS) approximation to the ADC(2) working equations. Besides excitation energies, the CVS-ADC(2) method also provides access to properties of core-excited states, thereby allowing for the calculation of X-ray absorption spectra. To demonstrate the potential of our implementation of CVS-ADC(2), we have chosen medium-sized molecules as examples that have either biological importance or find application in organic electronics. The calculated results of CVS-ADC(2) are compared with standard TD-DFT/B3LYP values and experimental data. In particular, the extended variant, CVS-ADC(2)-x, provides the most accurate results, and the agreement between the calculated values and experiment is remarkable. PMID:25130619

Wenzel, Jan; Wormit, Michael; Dreuw, Andreas

2014-10-01

230

NASA Technical Reports Server (NTRS)

A computer code base on an improved vortex filament/vortex core method for predicting aerodynamic characteristics of slender wings with edge vortex separations is developed. The code is applicable to camber wings, straked wings or wings with leading edge vortex flaps at subsonic speeds. The prediction of lifting pressure distribution and the computer time are improved by using a pair of concentrated vortex cores above the wing surface. The main features of this computer program are: (1) arbitrary camber shape may be defined and an option for exactly defining leading edge flap geometry is also provided; (2) the side edge vortex system is incorporated.

Pao, J. L.; Mehrotra, S. C.; Lan, C. E.

1982-01-01

231

Comparison of XSUSA and "Two-Step" Approaches for Full Core Uncertainty Quantification

While there are multiple sources of error that are introduced into the standard computational regime for simulating reactor cores, rigorous uncertainty analysis methods are available primarily for quantifying the effects of cross section uncertainties. Two methods for propagating cross section uncertainties through core simulators are the XSUSA statistical approach and the 'Two-Step' method. The XSUSA approach, which is based on the SUSA code package, is fundamentally a stochastic sampling method. Alternatively, the Two-Step method utilizes generalized perturbation theory in the first step and stochastic sampling in the second step. The consistency of these two methods in quantifying uncertainties in the multiplication factor and in the core power distribution will be examined in the framework of phase I-3 of the UAM Benchmark. Using the TMI core as a base model for analysis, the XSUSA and Two-Step methods are applied with certain limitations and the results are compared to those produced by other stochastic sampling-based codes. Based on the uncertainty analysis results, conclusions are made for which method is currently a more viable option for computing uncertainties in burnup and transient calculations.

Yankov, Artem [University of Michigan; Klein, M. [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS); Jessee, Matthew Anderson [ORNL; Zwermann, W. [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS); Velkov, Kiril [Gesellschaft fur Anlagen; Pautz, Andreas [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS); Collins, Benjamin [University of Michigan; Downar, Thomas [University of Michigan

2012-01-01

232

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

233

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 of the cracked fuel pellets and to mechanical constraints imposed by pellet-cladding friction. As part of a program to provide a better understanding of brittle-type failure of Zircaloy fuel cladding by pellet-cladding interaction (PCI) phenomenon, the stress-rupture properties and microstructural characteristics of high-burnup spent fuel cladding have been under investigation. This paper reports the results of the microstructural examinations by optical microscopy, scanning (SEM), 100-keV transmission (TEM), and 1 MeV high-voltage (HVEM) electron microscopies of the fractured spent fuel cladding with a specific empahsis on a correlation of the structural characteristics with the fracture behavior.

Chung, H.M.

1983-06-01

234

Analysis of high-burnup fuel performance during load-follow operation

In Japan, an objective of the burnup extension of nuclear fuel is to raise the licensing limit of burnup from 39 to 48 GWd\\/t for pressurized water reactors (PWRs) in the near future. Because of an increasing ratio of nuclear power generation, the necessity of the load-follow operation, which responds flexibly to changing power demands, is more apparent. To evaluate

T. Matsui; K. Fukuya; M. Kinoshita

1987-01-01

235

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

236

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

237

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

238

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

239

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

240

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

241

Impact of Integral Burnable Absorbers on PWR Burnup Credit Criticality Safety Analyses

The concept of taking credit for the reduction in reactivity of burned or spent nuclear fuel (SNF) due to fuel burnup is commonly referred to as burnup credit. The reduction in reactivity that occurs with fuel burnup is due to the net reduction of fissile nuclide concentrations and the production of actinide and fission-product neutron absorbers. The change in the inventory of these nuclides with fuel burnup, and the consequent reduction in reactivity, is dependent upon the depletion environment. Therefore, the use of burnup credit necessitates consideration of all possible fuel operating conditions, including the use of integral burnable absorbers (IBAs). The Interim Staff Guidance on burnup credit [1] issued by the Nuclear Regulatory Commission's (NRC) Spent Fuel Project Office recommends licensees restrict the use of burnup credit to assemblies that have not used burnable absorbers (e.g., IBAs or burnable poison rods, BPRs). This restriction eliminates a large portion of the currently discharged spent fuel assemblies from cask loading, and thus severely limits the practical usefulness of burnup credit. The reason for this restriction is that the presence of burnable absorbers during depletion hardens the neutron spectrum, resulting in lower {sup 235}U depletion and higher production of fissile plutonium isotopes. Enhanced plutonium production has the effect of increasing the reactivity of the fuel at discharge and beyond. Consequently, an assembly exposed to burnable absorbers may have a slightly higher reactivity for a given burnup than an assembly that has not been exposed to burnable absorbers. This paper examines the effect of IBAs on reactivity for various designs and enrichment/poison loading combinations as a function of burnup. The effect of BPRs, which are typically removed during operation, is addressed elsewhere [2].

Sanders, C.E.

2001-07-20

242

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

243

The U.S. Departments of Agriculture and Interior Abandoned Mine Lands (AML) Initiative is focused on the evaluation of the effect of past mining practices on the water quality and the riparian and aquatic habitats of impacted stream reaches downstream from historical mining districts located primarily on Federal lands. This problem is manifest in the eleven western states (west of longitude 102 degrees) where the majority of hardrock mines that had past production are located on Federal lands. In areas of temperate climate and moderate to heavy precipitation, the effects of rapid chemical and physical weathering of sulfides exposed on mine-waste dumps and acidic drainage from mines have resulted in elevated metal concentrations in the stream water and stream-bed sediment. The result of these mineral weathering processes has an unquantified impact on the quality of the water and the aquatic and riparian habitats that may limit their recreational resource value. One of the confounding factors in these studies is the determination of the component of metals derived from hydrothermally altered but unmined portions of these drainage basins. Several watersheds have been studied to evaluate the effects of acid mine drainage and acid rock drainage on the near-surface environment. The Animas River watershed in southwestern Colorado contains a large number of past-producing metal mines that have affected the watershed. Beginning in October 1996, the U.S. Geological Survey (USGS) began a collaborative study of these effects under the USGS-AML Initiative. In this report, we present the radionuclide and geochemical analytical results of sediment coring during 1997-1999 from two cores from oxbow lakes 0.5 mi. upstream from the 32nd Street Bridge near Durango, Colo., and from three cores from beaver ponds within the Mineral Creek drainage basin near Silverton, Colo.

Church, Stanley E.; Rice, Cyndi A.; Marot, Marci E.

2008-01-01

244

Calculation of the Phenix end-of-life test 'Control Rod Withdrawal' with the ERANOS code

The Inst. of Radiological Protection and Nuclear Safety (IRSN) acts as technical support to French public authorities. As such, IRSN is in charge of safety assessment of operating and under construction reactors, as well as future projects. In this framework, one current objective of IRSN is to evaluate the ability and accuracy of numerical tools to foresee consequences of accidents. Neutronic studies step in the safety assessment from different points of view among which the core design and its protection system. They are necessary to evaluate the core behavior in case of accident in order to assess the integrity of the first barrier and the absence of a prompt criticality risk. To reach this objective one main physical quantity has to be evaluated accurately: the neutronic power distribution in core during whole reactor lifetime. Phenix end of life tests, carried out in 2009, aim at increasing the experience feedback on sodium cooled fast reactors. These experiments have been done in the framework of the development of the 4. generation of nuclear reactors. Ten tests have been carried out: 6 on neutronic and fuel aspects, 2 on thermal hydraulics and 2 for the emergency shutdown. Two of them have been chosen for an international exercise on thermal hydraulics and neutronics in the frame of an IAEA Coordinated Research Project. Concerning neutronics, the Control Rod Withdrawal test is relevant for safety because it allows evaluating the capability of calculation tools to compute the radial power distribution on fast reactors core configurations in which the flux field is very deformed. IRSN participated to this benchmark with the ERANOS code developed by CEA for fast reactors studies. This paper presents the results obtained in the framework of the benchmark activity. A relatively good agreement was found with available measures considering the approximations done in the modeling. The work underlines the importance of burn-up calculations in order to have a fine core concentrations mesh for the calculation of the power distribution. (authors)

Tiberi, V. [Institut de Radioprotection et de Surete Nucleaire IRSN, PSN-EXP/SNC/LNR, BP 17, 92262 Fontenay-aux-Roses (France)

2012-07-01

245

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

246

Hybrid Reduced Order Modeling Algorithms for Reactor Physics Calculations

NASA Astrophysics Data System (ADS)

Reduced order modeling (ROM) has been recognized as an indispensable approach when the engineering analysis requires many executions of high fidelity simulation codes. Examples of such engineering analyses in nuclear reactor core calculations, representing the focus of this dissertation, include the functionalization of the homogenized few-group cross-sections in terms of the various core conditions, e.g. burn-up, fuel enrichment, temperature, etc. This is done via assembly calculations which are executed many times to generate the required functionalization for use in the downstream core calculations. Other examples are sensitivity analysis used to determine important core attribute variations due to input parameter variations, and uncertainty quantification employed to estimate core attribute uncertainties originating from input parameter uncertainties. ROM constructs a surrogate model with quantifiable accuracy which can replace the original code for subsequent engineering analysis calculations. This is achieved by reducing the effective dimensionality of the input parameter, the state variable, or the output response spaces, by projection onto the so-called active subspaces. Confining the variations to the active subspace allows one to construct an ROM model of reduced complexity which can be solved more efficiently. This dissertation introduces a new algorithm to render reduction with the reduction errors bounded based on a user-defined error tolerance which represents the main challenge of existing ROM techniques. Bounding the error is the key to ensuring that the constructed ROM models are robust for all possible applications. Providing such error bounds represents one of the algorithmic contributions of this dissertation to the ROM state-of-the-art. Recognizing that ROM techniques have been developed to render reduction at different levels, e.g. the input parameter space, the state space, and the response space, this dissertation offers a set of novel hybrid ROM algorithms which can be readily integrated into existing methods and offer higher computational efficiency and defendable accuracy of the reduced models. For example, the snapshots ROM algorithm is hybridized with the range finding algorithm to render reduction in the state space, e.g. the flux in reactor calculations. In another implementation, the perturbation theory used to calculate first order derivatives of responses with respect to parameters is hybridized with a forward sensitivity analysis approach to render reduction in the parameter space. Reduction at the state and parameter spaces can be combined to render further reduction at the interface between different physics codes in a multi-physics model with the accuracy quantified in a similar manner to the single physics case. Although the proposed algorithms are generic in nature, we focus here on radiation transport models used in support of the design and analysis of nuclear reactor cores. In particular, we focus on replacing the traditional assembly calculations by ROM models to facilitate the generation of homogenized cross-sections for downstream core calculations. The implication is that assembly calculations could be done instantaneously therefore precluding the need for the expensive evaluation of the few-group cross-sections for all possible core conditions. Given the generic natures of the algorithms, we make an effort to introduce the material in a general form to allow non-nuclear engineers to benefit from this work.

Bang, Youngsuk

247

The radial distribution of plutonium in high burnup UO 2 fuels

NASA Astrophysics Data System (ADS)

A new model (TUBRNP) is described which predicts the radial power density distribution as a function of burnup (and hence the radial burnup profile as a function of time) together with the radial profile of uranium and plutonium isotopes. Comparisons between measurements and the predictions of the TUBRNP model are made on fuels with enrichments in the range 2.9 to 8.25% and with burnups between 21000 and 64000 MWd/t. It is shown to be in excellent agreement with experimental measurements and is a marked improvement on earlier versions.

Lassmann, K.; O'Carroll, C.; van de Laar, J.; Walker, C. T.

1994-02-01

248

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

249

NASA Astrophysics Data System (ADS)

New rhenium(I) dicarbonyl complexes containing cis-{Re(CO)2}+ fragment with redox non-innocent NNS donor ligands L1/L2 (L1, 1-methyl-2-{(o-thiomethyl)phenylazo}imidazole and L2, 1-ethyl-2-{(o-thioethyl)phenylazo}imidazole) having general formula cis-[ReX(CO)3(L1/L2)] (1/2) (X = Cl (1) and Br (2)) have been synthesized and characterized by both experimental and theoretical studies. The structural confirmation has been carried out for 1b. The complexes show quasireversible ReI/ReII oxidation and ligand based reduction in the cyclic voltammetric studies. The electronic structure and the nature of Resbnd CO bonding has been explained by means of DFT and NBO calculations. The spin allowed singlet-singlet electronic transitions of 1b and 2b have been calculated with TDDFT method, and the experimental spectra of the complexes have been discussed on this basis.

Jana, Mahendra Sekhar; Pramanik, Ajoy Kumar; Sarkar, Deblina; Biswas, Sujan; Mondal, Tapan Kumar

2013-09-01

250

Assessment of high-burnup LWR fuel response to reactivity-initiated accidents

The economic advantages of longer fuel cycle, improved fuel utilization and reduced spent fuel storage have been driving the nuclear industry to pursue higher discharge burnup of Light Water Reactor (LWR) fuel. A design ...

Liu, Wenfeng, Ph.D. Massachusetts Institute of Technology

2007-01-01

251

Design strategies for optimizing high burnup fuel in pressurized water reactors

This work is focused on the strategy for utilizing high-burnup fuel in pressurized water reactors (PWR) with special emphasis on the full array of neutronic considerations. The historical increase in batch-averaged discharge ...

Xu, Zhiwen, 1975-

2003-01-01

252

National Technical Information Service (NTIS)

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 characteriz...

M. D. DeHart

1996-01-01

253

NASA Astrophysics Data System (ADS)

Recent observations of broadened spectral lines suggest that the dynamics of molecular clouds (MCs) are dominated by supersonic, turbulent motion. Furthermore, one observes sheet-like and filamentary structures on all resolvable scales. There exists recent theoretical work concerning the evolution of turbulent MC's and their fragmentation into dense cloud cores, which are presumably the progenitors of new stars. These calcualtions allowed to study the dynamics in MCs under the influence of thermal pressure and self-gravity. However, in high mass SF regions, the massive stars start interacting with their parental cloud via stellar winds and ionizing radiation. These feedback processes disrupt the MC by heating and ionization, but could at the same time induce new star formation by compression of the MC. This work focused on the development of a new computational method which allows the inclusion of ionizing radiation in hydrodynamical simulations using Smoothed Particle Hydrodynamics (sph). In contrast to grid-based methods, sph is especially suited for investigating the fragmentation of turbulent media due to the independence from a fixed grid geometry. The implementation of ionizing radiation into sph is an important step toward the ability to model numerically the feedback of young stars on their molecular clouds. The first application is the ionization-driven implosion of density enhancements in MC's. http://www.mpia-hd.mpg.de/THEORY/preprints/kessel/1999/dissertation/head.html

Kessel-Deynet, Olaf

1999-12-01

254

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

255

Burnup simulations and spent fuel characteristics of ZrO 2 based inert matrix fuels

NASA Astrophysics Data System (ADS)

Reducing the inventory of long lived isotopes that are contained in spent nuclear fuel is essential for maximizing repository capacity and extending the lifetime of related storage. Because of their non-fertile matrices, inert matrix fuels (IMF's) could be an ideal vehicle for using light-water reactors to help decrease the inventory of plutonium and other transuranics (neptunium, americium, curium) that are contained within spent uranium oxide fuel (UOX). Quantifying the characteristics of spent IMF is therefore of fundamental importance to determining its effect on repository design and capacity. We consider six ZrO 2 based IMF formulations with different transuranic loadings in a 1-8 IMF to UOX pin-cell arrangement. Burnup calculations are performed using a collision probability model where transport of neutrons through space is modeled using fuel to moderator transport and escape probabilities. The lethargy dependent neutron flux is treated with a high resolution multigroup thermalization method. The results of the reactor physics model are compared to a benchmark case performed with Montebruns and indicate that the approach yields reliable results applicable to high-level analyses of spent fuel isotopics. The data generated show that a fourfold reduction in the radiological and integrated thermal output is achievable in single recycle using IMF, as compared to direct disposal of an energy equivalent spent UOX.

Schneider, E. A.; Deinert, M. R.; Herring, S. T.; Cady, K. B.

2007-03-01

256

Methodologies to assess potential lifetime limits for extended burnup nuclear fuel

METHODOLOGIES TO ASSESS POTENTIAL LIFETIME LIMITS FOR EXTENDED BURNUP NUCLEAR FUEL A Thesis by CURTIS VINCENT DE VORE Submitted to the Graduate College of Texas A&M University in partial fulfillment of the requirements for the degree.... , Texas A&M University Chairman of Advisory Committee: Dr. K. L. Peddicord The investigation of the extended burnup performance of ceramic U02 fuel clad in Zircaloy tubes is of particular importance to nuclear utilities because of the very attractive...

De Vore, Curtis Vincent

2012-06-07

257

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

258

Improved lumped model for thermal analysis of high burn-up nuclear fuel rods

High burn-up nuclear fuel elements have been intensively studied for prolonged lifetime of existing reactors and for next-generation advanced reactors. This paper presents an improved lumped-differential formulation for one-dimensional transient heat conduction in a heat generating cylinder with temperature-dependent thermo-physical properties typical of high burn-up nuclear fuel rods. Two-points Hermite approximations for integrals (H1,1\\/H1,1) were used to obtain the average

Auro C. Pontedeiro; Renato M. Cotta; Jian Su

2008-01-01

259

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

260

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

261

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

262

Development of core fuel management code system for WWER-type reactors

In this article, a core fuel management program for hexagonal pressurized water type WWER reactors (CFMHEX) has been developed, which is based on advanced three-dimensional nodal method and integrated with thermal hydraulic code to realize the coupling of neutronics and thermal-hydraulics. In CFMHEX, all these feedback effects such as burnup, power distribution, moderator density, and control rod insertion are considered.

Bang-Yang XIA; Tao WANG; Zhong-Sheng XIE

2006-01-01

263

NASA Astrophysics Data System (ADS)

The work presented in this thesis is a continuation of a master's thesis research project conducted by the author to gain insight into the applicability of inverse methods to developing adaptive simulation capabilities for core physics problems. Use of adaptive simulation is intended to improve the fidelity and robustness of important core attributes predictions such as core power distribution, thermal margins and core reactivity. Adaptive simulation utilizes a selected set of past and current reactor measurements of reactor observables, i.e. in-core instrumentations readings, to adapt the simulation in a meaningful way. A meaningful adaption will result in high fidelity and robust adapted core simulators models. To perform adaption, we propose an inverse theory approach in which the multitudes of input data to core simulators, i.e. reactor physics and thermal-hydraulic data, are to be adjusted to improve agreement with measured observables while keeping core simulators models unadapted. At a first glance, devising such adaption for typical core simulators models would render the approach impractical. This follows, since core simulators are based on very demanding computational models, i.e. based on complex physics models with millions of input data and output observables. This would spawn not only several prohibitive challenges but also numerous disparaging concerns. The challenges include the computational burdens of the sensitivity-type calculations required to construct Jacobian operators for the core simulators models. Also, the computational burdens of the uncertainty-type calculations required to estimate the uncertainty information of core simulators input data presents a demanding challenge. The concerns however are mainly related to the reliability of the adjusted input data. We demonstrate that the power of our proposed approach is mainly driven by taking advantage of this unfavorable situation. Our contribution begins with the realization that to obtain numerical solutions to demanding computational models, matrix methods are often employed to produce approximately equivalent discretized computational models that may be manipulated further by computers. The discretized models are described by matrix operators that are often rank-deficient, i.e. ill-posed. We introduce a novel set of matrix algorithms, denoted by Efficient Subspace Methods (ESM), intended to approximate the action of very large, dense, and numerically rank-deficient matrix operators. We demonstrate that significant reductions in both computational and storage burdens can be attained for a typical BWR core simulator adaption problem without compromising the quality of the adaption. We demonstrate robust and high fidelity adaption utilizing a virtual core, e.g. core simulator predicted observables with the virtual core either based upon a modified version of the core simulator whose input data are to be adjusted or an entirely different core simulator. Further, one specific application of ESM is demonstrated, that is being the determination of the uncertainties of important core attributes such as core reactivity and core power distribution due to the available ENDF/B cross-sections uncertainties. The use of ESM is however not limited to adaptive core simulation techniques only, but a wide range of engineering applications may easily benefit from the introduced algorithms, e.g. machine learning and information retrieval techniques highly depends on finding low rank approximations to large scale matrices. In the appendix, we present a stand-alone paper that presents a generalized framework for ESM, including the mathematical theory behind the algorithms and several demonstrative applications that are central to many engineering arenas---(a) sensitivity analysis, (b) parameter estimation, and (c) uncertainty analysis. We choose to do so to allow other engineers, applied mathematicians, and scientists from other scientific disciplines to take direct advantage of ESM without having to sail across the sea of reactor core calculations.

Abdel-Khalik, Hany Samy

264

NASA Astrophysics Data System (ADS)

Frozen-core full-CI calculations of frequency-dependent dipole polarizabilities (FDPs) of ground state BeH 2 at the experimental distance of R = 2.506 a0 have been performed using an extended set of 208 contracted GTO functions [9s9p5d3f] on Be and [9s8p6d] on H involving about 58 × 10 6 symmetry-adapted Slater determinants. The Casimir-Polder integral was then evaluated analytically using eight optimized imaginary frequencies chosen according to a recently developed interpolation technique which allows for the evaluation of the three dipole dispersion constants for the BeH 2-BeH 2 homodimer, from which isotropic C6 and anisotropy ?6 coefficients are derived for the first time.

Bendazzoli, Gian Luigi; Monari, Antonio; Figari, Giuseppe; Rui, Marina; Costa, Camilla; Magnasco, Valerio

2005-10-01

265

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

266

Impacts of a high-burnup spent fuel on a geological disposal system design

The influence of a burnup increase of a spent nuclear fuel on a deep geological disposal system was evaluated in this study. First, the impact of a burnup increase on each aspect related to thermal and nuclear safety concerns was quantified. And then, the tunnel length, excavation volume, and the raw materials for a cast insert, copper, bentonite, and backfill needed to constitute a disposal system were comprehensively analyzed based on the spent fuel inventory to generate 1 Terawatt-year (TWa), to establish the overall effects and consequences on a geological disposal. As a result, impact of a burnup increase on the criticality safety and radiation shielding was shown to be negligible. The disposal area, however, is considerably affected because of a higher thermal load. And, it is reasonable to use a canister such as the Korean Reference Disposal Canister (KDC-1) containing 4 spent fuels up to 50 GWD/MtU, and to use a canister containing 3 spent fuels beyond 50 GWD/MtU. Although a considerable increased, 33 % in the tunnel length and 30 % in the excavation volume, was observed as the burnup increases from 50 to 60 GWD/MtU, because a decrease in the canister needs can offset an increase in the excavation volume, it can be concluded that a burnup increase of a spent fuel is not a critical concern for a geological disposal of a spent fuel. (authors)

Cho, D.K.; Lee, Y.; Lee, J.Y.; Choi, H.J.; Choi, J.W. [Korea Atomic Energy Research Institute, Yuseong-gu, Daejeon-city (Korea, Republic of)

2007-07-01

267

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

268

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

269

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

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

Post Irradiation Examination for Advanced Materials at Burnups Exceeding the Current Limit

Permitting fuel to be irradiated to higher burnups limits can reduce the amount of spent nuclear fuel (SNF) requiring storage and/or disposal and enable plants to operate with longer more economical cycle lengths and/or at higher power levels. Therefore, Framatome ANP (FANP) and the B&W Owner's Group (BWOG) have introduced a new fuel rod design with an advanced M5 cladding material and have irradiated several test fuel rods through four cycles. The U.S. Department of Energy (DOE) joined FANP and the BWOG in supporting this project during its final phase of collecting and evaluating high burnup data through post irradiation examination (PIE).

John H. Strumpell

2004-12-31

272

Fuel service conditions proposed for the very high temperature reactor 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 coated particle fuel development program in the 1980s. Of particular concern are the high burnup and high temperatures expected in the

John T. Maki; David A. Petti; Darrell L. Knudson; Gregory K. Miller

2007-01-01

273

A model of radial profiles of burnup, heat release, and accumulation of plutonium isotopes is described. The model was developed for use in the mechanistic RTOP fuel element code. The model is based on theoretical ideas about the mechanisms leading to the formation of the radial burnup profile and a simplified description of the neutron spectrum in the reactor, employing

S. Yu. Kurchatov; V. V. Likhanskii; A. A. Sorokin; O. V. Khoruzhii

2002-01-01

274

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...

275

ACCURATE NUCLEAR FUEL BURNUP ANALYSES. Quarterly Progress Report No. 6, March-May 1963

Work has continued on the development of accurate methods for ; determining nuclear fuel burnup. Thermal flux capsule GEV-1, has completed its ; irradiation in MTR and its cooling period. Hardened flux capsule GEV-2 is ; presently in the VBWR and has received about 2 x 10Â¹â¹ nvt. A fast fission ; capsule GEV-3, containing uranium-238 and thorium232 under cadmium,

B. F. Rider; C. P. Ruiz; J. P. Jr. Peterson; P. S. Jr. Luke

1963-01-01

276

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

277

Burnup verification measurements at a US nuclear utility using the FORK measurement system

The FORK measurement system, designed at Los Alamos National Laboratory (LANL) for the International Atomic Energy Agency (IAEA) safeguards program, has been used to examine spent reactor fuel assemblies at Duke Power Company`s Oconee Nuclear Station. The FORK system measures the passive neutron and gamma-ray emission from spent fuel assemblies while in the storage pool. These measurements can be correlated with burnup and cooling time, and can be used to verify the reactor site records. Verification measurements may be used to help ensure nuclear criticality safety when burnup credit is applied to spent fuel transport and storage systems. By taking into account the reduced reactivity of spent fuel due to its burnup in the reactor, burnup credit results in more efficient and economic transport and storage. The objectives of these tests are to demonstrate the applicability of the FORK system to verify reactor records and to develop optimal procedures compatible with utility operations. The test program is a cooperative effort supported by Sandia National Laboratories, the Electric Power Research Institute (EPRI), Los Alamos National Laboratory, and the Duke Power Company.

Ewing, R.I. [Sandia National Labs., Albuquerque, NM (United States); Bosler, G.E. [Los Alamos National Lab., NM (United States); Walden, G. [Duke Power Co., Charlotte, NC (United States)

1993-08-01

278

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

279

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

National Technical Information Service (NTIS)

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 u...

D. E. Mueller, I. C. Gauld

2005-01-01

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

I. C. Gauld; D. E. Mueller

2005-01-01

281

Most Monte Carlo neutronics analyses are performed for fresh cores. To model snapshots of the cores at different stages during burnup using MCNP, a method is presented that uses lumped fission product (LFP) cross sections generated by the WIMS-ANL code and processed for use in MCNP. Results of analyses for four very different reactor cores using MTR-type and Russian-designed fuel assemblies, with LEU and HEU fuels, are provided to demonstrate the use of this method.

Hanan, N. A.

1998-10-14

282

We cannot observe the magnetic field inside the earth's core directly, but there is likely to be a large toroidal part of 10-100 Gauss which, together with the dipole component, could produce a magnetic torque on the inner core that tends to rotate it. Estimates based on dynamo calculations give torques of 10 --N m which is large enough to

D. Gubbins

1981-01-01

283

Sodium-cooled mixed-oxide core design studies are performed with a target burnup of 150 GWd/t and possible measures against the recriticality issues in core disruptive accidents. Four types of core are compared from the viewpoints of core performance and reliability. Results show that all the types of core satisfy the target and that a homogeneous core with an axial blanket partial elimination subassembly is the superior concept, although experimental demonstration is required of molten fuel motion for mitigation of recriticality following fuel melting and loss of fuel pin integrity.

Mizuno, Tomoyasu; Niwa, Hajime [Japan Nuclear Cycle Development Institute (Japan)

2004-05-15

284

Test plan for high-burnup fuel cladding behavior under loss-of- coolant accident conditions

Excessive oxidation, hydriding, and extensive irradiation damage occur in high-burnup fuel cladding, and as result, mechanical properties of high-burnup fuels are degraded significantly. This may influence the current fuel cladding failure limits for loss-of- coolant-accident (LOCA) situations, which are based on fuel cladding behavior for zero burnup. To avoid cladding fragmentation and fuel dispersal during a LOCA, 10 CFR 50.46 requires that peak cladding temperature shall not exceed 1204 degrees C (2200 degrees F) and that total oxidation of the fuel cladding nowhere exceeds 0.17 times total cladding thickness before oxidation. Because of the concern, a new experimental program to investigate high-burnup fuel cladding behavior under LOCA situations has been initiated under the sponsorship of the U.S. Nuclear Regulatory Commission. A hot-cell test plan to investigate single-rod behavior under simulated LOCA conditions is described in this paper. In the meantime, industry fuel design and operating conditions are expected to undergo further changes as more advanced cladding materials are developed. Under these circumstances, mechanical properties of high-burnup fuel cladding require further investigation so that results from studies on LOCA, reactivity- initiated-accident (RIA), operational transient, and power-ramping situations, can be extrapolated to modified or advanced cladding materials and altered irradiation conditions without repeating major integral experiments in test reactors. To provide the applicable data base and mechanistic understanding, tests will be conducted to determine dynamic and static fracture toughness and tensile properties. Background and rationale for selecting the specific mechanical properties tests are also described.

Chung, H.M.; Neimark, L.A.; Kassner, T.F.

1996-10-01

285

Weapons-Grade MOX Fuel Burnup Characteristics in Advanced Test Reactor Irradiation

Mixed oxide (MOX) test capsules prepared with weapons-derived plutonium have been irradiated to a burnup of 50 GWd/t. The MOX fuel was fabricated at Los Alamos National Laboratory (LANL) by a master-mix process and has been irradiated in the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL). Previous withdrawals of the same fuel have occurred at 9, 21, 30, 40, and 50 GWd/t. Oak Ridge National Laboratory (ORNL) manages this test series for the Department of Energy’s Fissile Materials Disposition Program (FMDP). A UNIX BASH (Bourne Again SHell) script CMO has been written and validated at the Idaho National Laboratory (INL) to couple the Monte Carlo transport code MCNP with the depletion and buildup code ORIGEN-2 (CMO). The new Monte Carlo burnup analysis methodology in this paper consists of MCNP coupling through CMO with ORIGEN-2(MCWO). MCWO is a fully automated tool that links the Monte Carlo transport code MCNP with the radioactive decay and burnup code ORIGEN-2. The fuel burnup analyses presented in this study were performed using MCWO. MCWO analysis yields time-dependent and neutron-spectrum-dependent minor actinide and Pu concentrations for the ATR small I-irradiation test position. The purpose of this report is to validate both the Weapons-Grade Mixed Oxide (WG-MOX) test assembly model and the new fuel burnup analysis methodology by comparing the computed results against the neutron monitor measurements and the irradiated WG-MOX post irradiation examination (PIE) data.

G. S. Chang

2006-07-01

286

24. A CORE WORKER DISPLAYS THE CORE BOX AND CORES ...

24. A CORE WORKER DISPLAYS THE CORE BOX AND CORES FOR A BRASS GATE VALVE BODY MADE ON A CORE BOX, CA. 1950. - Stockham Pipe & Fittings Company, 4000 Tenth Avenue North, Birmingham, Jefferson County, AL

287

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

288

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

289

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

290

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

Germina Ilas; Ian C Gauld

2011-01-01

291

Influence of burnup on heat-rating-to-melting for UOâ-PuOâ fuel

This paper summarizes the results of an irradiation test (HEDL-P-20) conducted to evaluate the influence of burnup on the linear heat rating to incipient fuel melting, Q'\\/sub m\\/, for 75% UOâ-25% PuOâ fuel. The very short term irradiation was conducted in Row 1 of the EBR-II to simulate 115% of FTR full power using an irradiation history similar to that

R. D. Leggett; R. B. Baker; D. S. Dutt; S. A. Chastain

1974-01-01

292

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

293

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

294

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

295

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

296

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

297

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

298

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

299

Comparison of nuclear data uncertainty propagation methodologies for PWR burn-up simulations

Several methodologies using different levels of approximations have been developed for propagating nuclear data uncertainties in nuclear burn-up simulations. Most methods fall into the two broad classes of Monte Carlo approaches, which are exact apart from statistical uncertainties but require additional computation time, and first order perturbation theory approaches, which are efficient for not too large numbers of considered response functions but only applicable for sufficiently small nuclear data uncertainties. Some methods neglect isotopic composition uncertainties induced by the depletion steps of the simulations, others neglect neutron flux uncertainties, and the accuracy of a given approximation is often very hard to quantify. In order to get a better sense of the impact of different approximations, this work aims to compare results obtained based on different approximate methodologies with an exact method, namely the NUDUNA Monte Carlo based approach developed by AREVA GmbH. In addition, the impact of different covariance data is studied by comparing two of the presently most complete nuclear data covariance libraries (ENDF/B-VII.1 and SCALE 6.0), which reveals a high dependency of the uncertainty estimates on the source of covariance data. The burn-up benchmark Exercise I-1b proposed by the OECD expert group "Benchmarks for Uncertainty Analysis in Modeling (UAM) for the Design, Operation and Safety Analysis of LWRs" is studied as an example application. The burn-up simulations are performed with the SCALE 6.0 tool suite.

Carlos Javier Diez; Oliver Buss; Axel Hoefer; Dieter Porsch; Oscar Cabellos

2014-11-04

300

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

301

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

302

NASA Astrophysics Data System (ADS)

The main aim of this research is the development and validation of computational schemes for advanced lattice codes. The advanced lattice code which forms the primary part of this research is "DRAGON Version4". The code has unique features like self shielding calculation with capabilities to represent distributed and mutual resonance shielding effects, leakage models with space-dependent isotropic or anisotropic streaming effect, availability of the method of characteristics (MOC), burnup calculation with reaction-detailed energy production etc. Qualified reactor physics codes are essential for the study of all existing and envisaged designs of nuclear reactors. Any new design would require a thorough analysis of all the safety parameters and burnup dependent behaviour. Any reactor physics calculation requires the estimation of neutron fluxes in various regions of the problem domain. The calculation goes through several levels before the desired solution is obtained. Each level of the lattice calculation has its own significance and any compromise at any step will lead to poor final result. The various levels include choice of nuclear data library and energy group boundaries into which the multigroup library is cast; self shielding of nuclear data depending on the heterogeneous geometry and composition; tracking of geometry, keeping error in volume and surface to an acceptable minimum; generation of regionwise and groupwise collision probabilities or MOC-related information and their subsequent normalization thereof, solution of transport equation using the previously generated groupwise information and obtaining the fluxes and reaction rates in various regions of the lattice; depletion of fuel and of other materials based on normalization with constant power or constant flux. Of the above mentioned levels, the present research will mainly focus on two aspects, namely self shielding and depletion. The behaviour of the system is determined by composition of resonant and predominantly scattering isotopes. When the concentration of resonant isotopes is small, its presence does not affect the flux shape which is smooth. But when the concentration becomes high, there will be dips in the flux where resonances of the isotopes occur. This will affect the reaction rate, which is a product of cross section and flux. The reaction rate will thus be lower than that when one does not consider the flux dip. This is the phenomenon of self shielding. Self shielding treatment is thus a very important aspect of reactor lattice analysis code. This needs to be correctly modelled to obtain a physically sound and acceptable solution. In this research we will be looking into behaviour of the advanced self shielding models that have been incorporated in the code DRAGON Version4. The self shielding models are primarily classified into two broad groups, which are based on "equivalence in dilution" and "subgroup approach". These self shielding models will be tested against a variety of lattices which include Canada Deuterium Uranium (CANDU-6), CANDU-New Generation (CANDU-NG), Light Water Reactor (LWR), and High Conversion Light Water Reactor (HCLWR). The fuel composition will vary from natural uranium oxide to enriched uranium oxide and plutonium-uranium mixed oxide (MOX). We will also consider the presence of strong neutron absorbers like gadolinium and dysprosium in the lattice. The coolant/moderator chosen for the analysis will be light water/heavy water or a combination. The lattice geometry will vary from square, hexagonal and annular. Thus a broad spectrum of lattices will be analysed to assess the behaviour of advanced self shielding models. The results obtained using DRAGON will be validated against that obtained using Monte Carlo code MCNP5. The reference solutions for all situations will be provided by MCNP5. The depletion behaviour of any lattice will depend on the power or flux normalization that is considered. In general the flux in various regions is estimated with reference to a single neutron absorbed a

Ramamoorthy, Karthikeyan

303

NSDL National Science Digital Library

This learning community provides a forum for teachers, parents, and students to share ideas about teaching and learning. The concept of this forum is centered around the Common Core Standards to aide in moving mathematics education forward through shared knowledge. The website provides opportunities for community members to contribute, review, and learn through "Educational Objects".

2012-01-01

304

F-layer formation in the outer core with asymmetric inner core growth

NASA Astrophysics Data System (ADS)

Numerical calculations of thermochemical convection in a rotating, electrically conducting fluid sphere with heterogeneous boundary conditions are used to model effects of asymmetric inner core growth. With heterogeneous inner core growth but no melting, outer core flow consists of intense convection where inner core buoyancy release is high, weak convection where inner core buoyancy release is low, and large scale, mostly westward flow in the form of spiraling gyres. With localized inner core melting, outer core flow includes a gravity current of dense fluid that spreads over the inner core boundary, analogous to the seismic F-layer. An analytical model for gravity currents on a sphere connects the structure of the dense layer to the distribution of inner core melting and solidification. Predictions for F-layer formation by asymmetric inner core growth include large-scale asymmetric gyres below the core-mantle boundary and eccentricity of the geomagnetic field.

Deguen, Renaud; Olson, Peter; Reynolds, Evan

2014-05-01

305

Core restraint contributions to radial expansion reactivity

Bowing of core assemblies caused by thermal gradients, swelling gradients, and irradiation creep can cause significant changes in reactivity of an LMFBR during startup, overpower and loss-of-flow without scram transients. This paper summarizes calculations of bowing reactivity effects for both a small homogeneous and a small heterogeneous core design. It includes two core restraint concepts for each core design and concentrates on reactivity changes in the critical power-to-flow range of 1.0 to 2.0.

Moran, T.J.

1986-01-01

306

Accuracy of Monte Carlo Criticality Calculations During BR2 Operation

The Belgian Material Test Reactor BR2 is a strongly heterogeneous high-flux engineering test reactor at SCK-CEN (Centre d'Etude de l'Energie Nucleaire) in Mol with a thermal power of 60 to 100 MW. It deploys highly enriched uranium, water-cooled concentric plate fuel elements, positioned inside a beryllium reflector with a complex hyperboloid arrangement of test holes. The objective of this paper is to validate the MCNP and ORIGEN-S three-dimensional (3-D) model for reactivity predictions of the entire BR2 core during reactor operation. We employ the Monte Carlo code MCNP-4C to evaluate the effective multiplication factor k{sub eff} and 3-D space-dependent specific power distribution. The one-dimensional code ORIGEN-S is used to calculate the isotopic fuel depletion versus burnup and to prepare a database with depleted fuel compositions. The approach taken is to evaluate the 3-D power distribution at each time step and along with the database to evaluate the 3-D isotopic fuel depletion at the next step and to deduce the corresponding shim rod positions of the reactor operation. The capabilities of both codes are fully exploited without constraints on the number of involved isotope depletion chains or an increase of the computational time. The reactor has a complex operation, with important shutdowns between cycles, and its reactivity is strongly influenced by poisons, mainly {sup 3}He and {sup 6}Li from the beryllium reflector, and the burnable absorbers {sup 149}Sm and {sup 10}B in the fresh UAl{sub x} fuel. The computational predictions for the shim rod positions at various restarts are within 0.5 $ ({beta}{sub eff} = 0.0072)

Kalcheva, Silva; Koonen, Edgar; Ponsard, Bernard [SCK-CEN (Belgium)

2005-08-15

307

Air core pulse transformer design

Cylindrical-air-core pulse transformers capable of passing high-voltage\\/high-energy pulse waveforms with high efficiency and low distortion require a much more delicate design balance of physical dimensions and electrical parameters than iron or ferrite core units. Special computer codes were written to evaluate their performance. The analysis includes calculation of the self and mutual inductances as determined by the dimensions and insulation

J. P. O'Loughlin; J. D. Sidler; Gerry J. Rohwein

1988-01-01

308

Core restraint design for inherent safety

A simple analytical model is developed of core radial expansion for a fast reactor using a limited-free-bow core restraint design. Essentially elementary beam theory is used to calculate the elastic bow of a driver assembly at the core periphery subject to temperature dependent boundary conditions at the nozzle support, ACLP and TLP and subject to thermal and inelastic bowing deformations. The model is used to show the relative importance of grid plate temperature, core temperature rise, and restraint ring temperature in the inherent response of a limited-free-bow core restraint system to thermal transients. It is also used to explore this inherent core expansion. Limited verification of the model using detailed 3-D core restraint calculations is presented.

Moran, T.J.

1988-01-01

309

Modeling of fuel rod steady state and transient behavior over the full range of burnup

The heightened recent attention given to fuel rod behavior at high burnup has largely been the result of a few reactivity initiation experiments conducted in France and Japan. Regardless of the merits of these tests, their outcome has underscored the need for improved analytical methods for both steady-state performance evaluation and transient safety analysis. At issue is the ability to reliably predict fuel rod behavior over the full range of burnup under all conditions. The overall process can be divided into two essential components: the collection of high-burnup material properties data and the development of predictive computer codes with essential special-effects models. Lessons learned from the Electric Power Research Institute`s (EPRI`s) recent efforts in understanding and properly interpreting the reactivity insertion accident test results indicate that a first-principles approach to fuel rod behavioral modeling within a robust analytical capability is vital for describing the complex interaction between the various phenomena involved. Thus, the heavy reliance on analysis-by-correlation and the excessive use of adjustable parameters is not adequate, especially in extrapolating the analytical results beyond the correlation database. Guided by this philosophy, EPRI has initiated a major fuel rod modeling effort that builds on two existing EPRI codes: the ESCORE code for steady-state analysis and the FREY code for transient analysis. The new combined code, FALCON, is not a mere merging of the one-dimensional ESCORE and the two-dimensional FREY but rather an innovative construct of robust numerics and relevant material models.

Yagnik, S.K.; Yang, R.L. [Electric Power Research Institute, Palo Alto, NC (United States); Rashid, Y.R.; Montgomery, R.O. [Anatech Research Corp., San Diego, CA (United States)

1997-12-01

310

Matching meal insulin to carbohydrate intake, blood glucose, and activity level is recommended in type 1 diabetes management. Calculating an appropriate insulin bolus size several times per day is, however, challenging and resource demanding. Accordingly, there is a need for bolus calculators to support patients in insulin treatment decisions. Currently, bolus calculators are available integrated in insulin pumps, as stand-alone devices and in the form of software applications that can be downloaded to, for example, smartphones. Functionality and complexity of bolus calculators vary greatly, and the few handfuls of published bolus calculator studies are heterogeneous with regard to study design, intervention, duration, and outcome measures. Furthermore, many factors unrelated to the specific device affect outcomes from bolus calculator use and therefore bolus calculator study comparisons should be conducted cautiously. Despite these reservations, there seems to be increasing evidence that bolus calculators may improve glycemic control and treatment satisfaction in patients who use the devices actively and as intended. PMID:24876436

Schmidt, Signe; Nørgaard, Kirsten

2014-09-01

311

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

312

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

313

Calculation of VVER reactor computational benchmark pin cell variants

In the framework of the joint US/Russian project to update, verify, and validate reactor design/safety computer codes associated with mixed-oxide (MOX) usage in VVER reactors, a computational benchmark was constructed. The calculational results of pin cell variants from the benchmark are given in this paper. The pin cells contain low-enriched uranium (LEU), weapons-grade (W-G) MOX fuel, and reactor MOX fuel. The calculations call for a wide range of temperatures and soluble boron concentrations. Several of the variants include burnup analyses.

Kalashnikov, A.G. [IPPE (Russian Federation); Kalugin, M.A.; Lazarenko, A.P. [Kurchatov RRC (Russian Federation); Gehin, J.C. [Oak Ridge National Lab., TN (United States)

1998-12-31

314

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

315

EBSD and TEM characterization of high burn-up mixed oxide fuel

NASA Astrophysics Data System (ADS)

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 ?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 ?2.5× 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 ?25 ?m 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.

Teague, Melissa; Gorman, Brian; Miller, Brandon; King, Jeffrey

2014-01-01

316

Gravitationally driven inner core differential rotation

NASA Astrophysics Data System (ADS)

A heterogeneous heat flux at the core-mantle boundary can maintain time-averaged convective structures in the fluid core. This includes a steady pattern of heterogeneous heat flux at the inner core boundary which leads to aspherical inner core growth. If this growth pattern is longitudinally misaligned with the mantle-induced geoid, the latter would impart a gravitational torque on the newly created topography of the inner core. Allowing for continuous melting/solidification and viscous deformation of the inner core, a steady gravitationally driven differential rotation of the inner core with respect to the mantle can be sustained. In this work, we present calculations of inner core rotation driven by such a mechanism using recently published models of the heat flux at the inner core boundary and of the geoid at the base of the mantle. We show that, for a fast mean inner core growth rate of 1 mm/yr and a bulk viscous relaxation time longer than 100 yr, the inner core differential rotation can be as high as 100 deg/Myr, in the westward direction. Although this is much too slow (and in the wrong direction) to explain the seismically inferred inner core rotation (eastward, of the order of 0.2 deg/yr), this mechanism by itself would rotate the inner core by one full rotation in only a few million years. This gravitational torque would then partly offset the viscomagnetic torque from the steady eastward zonal flow near the inner core boundary, the driving mechanism typically invoked to explain a steady inner core super-rotation. Under the combined action of these two torques, the overall steady differential rotation of the inner core may then be small. This would allow the development of an inner core texture with a distinct longitudinal pattern connected to its aspherical freezing rate, as has been suggested to explain seismic observations.

Dumberry, Mathieu

2010-09-01

317

Particle physics: Hard-core revelations

Our description of how the atomic nucleus holds together has up to now been entirely empirical. Arduous calculations starting from the theory of the strong nuclear force provide a new way into matter's hard core.

Frank Wilczek

2007-01-01

318

This paper described the procedure of implementation and the possibilities of the BURNUP program. The purpose of the program is to predict the change in the nuclear composition of the materials of which a reactor is made in the course of its run and compute the radiation characteristics of the materials after their irradiation.

Yudkevich, M. S., E-mail: umark@adis.vver.kiae.ru [National Research Centre Kurchatov Institute (Russian Federation)

2012-12-15

319

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

320

EXOTIC-7: Irradiation of ceramic breeder materials to high lithium burnup

NASA Astrophysics Data System (ADS)

The EXOTIC-7 irradiation experiment in the high flux reactor (HFR) has been completed. Its aim has been to investigate the effects of high lithium-burnup on the mechanical stability and tritium release characteristics of candidate ceramic breeder materials, originating from the fusion programmes of CEA, FZK, ENEA, AECL and ECN. The tested ceramic breeder materials were pellets of Li 2ZrO 3, LiAlO 2 and Li 8ZrO 6 and pebbles of Li 4SiO 4 and Li 2ZrO 3, with a variety of characteristics, like grain size and porosity. The test matrix provided the simultaneous irradiation of eight independent capsules with on-line tritium monitoring. Two capsules contained a mixture of Li 4SiO 4 and beryllium pebbles. The experimental design, sample loading and main irradiation parameters are described. Some PIE results and analysis of in-situ tritium release behaviour are presented.

van der Laan, J. G.; Kwast, H.; Stijkel, M.; Conrad, R.; May, R.; Casadio, S.; Roux, N.; Werle, H.; Verrall, R. A.

1996-10-01

321

Comparison of nuclear data uncertainty propagation methodologies for PWR burn-up simulations

Several methodologies using different levels of approximations have been developed for propagating nuclear data uncertainties in nuclear burn-up simulations. Most methods fall into the two broad classes of Monte Carlo approaches, which are exact apart from statistical uncertainties but require additional computation time, and first order perturbation theory approaches, which are efficient for not too large numbers of considered response functions but only applicable for sufficiently small nuclear data uncertainties. Some methods neglect isotopic composition uncertainties induced by the depletion steps of the simulations, others neglect neutron flux uncertainties, and the accuracy of a given approximation is often very hard to quantify. In order to get a better sense of the impact of different approximations, this work aims to compare results obtained based on different approximate methodologies with an exact method, namely the NUDUNA Monte Carlo based approach developed by AREVA GmbH. In addition, the impact ...

Diez, Carlos Javier; Hoefer, Axel; Porsch, Dieter; Cabellos, Oscar

2014-01-01

322

NASA Technical Reports Server (NTRS)

This report gives an overall view of the CORE program at Goddard Space Flight Center (GSFC). It summarizes the different CORE sessions and gives information about the technical staff. The outlook summarizes the evolution of the different CORE programs.

Thomas, Cynthia; Vandenberg, Nancy

1999-01-01

323

Biases for current FFTF calculational methods

Uncertainties in nuclear data and approximate calculational methods used in safety design, and operational support of a reactor yield biased as well as uncertain results. Experimentally based biases for use in Fast Flux Test Facility (FFTF) core calculations have been evaluated and are presented together with a description of calculational methods. Experimental data for these evaluations were obtained from an

P. A. Ombrellaro; R. A. Bennett; J. W. Daughtry; K. D. Dobbin; R. A. Harris; J. V. Nelson; R. E. Peterson; R. B. Rothrock

1978-01-01

324

FIRST PRINCIPLES CALCULATIONS OF TOKAMAK ENERGY TRANSPORT

FIRST PRINCIPLES CALCULATIONS OF TOKAMAK ENERGY TRANSPORT M. KOTSCHENREUTHER, W. DORLAND, Q.P. LIU, California, United States of America Abstract FIRST PRINCIPLES CALCULATIONS OF TOKAMAK ENERGY TRANSPORT equation is presented. Calculations of core energy transport are compared with experimental results from

Hammett, Greg

325

NASA Astrophysics Data System (ADS)

The hemispherical asymmetry seen in the seismological properties of the inner core has recently been interpreted as resulting form a high-viscosity mode of inner core thermal convection, consisting in a translation of the inner core with melting on one hemisphere and solidification on the other. Inner core translation can potentially explain a significant part of the inner core structure, but its existence depends critically on the value of poorly constrained parameters. Being a thermal convection mode, a prerequisite for the existence of convective translation is that a superadiabatic temperature profile is maintained with the inner core. The necessary conditions for inner core superadiabaticity will be discussed, in particular in view of recent work suggesting a core thermal conductivity much larger than previously thought. If the inner core is indeed superadiabatic, linear stability analysis, asymptotic calculations, and direct numerical simulations consistently show that the convective translation mode dominates only if the viscosity is large enough, with a critical viscosity value estimated to be ˜ 3 × 1018 Pa~s. Inner core translation imposes highly asymmetric boundary conditions for the outer core flow, with a negative buoyancy flux in the melting hemisphere and a high positive flux in the crystallizing hemisphere. This might have profound - and potentially observable - implications for the outer core dynamics. The excessively asymmetric buoyancy release imposed by inner core translation would be expected to induce a persisting asymmetry in the outer core flow and the geomagnetic field. Numerical dynamos suggest a hemispherical pattern of buoyancy flux at the ICB would produce a large scale asymmetric anticyclonic jet reminiscent of the eccentric gyre inferred in the core from quasi-geostrophic core flow inversions. In addition, the large melt production associated with inner core translation may be at the origin of the seismic F-layer, with inner core melting providing the dense iron-rich melt necessary for the formation of the layer.

Deguen, R.; Alboussiere, T.; Olson, P.; Cardin, P.

2012-12-01

326

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

327

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

328

In 1985, prior to the completion of its nuclear power plant, Texas Utilities (TU) Electric established a corporate commitment to develop and license a fuel vendor-independent reactor analysis capability. Since TU Electric had no nuclear operating plants at that time, the analytical tools were selected without the benefit of comparisons of calculated results to measured data from TU Electric plants. After selection of the code package, reactor models were developed and bench-marked against measured data from two operating pressurized water reactors (PWRs). These benchmarks, along with the results of pin-cell and small-core analyses, formed the bases of a topical report submitted to the US Nuclear Regulatory Commission (NRC) in mid-1989. Included with that NRC submittal were predictions of Comanche Peak Steam Electric Station (CPSES) low-power physics tests results and other physics parameters routinely measured during power operation. The results of recent CPSES initial start-up tests as well as data obtained during power ascension testing continue to validate the high reliability of the three-dimensional nodal core model developed by TU Electric.

Killgore, M.R.; Janne, R.L.; Husain, A. (Texas Utilities, Dallas (United States))

1990-01-01

329

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.

Foundation, Wgbh E.

2010-12-23

330

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

331

Pseudopotentials for correlated-electron calculations

We describe a semiempirical method for constructing pseudopotentials for use in correlated wave-function calculations which involves using a combination of calculated and experimental quantities. The pseudopotentials are generated from single-valence-electron configurations and satisfy a norm-conservation condition. Core relaxation and core-polarization effects are taken into account. Detailed results for a typical atom with s and p valence electrons (silicon) and a

Y. Lee; P. R. C. Kent; M. D. Towler; R. J. Needs; G. Rajagopal

2000-01-01

332

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.

Barrow, Mandy

2008-01-01

333

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

334

Pressure Vessel Calculations for VVER-440 Reactors

NASA Astrophysics Data System (ADS)

Monte Carlo calculations were performed for a selected cycle of the Paks NPP Unit II to test a computational model. In the model the source term was calculated by the core design code KARATE and the neutron transport calculations were performed by the MCNP. Different forms of the source specification were examined. The calculated results were compared with measurements and in most cases fairly good agreement was found.

Hordósy, G.; Hegyi, Gy.; Keresztúri, A.; Maráczy, Cs.; Temesvári, E.; Vértes, P.; Zsolnay, É.

2003-06-01

335

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

336

A series of ex-reactor heating tests on low burnup U-26wt.%Pu-10wt.%Zr metallic fuel for the PRISM reactor was conducted to evaluate fuel\\/cladding metallurgical interaction and its effect on cladding integrity at elevated temperatures. The reaction between the fuel and cladding caused liquid-phase formation and dissolution of the inner surface of the cladding. The rate of cladding penetration was below the existing

Hanchung Tsai; Yung Y. Liu; Da-Yung Wang; J. M. Kramer

1991-01-01

337

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.

1999-01-01

338

Processing of low-burnup LEU (low enriched uranium) silicide targets

Currently much of the world's supply of /sup 99m/Tc for medical purposes is produced from ZZMo derived from the fissioning of high enriched uranium (HEU). Substitution of low enriched uranium (LEU) silicide fuel for the HEU alloy and aluminide fuels used in current target designs will allow equivalent ZZMo yields with no change in target geometries. In these studies, targets were irradiated to low burnup (10/sup /minus/5/%) to produce fission products and STZNp at concentrations conveniently measured by gamma spectroscopy. Processing was done by dissolution of LEU targets in acid or base followed by alumina column recovery ZZMo. Acid dissolution is more rapid, but precipitation of silica results in loss of ZZMo. Dissolution of U3Si2--Al targets in base requires more processing steps than the current process for UAl/sub x/--Al fuel. A two-step process of first dissolving the 6061Al cladding and fuel meat aluminum, and then dissolving the U3Si2 fuel particles, has the advantage of eliminating the aluminum from further processing. Loss of ZZMo during the aluminum dissolution is attributed to recoil of ZZMo out of the silicide particles during irradiation. A larger particle size would decrease this ZZMo loss. 6 refs., 4 figs., 2 tabs.

Kwok, J.D.; Vandegrift, G.F.; Matos, J.E.

1988-09-01

339

Microchemical study of high-burnup CANDU ® fuel by imaging-XPS

NASA Astrophysics Data System (ADS)

An advanced facility for characterization of highly radioactive materials by Imaging X-ray Photoelectron Spectroscopy (XPS) has been developed at the Chalk River Laboratories (CRL), based upon over a decade of prior experience with a prototype system. Auxiliary electron and ion guns provide additional in situ capabilities for scanning electron microscopy (SEM), scanning Auger microscopy (SAM) and composition depth profiling. The application of this facility to the characterization of irradiated fuel materials will be illustrated with selected results taken from a detailed study of the microchemistry at the fuel-sheath interface in a CANDU fuel element that was irradiated to extended burnup in the NRU (National Research Universal) reactor at CRL. Inside surfaces of the end caps and the welds between the sheath and the end caps as well as the thin-walled Zircaloy-4 sheath were analyzed. The in situ SEM capability was essential for selecting different areas on each sample, such as sheath locations with and without a visible retained CANLUB graphite layer, for XPS analysis. Effective infiltration of segregated fission products, especially cesium, into the graphite was demonstrated by depth profiling. A richer chemistry of segregated fission products was found on the end caps than on the sheath with elevated levels of barium, strontium, tellurium, iodine and cadmium as well as cesium. The results are consistent with current understanding of the primary migration route for fission products to the sheath and also indicate that the CANLUB layer functions as a chemical rather than a physical barrier to segregated fission products.

Do, Than; Irving, Karen G.; Hocking, William H.

2008-12-01

340

Academic Rigor: The Core of the Core

ERIC Educational Resources Information Center

Some educators see the Common Core State Standards as reason for stress, most recognize the positive possibilities associated with them and are willing to make the professional commitment to implementing them so that academic rigor for all students will increase. But business leaders, parents, and the authors of the Common Core are not the only…

Brunner, Judy

2013-01-01

341

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

342

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

343

NSDL National Science Digital Library

This tool lets you calculate the probability that a random variable X is in a specified range, for a variety of probability distributions for X: the normal distribution, the binomial distribution with parameters n and p, the chi-square distribution, the exponential distribution, the geometric distribution, the hypergeometric distribution, the negative binomial distribution, the Poisson distribution, and Student's t-distribution. The first choice box lets you select a probability distribution. Depending on the distribution you select, text areas will appear for you to enter the values of the parameters of the distribution. Parameters that are probabilities (e.g., the chance of success in each trial for a binomial distribution) can be entered either as decimal numbers between 0 and 1, or as percentages. If you enter a probability as a percentage, be sure to include the percent sign (%) after the number.

Stark, Philip B.

2009-01-08

344

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

345

POKE. HTGR Core Thermalhydraulic Analysis Code

POKE solves for the steady-state coolant flow rate and temperature distributions in an HTGR core. Core regions are represented by an equivalent number of coolant channels and flow is partitioned by balancing pressure drop across these coolant channels. Solid temperatures are then calculated using either equivalent conductances or an analytic solution to the unit cell within a standard HTGR fuel block. Output includes both printed results and temperature plot files.

Kapernick, R. [CEGA Corporation, San Diego, CA (United States)

1993-11-01

346

BN-600 full MOX core benchmark analysis.

As a follow-up of the BN-600 hybrid core benchmark, a full MOX core benchmark was performed within the framework of the IAEA co-ordinated research project. Discrepancies between the values of main reactivity coefficients obtained by the participants for the BN-600 full MOX core benchmark appear to be larger than those in the previous hybrid core benchmarks on traditional core configurations. This arises due to uncertainties in the proper modelling of the axial sodium plenum above the core. It was recognized that the sodium density coefficient strongly depends on the core model configuration of interest (hybrid core vs. fully MOX fuelled core with sodium plenum above the core) in conjunction with the calculation method (diffusion vs. transport theory). The effects of the discrepancies revealed between the participants results on the ULOF and UTOP transient behaviours of the BN-600 full MOX core were investigated in simplified transient analyses. Generally the diffusion approximation predicts more benign consequences for the ULOF accident but more hazardous ones for the UTOP accident when compared with the transport theory results. The heterogeneity effect does not have any significant effect on the simulation of the transient. The comparison of the transient analyses results concluded that the fuel Doppler coefficient and the sodium density coefficient are the two most important coefficients in understanding the ULOF transient behaviour. In particular, the uncertainty in evaluating the sodium density coefficient distribution has the largest impact on the description of reactor dynamics. This is because the maximum sodium temperature rise takes place at the top of the core and in the sodium plenum.

Kim, Y. I.; Hill, R. N.; Grimm, K.; Rimpault, G.; Newton, T.; Li, Z. H.; Rineiski, A.; Mohanakrishan, P.; Ishikawa, M.; Lee, K. B.; Danilytchev, A.; Stogov, V.; Nuclear Engineering Division; International Atomic Energy Agency; CEA /Cadarache; SERCO Assurance; China Inst. of Atomic Energy; Forschnungszentrum Karlsruhe; Indira Gandhi Centre for Atomic Research; Japan Nuclear Cycle Development Inst.; Korea Atomic Energy Research Inst.; Inst. of Physics and Power Engineering

2004-01-01

347

An apparatus is described for taking core samples from the sidewall of a borehole in a well, the apparatus comprising: a string of drill pipe; at least one gun housing connected to the downhole end of the drill string; at least one coring bullet radially disposed within the gun housing, the coring bullet arranged for securing formation samples from the

E. A. Jr. Colle; D. N. Jr. Yates; E. F. Brieger

1986-01-01

348

Coring Sample Acquisition Tool

NASA Technical Reports Server (NTRS)

A sample acquisition tool (SAT) has been developed that can be used autonomously to sample drill and capture rock cores. The tool is designed to accommodate core transfer using a sample tube to the IMSAH (integrated Mars sample acquisition and handling) SHEC (sample handling, encapsulation, and containerization) without ever touching the pristine core sample in the transfer process.

Haddad, Nicolas E.; Murray, Saben D.; Walkemeyer, Phillip E.; Badescu, Mircea; Sherrit, Stewart; Bao, Xiaoqi; Kriechbaum, Kristopher L.; Richardson, Megan; Klein, Kerry J.

2012-01-01

349

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

350

23. CORE WORKER OPERATING A COREBLOWER THAT PNEUMATICALLY FILLED CORE ...

23. CORE WORKER OPERATING A CORE-BLOWER THAT PNEUMATICALLY FILLED CORE BOXES WITH RESIGN IMPREGNATED SAND AND CREATED A CORE THAT THEN REQUIRED BAKING, CA. 1950. - Stockham Pipe & Fittings Company, 4000 Tenth Avenue North, Birmingham, Jefferson County, AL

351

Calculation Methods for Core Distortions and Mechanical Behavior.

National Technical Information Service (NTIS)

This paper describes ABADAN, a general purpose, nonlinear, multi-dimensional finite element structural analyses computer code developed for the express purpose of solving large nonlinear problems as typified by the Liquid Metal Fast Breeder Reactor (LMFBR...

W. H. Sutherland

1984-01-01

352

Calculating conditional core damage probabilities for nuclear power plant operations

A part of managing nuclear power plant operations is the control of plant risk over time as components are taken out of service or plant upsets are caused by initiating events. Unfortunately, measuring risk over time proves to be challenging, even with modern probabilistic risk analyses (PRAs) and PRA tools. In general, the process of measuring the operational risk would

Curtis L. Smith

1998-01-01

353

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

354

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

355

A procedure for structuring generally contracted valence-core/valence basis sets of Gaussian-type functions for use with relativistic effective core potentials (gcv-c/v-RECP basis sets) is presented. Large valence basis sets are enhanced using a compact basis set derived for outer core electrons in the presence of small-core RECPs. When core electrons are represented by relativistic effective core potentials (RECPs), and appropriate levels of theory, these basis sets are shown to provide accurate representations of atomic and molecular valence and outer-core electrons. Core/valence polarization and correlation effects can be calculated using these basis sets through standard methods for treating electron correlation. Calculations of energies and spectra for Ru, Os, Ir, In and Cs are reported. Spectroscopic constants for RuO2+, OsO2+, Cs2 and InH are calculated and compared with experiment.

Ermler, Walter V.; Tilson, Jeffrey L.

2012-12-15

356

Dublin Core Metadata Initiative

NSDL National Science Digital Library

Dublin Core metadata has been implemented in several ways, including as HTML metatags and as database elements, as it is used in the Scout Archives (discussed in the June 20, 1997 issue of the Scout Report). The DC elements are title, author, subject, description, publisher, other contributor, date, resource type, format, resource identifier, source, language, relation, coverage, and rights management. Information about the Dublin Core Workshop Series, DC semantics and syntax, working papers, and projects that have implemented Dublin Core metadata can be found at the Dublin Core Metadata homepage.

1995-01-01

357

FLODIS. Thermal Response of FSV HTGR Core

FLODIS was developed to analyze shutdown transients for the Fort St. Vrain high-temperature gas-cooled reactor (HTGR) core. The program is a lumped node representation of the 37 refueling regions in the active core, the side reflector blocks, the gas annulus between the core barrel and the prestressed concrete reactor vessel (PCRV) liner, and the PCRV cooling system. Heat conduction in all three coordinate directions and to the coolant is modeled. The calculation uses the specified operating conditions for the reactor at power to determine appropriate loss coefficients for the variable orifices in each refueling region. Iterative techniques are used to determine the distribution of coolant flow as a function of time during the transient. Both forced and natural convection flows can be calculated. FLODIS can be adapted to other HTGR systems with minor modifications.

Paul, D.D. [Oak Ridge National Lab., TN (United States)

1987-03-01

358

KSI's Cross Insulated Core Transformer Technology

NASA Astrophysics Data System (ADS)

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

2009-08-01

359

Glissile dislocations with transient cores in silicon.

We report an unexpected characteristic of dislocation cores in silicon. Using first-principles calculations, we show that all of the stable core configurations for a nondissociated 60 degrees dislocation are sessile. The only glissile configuration, previously obtained by nucleation from surfaces, surprisingly corresponds to an unstable core. As a result, the 60 degrees dislocation motion is solely driven by stress, with no thermal activation. We predict that this original feature could be relevant in situations for which large stresses occur, such as mechanical deformation at room temperature. Our work also suggests that postmortem observations of stable dislocations could be misleading and that mobile unstable dislocation cores should be taken into account in theoretical investigations. PMID:19792584

Pizzagalli, Laurent; Godet, Julien; Brochard, Sandrine

2009-08-01

360

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

361

Calculation of kinetic spatial weighting factors in power reactors

Ex-core neutron detector kinetic (frequency-dependent) spatial sensitivities (weighting factors) for in-core neutron sources were determined by performing space-dependent, transport and diffusion theory, kinetic detector adjoint calculations in which both source propagation through fission processes and the frequency dependence of the reactivity-to-power transfer function were considered. This study was pursued to overcome the shortcomings of previous calculations of ex-core detector weighting factors for in-core neutron sources using discrete-ordinate shielding or point kernel techniques.

Sweeney, F.J.; Renier, J.P.

1982-01-01

362

ERIC Educational Resources Information Center

Core concepts of kinesiology are the basis of communication about movement that facilitate progression of skill levels. The article defines and exemplifies each of 10 core concepts: range of motion, speed of motion, number of segments, nature of segments, balance, coordination, compactness, extension at release/contact, path of projection, and…

Hudson, Jackie L.

1995-01-01

363

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

364

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

365

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

366

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

367

NASA Technical Reports Server (NTRS)

This report gives a synopsis of the activities of the CORE (continuous observations of the rotation of the Earth) Operating Center from March 1999 to December 2000. The report forecasts activities planned for the year 2001. The outlook summarizes the evolution of the different CORE programs.

Thomas, Cynthia C.

2001-01-01

368

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

369

Measurement of flux distribution in toroidal and multiaperture cores

Recently, magnetic devices utilizing a partial switching have been used in many fields. From a point of core design, a flux distribution must be known. It is possible to calculate them under some assumptions, but it is difficult to measure them experimentally. This paper describes a technique of determining a flux distribution in toroidal and multiaperture ferrite cores. The patterns

Y. Ohbuchi; T. Urabe; Y. Sakurai

1971-01-01

370

Dust Growth in Core-Collapse and Thermonuclear Supernovae

NASA Astrophysics Data System (ADS)

We here present calculations of dust condensation in core-collapse (Type II) and thermonuclear (Type Ia) supernovae, to understand the production of large dust grains in Type II supernovae and small dust grains in thermonuclear supernovae.

Yu, T.; Meyer, B. S.; Clayton, D. D.

2013-09-01

371

Utilization of Mixed Uranium–Plutonium Oxide (MOX) fuel in VVER-1000 reactors envisages the core physics analysis using computational methods and validation of the related computer codes. Towards this objective, an international experts group has been established at OECD\\/NEA. The experts group facilitates sharing of existing information on physics parameters and fuel behaviour. Several benchmark exercises have been proposed by them with

L. Thilagam; C. Sunil Sunny; V. Jagannathan; K. V. Subbaiah

2009-01-01

372

Collapse and fragmentation of molecular cloud cores. I - Moderately centrally condensed cores

NASA Technical Reports Server (NTRS)

3D calculations of the collapse of moderately centrally condensed molecular cloud cores with varied thermal and rotational energies are presented. The calculations are carried out using a newly developed and tested second-order accurate radiative hydrodynamics code. Because of the use of a second-order accurate numerical scheme and initial clouds that resemble both observed prolate molecular cloud cores and magnetically supported clouds at the initiation of the dynamic collapse phase, the new models provide a superior estimate of the likelihood of fragmentation as a mechanism for binary star formation.

Boss, Alan P.

1993-01-01

373

NASA Astrophysics Data System (ADS)

In this thesis I re-analyse the SCUBA archive data for the L1688 main cloud of Ophiuchus, incorporating all available scan-map and jiggle-map data. I create a new core mass function (CMF) for L1688 using updated values for the distance to this region, as well as new estimates for the temperatures and masses of the cores. I show that the CMF for LI688 is consistent with a three part power-law, with slopes the same as those seen in the stellar IMF. The deeper maps allows the discovery of a turnover in the CMF at 0.7A/, which shows that the core mass function appears to mimic the stellar initial mass function. This concordance is indicative that the stellar IMF is determined at the prestellar core phase. I also present HCO* (J=4 > 3) spectral line observations from HARP on the JCMT. Data are presented for 59 of the prestellar cores mapped using SCUBA. Using these data. I present a proposed evolutionary diagram for prestellar cores in the form of a radius-mass plot. I hypothesise that a core is formed in the low-mass, low-radius region of the plot. It then accretes quasi-statically, increasing in both mass and radius. When it crosses the limit of gravitational instability it begins to collapse, decreasing in radius, towards the region of the diagram where protostellar cores are seen. My predictions are borne out when I plot the collapsing cores on this diagram. I outline an analytical model, created by Whitworth & Ward-Thompson (2001), describing the collapse of a starless core with a Plummer-like density profile. I describe my addition of a simple radiative transfer code, which allows simulated spectral line profiles to be created for such cores. The model is shown to be consistent with previous models, and with the observed physical properties of prestellar cores. This model is applied to 20 of the spectral line profiles from the HCO+ (J=4 + 3) data. These 20 modelled cores are placed onto the proposed evolutionary diagram. Their modelled physical states are found to be consistent with the proposed evolutionary track. In conjunction with the SCUBA and HARP data, these fits allow a potential timeline for the LI688 cloud to be established for the first time.

Simpson, Robert J.

374

The dynamic behavior of partially delaminated at the skin\\/core interface sandwich plates with flexible cores is studied. The commercial finite element code ABAQUS is used to calculate natural frequencies and mode shapes of the sandwich plates containing a debonding zone. The influence of the debonding size, debonding location and types of debonding on the modal parameters of damaged sandwich plates

Vyacheslav N. Burlayenko; Tomasz Sadowski

2010-01-01

375

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

376

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

377

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

378

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

379

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

380

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

381

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

382

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

383

On-Site was developed to provide modelers and model reviewers with prepackaged tools ("calculators") for performing site assessment calculations. The philosophy behind OnSite is that the convenience of the prepackaged calculators helps provide consistency for simple calculations,...

384

Experimental constraints on Mercury's core composition

NASA Astrophysics Data System (ADS)

The recent discovery of high S concentrations on the surface of Mercury by spacecraft measurements from the MESSENGER mission provides the potential to place new constraints on the composition of Mercury's large metallic core. In this work, we conducted a set of systematic equilibrium metal-silicate experiments that determined the effect of different metallic compositions in the Fe-S-Si system on the S concentration in the coexisting silicate melt. We find that metallic melts with a range of S and Si combinations can be in equilibrium with silicate melts with S contents consistent with Mercury's surface, but that such silicate melts contain Fe contents lower than measured for Mercury's surface. If Mercury's surface S abundance is representative of the planet's bulk silicate composition and if the planet experienced metal-silicate equilibrium during planetary core formation, then these results place boundaries on the range of possible combinations of Si and S that could be present as the light elements in Mercury's core and suggest that Mercury's core likely contains Si. Except for core compositions with extreme abundances of Si, bulk Mercury compositions calculated by using the newly determined range of potential S and Si core compositions do not resemble primitive meteorite compositions.

Chabot, Nancy L.; Wollack, E. Alex; Klima, Rachel L.; Minitti, Michelle E.

2014-03-01

385

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

386

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

387

National Technical Information Service (NTIS)

The non-destructive gamma-spectrometric method (HRGS) and the passive neutron technique (PNT) were applied to the determination of WWER 440 reactor spent fuel assemblies burn-up for safeguard purposes. Rapid codes FISPR-2 and BUNECO were compiled on HP-85...

S. Rohar, P. Liptak, L. Krajci, V. Petenyi, R. Arlt

1988-01-01

388

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

389

Biopolymers & Proteomics Core Facility

Biopolymers & Proteomics Core Facility 76-185 x4-2609 Proteomics Sample Form (one form per sample information, previously acquired mass spectral data, etc.) Proteomics Division Use Only DATES RECEIVED

Sabatini, David M.

390

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.

391

Central core disease (CCD) is an inherited neuromuscular disorder characterised by central cores on muscle biopsy and clinical\\u000a features of a congenital myopathy. Prevalence is unknown but the condition is probably more common than other congenital myopathies.\\u000a CCD typically presents in infancy with hypotonia and motor developmental delay and is characterized by predominantly proximal\\u000a weakness pronounced in the hip girdle;

Heinz Jungbluth

2007-01-01

392

A series of ex-reactor heating tests on low burnup U-26wt.%Pu-10wt.%Zr metallic fuel for the PRISM reactor was conducted to evaluate fuel/cladding metallurgical interaction and its effect on cladding integrity at elevated temperatures. The reaction between the fuel and cladding caused liquid-phase formation and dissolution of the inner surface of the cladding. The rate of cladding penetration was below the existing design correlation, which provides a conservative margin to cladding failure. In a test which enveloped a wide range of postulated reactor transient events, a substantial temporal cladding integrity margin was demonstrated for an intact, whole fuel pin. The cause of the eventual pin breach was reaction-induced cladding thinning combined with fission-gas pressure loading. The behavior of the breached pin was benign. 7 refs., 7 figs., 1 tab.

Tsai, Hanchung; Liu, Yung Y.; Wang, Da-Yung; Kramer, J.M.

1991-07-01

393

MOX capsule post-irradiation examination. Volume 2: Test plan for 30-GWd/MT burnup fuel

This test plan is a Level-2 document as defined in the Fissile Materials Disposition Program Light-Water Reactor Mixed-Oxide Fuel Irradiation Test Project Plan. The planned post-irradiation examination (PIE) work to be performed on the mixed uranium and plutonium oxide fuel capsules that have received burnups of approximately 30 GWd/MT is described. The major emphasis of this PIE task will be material interactions, particularly indications of gallium transport and interactions. This PIE will include gamma scanning, ceramography, metallography, pellet radial gallium analysis, and clad gallium analysis. A preliminary PIE report will be generated before all the work is completed so that the progress of the fuel irradiation may be known in a timely manner.

Morris, R.N.

1997-12-01

394

NASA Astrophysics Data System (ADS)

The radial distribution of grain boundary gas in a PWR and a BWR fuel is reported. The measurements were made using a new approach involving X-ray fluorescence analysis and electron probe microanalysis. In both fuels the concentration of grain boundary gas was much higher than hitherto suspected. The gas was mainly contained in the bubble/pore structure. The factors that determined the fraction of gas released from the grains and the level of gas retention on the grain boundaries are identified and discussed. The variables involved are the local fuel stoichiometry, the amount of open porosity, the magnitude of the local compressive hydrostatic stress and the interaction of metallic precipitates with gas bubbles on the grain faces. It is concluded that under transient conditions the interlinkage of gas bubbles on the grain faces and the subsequent formation of grain edge tunnels is the rate determining step for gas release; at least when high burn-up fuel is involved.

Walker, C. T.; Mogensen, M.

1987-07-01

395

Tidal flow and dissipation in planetary cores

NASA Astrophysics Data System (ADS)

Tidal forcing at diurnal periods drives a flow in the liquid cores of planets, altering the nutation of these bodies. An approximate description of the flow in an inviscid fluid is comprised of a uniform vorticity and a potential flow to satisfy boundary conditions. This flow (often called Poincare flow) is unable to satisfy boundary conditions when the solid inner core tilts relative to the mantle or when the ellipticity of the inner-core boundary differs from the ellipticity of the core-mantle boundary. A numerical solution is used to show that internal shear layers develop when Poincare flow fails. The thickness of the shear layers scale as E1/3, where E is the Ekman number. The total viscous dissipation also scales as E1/3 when stress-free boundary conditions are assumed. Extrapolating E to realistic values for planetary cores suggests that viscous dissipation is negligible. However, shear layers also induce electric currents when the core is permeated with a magnetic field. Scaling arguments for a weak magnetic field suggest that the resulting ohmic dissipation is proportional to E-2/3. This estimate is supported by numerical calculations with E = 10-4 to 10-6. The singularity at vanishingly small values of E is avoided by the influence of the Lorentz force on the flow. The resulting ohmic dissipation should be detectable in observations of the Earth’s nutation, but it is insufficient to explain all of the dissipation that is actually observed. Strong magnetic coupling at the core-mantle and inner-core boundaries probably accounts for most of the observed dissipation.

Buffett, B. A.

2009-12-01

396

... Subject: Initial Environmental Evaluation (Deep Ice Core Drilling at Vostok Station, Antarctica ... for the National Science Foundation's Deep Ice Core Drilling project at Vostok Station, prepare an ...

397

Calculations of thermal-reactor spent-fuel nuclide inventories and comparisons with measurements

Comparisons with integral measurements have demonstrated the accuracy of CINDER codes and libraries in calculating aggregate fission-product properties, including neutron absorption, decay power, and decay spectra. CINDER calculations have, alternatively, been used to supplement measured integral data describing fission-product decay power and decay spectra. Because of the incorporation of the extensive actinide library and the use of ENDF/B-V data, it is desirable to compare the inventory of individual nuclides obtained from tandem EPRI-CELL/CINDER-2 calculations with those determined in documented benchmark inventory measurements of spent reactor fuel. The development of the popular /sup 148/Nd burnup measurement procedure is outlined, and areas of uncertainty in it and lack of clarity in its interpretation are indicated. Six inventory samples of varying quality and completeness are examined. The power histories used in the calculations have been listed for other users.

Wilson, W.B.; LaBauve, R.J.; England, T.R.

1982-01-01

398

NASA Technical Reports Server (NTRS)

This presentation is a technical progress report and near-term outlook for NASA-internal and NASA-sponsored external work on core (combustor and turbine) noise funded by the Fundamental Aeronautics Program Subsonic Fixed Wing (SFW) Project. Sections of the presentation cover: the SFW system level noise metrics for the 2015, 2020, and 2025 timeframes; the emerging importance of core noise and its relevance to the SFW Reduced-Noise-Aircraft Technical Challenge; the current research activities in the core-noise area, with some additional details given about the development of a high-fidelity combustion-noise prediction capability; the need for a core-noise diagnostic capability to generate benchmark data for validation of both high-fidelity work and improved models, as well as testing of future noise-reduction technologies; relevant existing core-noise tests using real engines and auxiliary power units; and examples of possible scenarios for a future diagnostic facility. The NASA Fundamental Aeronautics Program has the principal objective of overcoming today's national challenges in air transportation. The SFW Reduced-Noise-Aircraft Technical Challenge aims to enable concepts and technologies to dramatically reduce the perceived aircraft noise outside of airport boundaries. This reduction of aircraft noise is critical for enabling the anticipated large increase in future air traffic. Noise generated in the jet engine core, by sources such as the compressor, combustor, and turbine, can be a significant contribution to the overall noise signature at low-power conditions, typical of approach flight. At high engine power during takeoff, jet and fan noise have traditionally dominated over core noise. However, current design trends and expected technological advances in engine-cycle design as well as noise-reduction methods are likely to reduce non-core noise even at engine-power points higher than approach. In addition, future low-emission combustor designs could increase the combustion-noise component. The trend towards high-power-density cores also means that the noise generated in the low-pressure turbine will likely increase. Consequently, the combined result from these emerging changes will be to elevate the overall importance of turbomachinery core noise, which will need to be addressed in order to meet future noise goals.

Hultgren, Lennart S.

2010-01-01

399

CORRELATING INFALL WITH DEUTERIUM FRACTIONATION IN DENSE CORES

We present a survey of HCO{sup +} (3-2) observations pointed toward dense cores with previous measurements of N(N{sub 2}D{sup +})/N(N{sub 2}H{sup +}). Of the 26 cores in this survey, 5 show the spectroscopic signature of outward motion, 9 exhibit neither inward nor outward motion, 11 appear to be infalling, and 1 is not detected. We compare the degree of deuterium fractionation with infall velocities calculated from the HCO{sup +} spectra and find that those cores with [D]/[H] > 0.1 are more likely to have the signature of inward motions than cores with smaller [D]/[H] ratios. Infall motions are also much more common in cores with masses exceeding their thermal Jeans masses. The fastest infall velocity measured belongs to one of the two protostellar cores in our survey, L1521F, and the observed motions are typically on the order of the sound speed.

Schnee, Scott; Brunetti, Nathan; Friesen, Rachel [National Radio Astronomy Observatory, 520 Edgemont Road, Charlottesville, VA 22903 (United States); Di Francesco, James; Johnstone, Doug; Pon, Andy [National Research Council Canada, Herzberg Institute of Astrophysics, 5071 West Saanich Road Victoria, BC V9E 2E7 (Canada); Caselli, Paola, E-mail: sschnee@nrao.edu [School of Physics and Astronomy, University of Leeds, Leeds LS2 9JT (United Kingdom)

2013-11-10

400

Core Noise - Increasing Importance

NASA Technical Reports Server (NTRS)

This presentation is a technical summary of and outlook for NASA-internal and NASA-sponsored external research on core (combustor and turbine) noise funded by the Fundamental Aeronautics Program Subsonic Fixed Wing (SFW) Project. Sections of the presentation cover: the SFW system-level noise metrics for the 2015, 2020, and 2025 timeframes; turbofan design trends and their aeroacoustic implications; the emerging importance of core noise and its relevance to the SFW Reduced-Perceived-Noise Technical Challenge; and the current research activities in the core-noise area, with additional details given about the development of a high-fidelity combustor-noise prediction capability as well as activities supporting the development of improved reduced-order, physics-based models for combustor-noise prediction. The need for benchmark data for validation of high-fidelity and modeling work and the value of a potential future diagnostic facility for testing of core-noise-reduction concepts are indicated. The NASA Fundamental Aeronautics Program has the principal objective of overcoming today's national challenges in air transportation. The SFW Reduced-Perceived-Noise Technical Challenge aims to develop concepts and technologies to dramatically reduce the perceived aircraft noise outside of airport boundaries. This reduction of aircraft noise is critical to enabling the anticipated large increase in future air traffic. Noise generated in the jet engine core, by sources such as the compressor, combustor, and turbine, can be a significant contribution to the overall noise signature at low-power conditions, typical of approach flight. At high engine power during takeoff, jet and fan noise have traditionally dominated over core noise. However, current design trends and expected technological advances in engine-cycle design as well as noise-reduction methods are likely to reduce non-core noise even at engine-power points higher than approach. In addition, future low-emission combustor designs could increase the combustion-noise component. The trend towards high-power-density cores also means that the noise generated in the low-pressure turbine will likely increase. Consequently, the combined result from these emerging changes will be to elevate the overall importance of turbomachinery core noise, which will need to be addressed in order to meet future noise goals.

Hultgren, Lennart S.

2011-01-01

401

Vintage Calculators Web Museum

NSDL National Science Digital Library

This "web museum" devoted to vintage calculators shows "the evolution from mechanical calculator to hand held electronic calculator." Some items featured include: Mechanical and early electronic desk calculators, "strange hand-held calculators," and articles, photographs, and databases from the archives of the International Association of Calculator Collectors. A history of the technology and information on British and sterling currency calculators are also posted here. The website also offers a Calculator time-line (chronicling calculator developments), background on the technology used by mechanical and early electronic calculators, and information on The Calculator Business. An index allows visitors to search the calculators featured on this site. The Puzzle Corner section asks visitors to contact them with any information that may answer unresolved questions regarding vintage calculators.

Tout, Nigel

402

Solubility of iron in metallic hydrogen and stability of dense cores in giant planets

Solubility of iron in metallic hydrogen and stability of dense cores in giant planets Sean M. Wahl. In this study, we perform ab initio calculations to study the solubility of an innermost metallic core. We find. We compare with and summarize the results for solubilities on other probable core constituents

Militzer, Burkhard

403

Energy-Aware Application Scheduling on a Heterogeneous Multi-core System

Energy-Aware Application Scheduling on a Heterogeneous Multi-core System Jian Chen and Lizy K. John to matching cores that can deliver the most efficient program execution. This paper presents an energy-aware scheduling mechanism that employs fuzzy logic to calculate the suitability between programs and cores

John, Lizy Kurian

404

NSDL National Science Digital Library

The Arts at the Core Initiative is part of The College Board's Advocacy & Policy Center, created "to help transform education in America." Part of the Center's work involves the Arts at the Core project, whose goal is "to empower education leaders, particularly in under-resourced districts, to implement rigorous arts programming in their schools." Under the Our Progress section, visitors learn about some of the resources created to achieve this goal. Moving on, the News & Events area contains links to recent success stories of bringing arts education programs to schools, along with updates from the field of research into this area. Visitors shouldn't miss the Publications area, which contains a brochure about flagship programs and a summary of key recommendations for school systems seeking to move arts to the core of their mission.

405

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

The response of fuel in a boiling water reactor to the rod drop accident (RDA) was studied using the RAMONA-4B computer code. In this study, a fit of RAMONA-4B bundle powers was used to estimate the local power peaking. It was determined that the peaking factor could be 25% higher than the factor usually assumed for RDA analysis. Combining this error with the 2 sigma random error means that for this analysis the actual fuel rod enthalpy could be 100% larger than calculated by RAMONA-4B. This is much larger than the uncertainty in most parameters that are calculated with best-estimate methods for other design-basis events.

Diamond, D.J.; Neymotin, L.

1997-02-01

406

Analysis of fresh fuel critical experiments appropriate for burnup credit validation

The ANS/ANS-8.1 standard requires that calculational methods used in determining criticality safety limits for applications outside reactors be validated by comparison with appropriate critical experiments. This report provides a detailed description of 34 fresh fuel critical experiments and their analyses using the SCALE-4.2 code system and the 27-group ENDF/B-IV cross-section library. The 34 critical experiments were selected based on geometry, material, and neutron interaction characteristics that are applicable to a transportation cask loaded with pressurized-water-reactor spent fuel. These 34 experiments are a representative subset of a much larger data base of low-enriched uranium and mixed-oxide critical experiments. A statistical approach is described and used to obtain an estimate of the bias and uncertainty in the calculational methods and to predict a confidence limit for a calculated neutron multiplication factor. The SCALE-4.2 results for a superset of approximately 100 criticals are included in uncertainty analyses, but descriptions of the individual criticals are not included.

DeHart, M.D.; Bowman, S.M.

1995-10-01

407

NSDL National Science Digital Library

Students access the ice core data archived at Lamont-Doherty Geological Observatory. They select a core (Greenland, Antarctica, Quelcaya), pose a working hypothesis regarding the data, import the data in an Excel-readable format, and examine the data to determine correlations between variables and cause/effect as recorded in leads and lags. They generate a written and graphical analysis of the data and, in the next lab period, discuss the similarities and differences among their group outputs in terms of demonstrated correlations, assumptions required, effects of latitude, and any other item that arises.

Locke, William

408

Advanced Core Design And Fuel Management For Pebble-Bed Reactors

A method for designing and optimizing recirculating pebble-bed reactor cores is presented. At the heart of the method is a new reactor physics computer code, PEBBED, which accurately and efficiently computes the neutronic and material properties of the asymptotic (equilibrium) fuel cycle. This core state is shown to be unique for a given core geometry, power level, discharge burnup, and fuel circulation policy. Fuel circulation in the pebble-bed can be described in terms of a few well?defined parameters and expressed as a recirculation matrix. The implementation of a few heat?transfer relations suitable for high-temperature gas-cooled reactors allows for the rapid estimation of thermal properties critical for safe operation. Thus, modeling and design optimization of a given pebble-bed core can be performed quickly and efficiently via the manipulation of a limited number key parameters. Automation of the optimization process is achieved by manipulation of these parameters using a genetic algorithm. The end result is an economical, passively safe, proliferation-resistant nuclear power plant.

Hans D. Gougar; Abderrafi M. Ougouag; William K. Terry

2004-10-01

409

Core radii and common-envelope evolution

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 $\\alpha$. 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 ra...

Hall, Philip D

2014-01-01

410

Core-induced chaos in intrashell transitions

Intrashell transitions in Rydberg atoms with a nonhydrogenic core were studied. The degeneracy of the energy shell was lifted by constant electric and magnetic fields and transitions from initial coherent elliptic states (CES) were driven by a harmonic electric field. Evenly spaced hydrogenic resonances dominate at low eccentricities e of the CES. At high e, the hydrogenic structures disappear and new structures of a more irregular nature emerge. Classical and quantal calculations are described and are shown to agree fairly well with the experiment. Further, the classical calculations suggest that many of the observed phenomena are a consequence of the underlying classical dynamics being chaotic.

Fregenal, D.; Henningsen, B.; Horsdal, E. [Department of Physics and Astronomy, Aarhus University, DK-8000 Aarhus C (Denmark); Foerre, M. [Department of Physics and Technology, University of Bergen, N-5007 Bergen (Norway); Richards, D. [Faculty of Mathematics, Open University, Milton Keynes MK7 6AA (United Kingdom)

2007-11-15

411

Core-induced chaos in intrashell transitions

NASA Astrophysics Data System (ADS)

Intrashell transitions in Rydberg atoms with a nonhydrogenic core were studied. The degeneracy of the energy shell was lifted by constant electric and magnetic fields and transitions from initial coherent elliptic states (CES) were driven by a harmonic electric field. Evenly spaced hydrogenic resonances dominate at low eccentricities e of the CES. At high e , the hydrogenic structures disappear and new structures of a more irregular nature emerge. Classical and quantal calculations are described and are shown to agree fairly well with the experiment. Further, the classical calculations suggest that many of the observed phenomena are a consequence of the underlying classical dynamics being chaotic.

Fregenal, D.; Førre, M.; Henningsen, B.; Horsdal, E.; Richards, D.

2007-11-01

412

Core-valence correlations for atoms with open shells

We present an efficient method of inclusion of the core-valence correlations into the configuration interaction (CI) calculations. These correlations take place in the core area where the potential of external electrons is approximately constant. A constant potential does not change the core electron wave functions and Green's functions. Therefore, all operators describing interaction of $M$ valence electrons and $N-M$ core electrons (the core part of the Hartree-Fock Hamiltonian $V^{N-M}$, the correlation potential $\\hat\\Sigma_1({\\bf r},{\\bf r'},E)$ and the screening of interaction between valence electrons by the core electrons $\\hat\\Sigma_2$) may be calculated with all $M$ valence electrons removed. This allows one to avoid subtraction diagrams which make accurate inclusion of the core-valence correlations for $M>2$ prohibitively complicated. Then the CI Hamiltonian for $M$ valence electrons is calculated using orbitals in complete $V^{N}$ potential (the mean field produced by all electrons); $\\hat\\Sigma_1...

Dzuba, V A

2007-01-01

413

Core-valence correlations for atoms with open shells

We present an efficient method of inclusion of the core-valence correlations into the configuration interaction (CI) calculations. These correlations take place in the core area where the potential of external electrons is approximately constant. A constant potential does not change the core electron wave functions and Green's functions. Therefore, all operators describing interaction of M valence electrons and N-M core electrons [the core part of the Hartree-Fock Hamiltonian V{sup N-M}, the correlation potential {sigma}{sub 1}(r,r{sup '},E), and the screening of interaction between valence electrons by the core electrons {sigma}{sub 2}] may be calculated with all M valence electrons removed. This allows one to avoid subtraction diagrams which make accurate inclusion of the core-valence correlations for M>2 prohibitively complicated. Then the CI Hamiltonian for M valence electrons is calculated using orbitals in complete V{sup N} potential (the mean field produced by all electrons); {sigma}{sub 1}+{sigma}{sub 2} are added to the CI Hamiltonian to account for the core-valence correlations. We calculate {sigma}{sub 1} and {sigma}{sub 2} using many-body perturbation theory in which dominating classes of diagrams are included in all orders. We use neutral Xe I and all positive ions up to Xe VIII as a testing ground. We found that the core electron density for all these systems is practically the same. Therefore, we use the same {sigma}{sub 1} and {sigma}{sub 2} to build the CI Hamiltonian in all these systems (M=1,2,3,4,5,6,7,8). Good agreement with experiment for energy levels and Lande factors is demonstrated for all cases from Xe I to Xe VIII.

Dzuba, V. A.; Flambaum, V. V. [School of Physics, University of New South Wales, Sydney 2052 (Australia)

2007-05-15

414

Theoretical Calculations of the Characteristics of Precious Metal Clusters

NASA Astrophysics Data System (ADS)

It is reported that core/shell type Pd/Pt bimetallic nanoclusters where the inner atoms of a Pd cluster are substituted by Pt atoms have extremely enhanced catalytic activity for cyclooctadiene hydrogenation. In order to discuss the electronic states of core/shell clusters, DFT calculations were carried out for Au13, Pd13, Pt13, Pt/Pd12, Pd/Pt12, Pd/Au12, Pd38, and Pd6/Pt32 clusters. From these calculations, it was found that the charge transfer between the core atoms and the shell atoms played an important role in the modification of the electronic state of the surface atoms.

Okumura, Mitsutaka; Kitagawa, Yasutaka; Kawakami, Takashi; Haruta, Masatake; Yamaguchi, Kizashi

2008-09-01

415

Assembly Based Modular Ray Tracing and CMFD Acceleration for BWR Cores with Different Fuel Lattices

The geometry module of the DeCART direct whole core calculation code has been extended in order to analyze BWR cores which might have a mixed loading of different fuel types. First, an assembly based modular ray tracing scheme was implemented for the Method of Characteristic (MOC) calculation, and a CMFD formulation applicable for unaligned mesh conditions was then developed for

J. W. Thomas; Y. Xu; M. Kalugin; B. Kochunas; T. J. Downar; H. G. Joo

2006-01-01

416

Core Composition and the Magnetic Field of Mercury

NASA Astrophysics Data System (ADS)

The density of Mercury suggests a core of approximately 1800 km radius and a mantle of approximately 600 km thickness. Convection in the mantle is often claimed to be capable of freezing the core over the lifetime of the solar system if the core is nearly pure iron. The thermal history calculations of Stevenson et al. (1983) and Schubert et al. (1988) suggest that about 5 weight-% sulphur are required to lower the core liquidus sufficiently to prevent complete freezing of the core and maintain a significant fluid outer core shell. Other candidates for a light alloying element require similarly large concentrations. The requirement of a significant concentration of volatile elements in the core is likely to be at variance with cosmochemical arguments for a mostly refractory, volatile poor composition of the planet. We have re-addressed the question of the freezing of Mercury's core using parameterized convection models based on the stagnant lid theory of planetary mantle convection. We have compared these results to earlier calculations (Conzelmann and Spohn, 1999) of Hermian mantle convection using a finite-amplitude convection code. We find consistently that the stagnant lid tends to thermally insulate the deep interior and we find mantle and core temperatures significantly larger than those calculated by Stevenson et al. (1983) and Schubert et al. (1988). As a consequence we find fluid outer core shells for reasonable mantle rheology parameters even for compositions with as little as 0.1 weight-% sulphur. Stevenson, D.J., T. Spohn, and G. Schubert. Icarus, 54, 466, 1983. Schubert, G. M.N. Ross, D.J. Stevenson, and T. Spohn, in Mercury, F. Vilas, C.R. Chapman and M.S. Matthews, eds., p.429, 1988. Conzelmann, V. and T. Spohn, Bull. Am. Astr. Soc., 31, 1102, 1999.

Spohn, T.; Breuer, D.

2005-05-01

417

NSDL National Science Digital Library

This site explains how core samples are taken from the ocean floor. Topics include how research cruises are planned, who makes up the crew of a research vessel, and what a cruise track is. Links to additional information are embedded in the text.

418

NSDL National Science Digital Library

This Ocean and Climate Change Institute module features a brief, but image-rich overview of ocean drilling and sediment analysis to determine paleoclimate (past climate). This site is the first of a 3-page module, the other two sites (Describing the Core; Sampling Techniques) are linked at the top of the article.

Woods Hole Oceanographic Institution; Ocean and Climate Change Institute

419

ERIC Educational Resources Information Center

This document consists of four papers presented at a symposium on core directions in human resource development (HRD) moderated by Verna Willis at the 1996 conference of the Academy of Human Resource Development. "Reengineering the Organizational HRD Function: Two Case Studies" (Neal Chalofsky) reports an action research study in which the…

1996

420

Core competence (knowledge) (skill)

Core competence 8 5~8 2 3 4 5 6 7 8 PPS003 Ver. 1.1 2011/03/07 #12; 2 (knowledge) (skill) (attitude) Set of skill, knowledge or attitude which should be learned or acquired by each, 2000) (knowledge) (skill) (attitude) Set of skill, knowledge or attitude which should be learned

Wu, Yih-Min

421

ERIC Educational Resources Information Center

The core curriculum accompanied the development of the academic discipline with multiple names such as Kinesiology, Exercise and Sport Science, and Health and Human Performance. It provides commonalties for undergraduate majors. It is timely to renew this curriculum. Renewal involves strategic reappraisals. It may stimulate change or reaffirm the…

Lawson, Hal A.

2007-01-01

422

ERIC Educational Resources Information Center

The nature of the earth's core is described. Indirect evidence (such as that determined from seismological data) indicates that it is an iron alloy, solid toward its center but otherwise liquid. Evidence also suggests that it is the turbulent flow of the liquid that generates the earth's magnetic field. (JN)

Jeanloz, Raymond

1983-01-01

423

NSDL National Science Digital Library

Students learn about one method used in environmental site assessments. They practice soil sampling by creating soil cores, studying soil profiles and characterizing soil profiles in borehole logs. They use their analysis to make predictions about what is going on in the soil and what it might mean to an engineer developing the area.

Integrated Teaching And Learning Program

424

ERIC Educational Resources Information Center

This article presents a debate over the Common Core State Standards Initiative as it has rocketed to the forefront of education policy discussions around the country. The author contends that there is value in having clear cross state standards that will clarify the new online and blended learning that the growing use of technology has provided…

McShane, Michael Q.

2014-01-01

425

ERIC Educational Resources Information Center

When educators think about what makes learning relevant to students, often they narrow their thinking to electives or career technical education. While these provide powerful opportunities for students to make relevant connections to their learning, they can also create authentic experiences in the core curriculum. In the San Juan Unified School…

Kukral, Nicole; Spector, Stacy

2012-01-01

426

Electromagnetic pump stator core

A stator core for supporting an electrical coil includes a plurality of groups of circumferentially abutting flat laminations which collectively form a bore and perimeter. A plurality of wedges are interposed between the groups, with each wedge having an inner edge and a thicker outer edge. The wedge outer edges abut adjacent ones of the groups to provide a continuous path around the perimeter.

Fanning, Alan W. (San Jose, CA); Olich, Eugene E. (Aptos, CA); Dahl, Leslie R. (Livermore, CA)

1995-01-01

427

Charge The QEP Core Committee creates, writes, and coordinates the implementation of the Quality Enhancement Plan. It is the Committee's responsibility to ensure that the plan is technically responsive to requirements outlined in the Quality Enhancement Plan section of the DRAFT Handbook for Institutional Analysis by the Southern Association of Colleges and Schools, Commission on Colleges, 2003; to keep the

Suzanne Morales; John Frith; Terri Jackson

428

ERIC Educational Resources Information Center

This author contends that the United States neglects creativity in its education system. To see this, he states, one may look at the Common Core State Standards. If one searches the English Language Arts and Literacy standards for the words "creative," "innovative," and "original"--and any associated terms, one will find scant mention of the words…

Ohler, Jason

2013-01-01

429

Core Security Requirements Artefacts

Although security requirements engineering has recently attracted increasing attention, it has lacked a context in which to operate. A number of papers have described how security requirements may be violated, but apart from a few hints in the general literature, none have described satisfactorily what security requirements are. This paper proposes a framework of core security requirements artefacts, which unifies

Jonathan D Moffett Charles; B Haley; Bashar Nuseibeh

2004-01-01

430

NASA Astrophysics Data System (ADS)

In this paper, a method is developed to calculate the core loss in a switched reluctance machine. The magnetic circuit of the motor is described as a magnetic network. The electromagnetic behavior of each magnetic network element takes into account the iron loss using the Preisach model and the principle of loss separation. Using the numerical routines, the local core loss in the different motor sections is calculated. The global core loss is compared with the experimentally determined core loss.

Dupré, Luc; Sergeant, Peter; Vandenbossche, Lode

2005-05-01

431

Calculations of radiation levels during reactor operations for safeguard inspections

NASA Astrophysics Data System (ADS)

It is necessary to calculate the total level of radioactivity produced by the reactor operation when the reactor cover is opened during periodical safeguard inspections for safety requirements. Also the calculations are performed for different reactor core and spent fuel storage loadings during refuelling after reactor shutdown. When an internal core spent fuel storage is used in the shield tank to accommodate a large number of spent fuel baskets, physical calculations are performed to determine the number of these spent fuel elements which can be accommodated and still maintain the gamma activity outside under the permissible limit. The corresponding reactor power level is determined. The radioactivity calculations are performed for this internal storage at different axial levels to avoid the criticality of the reactor core. Transport theory is used in calculations based on collision probability for multi group cell calculations. Diffusion theory is used in three dimensions in the core calculations. The nuclear fuel history is traced and radioactive decay is calculated, since reactor fission products are very sensitive to power level. The radioactivity is calculated with a developed formula based on fuel basket loading integrity.

Sobhy, M.

1996-05-01

432

Reactor core design and modeling of the MIT research reactor for conversion to LEU

Feasibility design studies for conversion of the MIT Research Reactor (MITR) to LEU are described. Because the reactor fuel has a rhombic cross section, a special input processor was created in order to model the reactor in great detail with the REBUS-PC diffusion theory code, in 3D (triangular-z) geometry. Comparisons are made of fuel assembly power distributions and control blade worth vs. axial position, between REBUS-PC results and Monte Carlo predictions from the MCNP code. Results for the original HEU core at zero burnup are also compared with measurement. These two analysis methods showed remarkable agreement. Ongoing fuel cycle studies are summarized. A status report will be given as to results thus far that affect key design decisions. Future work plans and schedules to achieve completion of the conversion are presented. (author)

Newton, Thomas H. Jr. [Nuclear Reactor Laboratory, Massachusetts Institute of Technology, 138 Albany St., Cambridge, MA 02139 (United States); Olson, Arne P.; Stillman, John A. [RERTR Program, Argonne National Laboratory, Argonne, IL 60439 (United States)

2008-07-15

433

Reference: Trial I Page 35 2 Observed Signal Intensities for Core ACH03 - Agarose Reference: Trial I 35 3 Observed Signal Intensities for Core ACH12 - 80% D20 Reference: Trial 1 36 4 Observed Signal Intensities for Core ACH03 - 80% D20 Reference...: Trial 1 36 5 Comparison of Calculated Porosities: Trial I 6 Observed Signal Intensities for Core AC2 - NA=1: Trial 2 7 Observed Signal Intensities for Core JCR3 - NA=l: Trial 2 . . 8 Observed Signal Intensities for Core GS1 ? NA=1: Trial 2 9...

Sherman, Byron Blake

2012-06-07

434

Rotation and convective core overshoot in ? Ophiuchi

NASA Astrophysics Data System (ADS)

Context. Recent work on several ? Cephei stars has succeeded in constraining both their interior rotation profile and their convective core overshoot. In particular, a recent study focusing on ? Ophiuchi has shown that a convective core overshoot parameter of ?_ov = 0.44 is required to model the observed pulsation frequencies, significantly higher than for other stars of this type. Aims: We investigate the effects of rotation and overshoot in early type main sequence pulsators, such as ? Cephei stars, and attempt to use the low order pulsation frequencies to constrain these parameters. This will be applied to a few test models and the ? Cephei star ? Ophiuchi. Methods: We use the 2D stellar evolution code ROTORC and the 2D linear adiabatic pulsation code NRO to calculate pulsation frequencies for 9.5 M? models evolved to an age of 15.6 Myr. We calculate low order p-modes (? ? 2) for models with a range of rotation rates and convective core overshoot parameters. These low order modes are the same range of modes observed in ? Ophiuchi. Results: Using these models, we find that the convective core overshoot has a larger effect on the pulsation frequencies than the rotation, except in the most rapidly rotating models considered. When the differences in radii are accounted for by scaling the frequencies by ?(GM/R(40°)^3), the effects of rotation diminish, but are not entirely accounted for. Thus, this scaling emphasizes the differences produced by changing the convective core overshoot. We find that increasing the convective core overshoot decreases the large separation, while producing a slight increase in the small separations. We created a model frequency grid which spanned several rotation rates and convective core overshoot values. We used this grid to define a modified ?^2 statistic in order to determine the best fitting parameters from a set of observed frequencies. Using this statistic, we are able to recover the rotation velocity and convective core overshoot for a few test models. We have also performed a “hare and hound” exercise to see how well 1D models can recover these parameters. Finally, we discuss the case of the ? Cephei star ? Oph. Using the observed frequencies and a fixed mass and metallicity, we find a lower overshoot than previously determined, with ?_ov = 0.28 ± 0.05. Our determination of the rotation rate agrees well with both previous work and observations, around 30 km s-1.

Lovekin, C. C.; Goupil, M.-J.

2010-06-01

435

Critical Experiment Analyses by CHAPLET3D Code in Two and Three-Dimensional Core Models

A series of critical experiments has been analyzed by the deterministic method code CHAPLET-3D in two- and three-dimensional core configurations in which explicit core structures are represented. The results show that the three-dimensional core calculation model employed in CHAPLET-3D code is valid and useful to obtain fine resolution results by the deterministic method. Moreover, the conventional two-dimensional axial buckling calculation

Shinya KOSAKA; Toshikazu TAKEDA

2005-01-01

436

Background This is a progress report of the Alzheimer's Disease Neuroimaging Initiative (ADNI) PET Core. Methods The Core has supervised the acquisition, quality control, and analysis of longitudinal [18F]fluorodeoxyglucose PET (FDG-PET) data in approximately half of the ADNI cohort. In an “add on” study, approximately 100 subjects also underwent scanning with [11C]PIB-PET for amyloid imaging. The Core developed quality control procedures and standardized image acquisition by developing an imaging protocol that has been widely adopted in academic and pharmaceutical industry studies. Data processing provides users with scans that have identical orientation and resolution characteristics despite acquisition on multiple scanner models. The Core labs have used a number of different approaches to characterize differences between subject groups (AD, MCI, controls), to examine longitudinal change over time in glucose metabolism and amyloid deposition, and to assess the use of FDG-PET as a potential outcome measure in clinical trials. Results ADNI data indicate that FDG-PET increases statistical power over traditional cognitive measures, might aid subject selection, and could substantially reduce the sample size in a clinical trial. PIB-PET data showed expected group differences, and identified subjects with significant annual increases in amyloid load across the subject groups. The next activities of the PET core in ADNI will entail developing standardized protocols for amyloid imaging using the [18F]-labeled amyloid imaging agent AV45, which can be delivered to virtually all ADNI sites. Conclusions ADNI has demonstrated the feasibility and utility of multicenter PET studies and is helping to clarify the role of biomarkers in the study of aging and dementia. PMID:20451870

Jagust, William J.; Bandy, Dan; Chen, Kewei; Foster, Norman L.; Landau, Susan M.; Mathis, Chester A.; Price, Julie C.; Reiman, Eric M.; Skovronsky, Daniel; Koeppe, Robert A.

2010-01-01

437

The Underwater Blast Resistance of Metallic Sandwich Beams With Prismatic Lattice Cores

The finite element method is used to evaluate the underwater blast resistance of mono- lithic beams and sandwich beams containing prismatic lattice cores (Y-frame and corru- gated core) and an ideal foam core. Calculations are performed on both free-standing and end-clamped beams, and fluid-structure interaction effects are accounted for. It is found that the degree of core compression in the

G. J. McShane; V. S. Deshpande; N. A. Fleck

2007-01-01

438

Dislocation core structures and yield stress anomalies in molybdenum disilicide

Stacking fault energies in MoSi2 due to shear along have been calculated by modified embedded atom method (MEAM) calculations. Preliminary calculations have also been made of dislocation core structures and their response to applied stress. The results are used to investigate the configuration and mobility of 1\\/2 dislocations. Shear of 1\\/6 in the {103} plane of MoSi2 produces an anti-phase

T. E. Mitchell; M. I. Baskes; R. G. Hoagland; A. Misra

2001-01-01

439

Programmable calculator stress analysis

Advanced programmable alphanumeric calculators are well suited for closed-form calculation of pressure-vessel stresses. They offer adequate computing power, portability, special programming features, and simple interactive execution procedures. Representative programs that demonstrate calculator capabilities are presented. Problems treated are stress and strength calculations in thick-walled pressure vessels and the computation of stresses near head/pressure-vessel junctures.

Van Gulick, L.A.

1983-01-01

440

Core hole screening and decay rates of double core ionized first row hydrides.

Because of the high intensity, X-ray free electron lasers allow one to create and probe double core ionized states in molecules. The decay of these multiple core ionized states crucially determines the evolution of radiation damage in single molecule diffractive imaging experiments. Here we have studied the Auger decay in hydrides of first row elements after single and double core ionization by quantum mechanical ab initio calculations. In our approach the continuum wave function of the emitted Auger electron is expanded into spherical harmonics on a radial grid. The obtained decay rates of double K-shell vacancies were found to be systematically larger than those for the respective single K-shell vacancies, markedly exceeding the expected factor of two. This enhancement is attributed to the screening effects induced by the core hole. We propose a simple model, which is able to predict core hole decay rates in molecules with low Z elements based on the electron density in the vicinity of the core hole. PMID:23635135

Inhester, L; Groenhof, G; Grubmüller, H

2013-04-28

441

Effect of burn-up on the thermal conductivity of uranium-gadolinium dioxide up to 100 GWd/tHM

NASA Astrophysics Data System (ADS)

The thermal diffusivity of reactor irradiated (U,Gd)O2 fuels has been measured, for burn-ups from 33 to 97 GWd tHM-1 and for irradiation temperatures from 670 to 1580 K. Measurements under thermal annealing cycles were performed in order to investigate the recovery of the thermal conductivity as a function of temperature. The analysis of the results showed a lower thermal conductivity for (U,Gd)O2 when compared to UO2, with similar effects of the burn-up and irradiation temperature. A correlation for the thermal conductivity could be proposed on the basis of that for UO2 presented in an earlier work, which describes the separate effects of soluble fission products, of fission gas frozen in dynamical solution and of radiation damage.

Staicu, D.; Rondinella, V. V.; Walker, C. T.; Papaioannou, D.; Konings, R. J. M.; Ronchi, C.; Sheindlin, M.; Sasahara, A.; Sonoda, T.; Kinoshita, M.

2014-10-01

442

Air ingression calculations for selected plant transients using MELCOR

Two sets of MELCOR calculations have been completed studying the effects of air ingression on the consequences of various severe accident scenarios. One set of calculations analyzed a station blackout with surge line failure prior to vessel breach, starting from nominal operating conditions; the other set of calculations analyzed a station blackout occurring during shutdown (refueling) conditions. Both sets of analyses were for the Surry plant, a three-loop Westinghouse PWR. For both accident scenarios, a basecase calculation was done, and then repeated with air ingression from containment into the core region following core degradation and vessel failure. In addition to the two sets of analyses done for this program, a similar air-ingression sensitivity study was done as part of a low-power/shutdown PRA, with results summarized here; that PRA study also analyzed a station blackout occurring during shutdown (refueling) conditions, but for the Grand Gulf plant, a BWR/6 with Mark III containment. These studies help quantify the amount of air that would have to enter the core region to have a significant impact on the severe accident scenario, and demonstrate that one effect, of air ingression is substantial enhancement of ruthenium release. These calculations also show that, while the core clad temperatures rise more quickly due to oxidation with air rather than steam, the core also degrades and relocates more quickly, so that no sustained, enhanced core heatup is predicted to occur with air ingression.

Kmetyk, L.N.

1994-01-01

443

NSDL National Science Digital Library

The Voice over IP Calculator Web site actually consists of four free online tools that can be used to estimate bandwidth requirements and voice paths for a planned VoIP system. The four tools are: Lines and IP Bandwidth Calculator, Erlangs and Bandwidth Calculator, Minutes and Lines Calculator, and Erlangs and Lines Calculator. Each utility is very easy to use, but is mainly intended for experienced IT workers.

2008-02-12

444

Computerised gamma-ray emission tomography has been applied to single PWR UO2 fuel rods, with pellet averaged burnups of 52, 71, 91 and 126GWd\\/t respectively, for the determination of 134Cs, 137Cs and 154Eu internal radial distributions. State-of-the-art image reconstruction techniques, analytical and iterative, have been applied, evaluated and compared using test phantoms first and, in a second step, the actual measured

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

2009-01-01

445

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 (ASFAD and FEMAXI-V), which provide all necessary input for this model, mainly local temperatures and fuel matrix density across the radius. Comparisons between measurements and predictions of the PLUTON model are made on fuels with enrichments in the range 2.9 to 8.25% and with burnup between 21 000 and 64 000 MWd/t. It is shown that the PLUTON predictions are in good agreement with measurements as well as with predictions of the well-known TUBRNP model. The proposed model is flexibly applicable for all types of light water reactor (LWR) fuels, including mixed oxide, and for fuel tested in the Organization for Economic Corporation and Development's Halden heavy water reactor. The PLUTON three-group model is based on analytical (theoretical) consideration of neutron absorption in a resonant region of the fuel in its apparent form. It makes the model more flexible in comparison with the semi-empirical TUBRNP one-group model and allows the physically based model analysis of commercial LWR-type fuels at high burnup as well as analysis of experimental fuel rods tested in the Halden heavy water reactor, which is one of the main test reactors in the world. The differences in fuel behavior in the Halden reactor in terms of burnup distribution and plutonium buildup can be more clearly understood with the PLUTON model.

Lemehov, Sergei; Nakamura, Jinichi; Suzuki, Motoe [Japan Atomic Energy Research Institute (Japan)

2001-02-15

446

High-burnup fuel failure during a reactivity-initiated accident has been a subject of safety-related concern. Because of wide variations in cladding metallurgical and simulation test conditions, it has been difficult to understand the complex failure behavior observed in tests in the SPERT, NSRR and CABRI reactors. In this paper, we propose a failure model that is based on temperature-sensitive tensile properties

H. M. Chung; T. F. Kassner

1998-01-01