Two level calculation of assembly neutronic data libraries; Schema de calcul de bibliotheques a deux niveaux

Energy Technology Data Exchange (ETDEWEB)

Benomar, M

1998-09-01

The neutronic modeling of a nuclear reactor core requires 2 steps. The first step that is called transport calculation, is an accurate modeling of each type of assemblies put in a simple configuration. APOLLO2, a French neutronic code is used. This step allows the constitution of assembly data libraries. The second step represents the computing of the whole core by the diffusion theory and by using the data libraries defined in the first step. This work is dedicated to the improvement of the first step by allowing both a 172 group energy meshing and a two-dimension spatial processing. (A.C.) 7 refs.

Integrated Radiation Transport and Nuclear Fuel Performance for Assembly-Level Simulations

Energy Technology Data Exchange (ETDEWEB)

Clarno, Kevin T [ORNL; Hamilton, Steven P [ORNL; Philip, Bobby [ORNL; Berrill, Mark A [ORNL; Sampath, Rahul S [ORNL; Allu, Srikanth [ORNL; Pugmire, Dave [ORNL; Dilts, Gary [Los Alamos National Laboratory (LANL); Banfield, James E [ORNL

2012-02-01

The Advanced Multi-Physics (AMP) Nuclear Fuel Performance code (AMPFuel) is focused on predicting the temperature and strain within a nuclear fuel assembly to evaluate the performance and safety of existing and advanced nuclear fuel bundles within existing and advanced nuclear reactors. AMPFuel was extended to include an integrated nuclear fuel assembly capability for (one-way) coupled radiation transport and nuclear fuel assembly thermo-mechanics. This capability is the initial step toward incorporating an improved predictive nuclear fuel assembly modeling capability to accurately account for source-terms and boundary conditions of traditional (single-pin) nuclear fuel performance simulation, such as the neutron flux distribution, coolant conditions, and assembly mechanical stresses. A novel scheme is introduced for transferring the power distribution from the Scale/Denovo (Denovo) radiation transport code (structured, Cartesian mesh with smeared materials within each cell) to AMPFuel (unstructured, hexagonal mesh with a single material within each cell), allowing the use of a relatively coarse spatial mesh (10 million elements) for the radiation transport and a fine spatial mesh (3.3 billion elements) for thermo-mechanics with very little loss of accuracy. In addition, a new nuclear fuel-specific preconditioner was developed to account for the high aspect ratio of each fuel pin (12 feet axially, but 1 4 inches in diameter) with many individual fuel regions (pellets). With this novel capability, AMPFuel was used to model an entire 17 17 pressurized water reactor fuel assembly with many of the features resolved in three dimensions (for thermo-mechanics and/or neutronics), including the fuel, gap, and cladding of each of the 264 fuel pins; the 25 guide tubes; the top and bottom structural regions; and the upper and lower (neutron) reflector regions. The final, full assembly calculation was executed on Jaguar using 40,000 cores in under 10 hours to model over 162

Feasibility study on embedded transport core calculations

International Nuclear Information System (INIS)

Ivanov, B.; Zikatanov, L.; Ivanov, K.

2007-01-01

The main objective of this study is to develop an advanced core calculation methodology based on embedded diffusion and transport calculations. The scheme proposed in this work is based on embedded diffusion or SP 3 pin-by-pin local fuel assembly calculation within the framework of the Nodal Expansion Method (NEM) diffusion core calculation. The SP 3 method has gained popularity in the last 10 years as an advanced method for neutronics calculation. NEM is a multi-group nodal diffusion code developed, maintained and continuously improved at the Pennsylvania State University. The developed calculation scheme is a non-linear iteration process, which involves cross-section homogenization, on-line discontinuity factors generation, and boundary conditions evaluation by the global solution passed to the local calculation. In order to accomplish the local calculation, a new code has been developed based on the Finite Elements Method (FEM), which is capable of performing both diffusion and SP 3 calculations. The new code will be used in the framework of the NEM code in order to perform embedded pin-by-pin diffusion and SP 3 calculations on fuel assembly basis. The development of the diffusion and SP 3 FEM code is presented first following by its application to several problems. Description of the proposed embedded scheme is provided next as well as the obtained preliminary results of the C3 MOX benchmark. The results from the embedded calculations are compared with direct pin-by-pin whole core calculations in terms of accuracy and efficiency followed by conclusions made about the feasibility of the proposed embedded approach. (authors)

Transport synthetic acceleration for long-characteristics assembly-level transport problems

Energy Technology Data Exchange (ETDEWEB)

Zika, M R; Adams, M L

2000-02-01

The authors apply the transport synthetic acceleration (TSA) scheme to the long-characteristics spatial discretization for the two-dimensional assembly-level transport problem. This synthetic method employs a simplified transport operator as its low-order approximation. Thus, in the acceleration step, the authors take advantage of features of the long-characteristics discretization that make it particularly well suited to assembly-level transport problems. The main contribution is to address difficulties unique to the long-characteristics discretization and produce a computationally efficient acceleration scheme. The combination of the long-characteristics discretization, opposing reflecting boundary conditions (which are present in assembly-level transport problems), and TSA presents several challenges. The authors devise methods for overcoming each of them in a computationally efficient way. Since the boundary angular data exist on different grids in the high- and low-order problems, they define restriction and prolongation operations specific to the method of long characteristics to map between the two grids. They implement the conjugate gradient (CG) method in the presence of opposing reflection boundary conditions to solve the TSA low-order equations. The CG iteration may be applied only to symmetric positive definite (SPD) matrices; they prove that the long-characteristics discretization yields an SPD matrix. They present results of the acceleration scheme on a simple test problem, a typical pressurized water reactor assembly, and a typical boiling water reactor assembly.

Transport synthetic acceleration for long-characteristics assembly-level transport problems

International Nuclear Information System (INIS)

Zika, M.R.; Adams, M.L.

2000-01-01

The authors apply the transport synthetic acceleration (TSA) scheme to the long-characteristics spatial discretization for the two-dimensional assembly-level transport problem. This synthetic method employs a simplified transport operator as its low-order approximation. Thus, in the acceleration step, the authors take advantage of features of the long-characteristics discretization that make it particularly well suited to assembly-level transport problems. The main contribution is to address difficulties unique to the long-characteristics discretization and produce a computationally efficient acceleration scheme. The combination of the long-characteristics discretization, opposing reflecting boundary conditions (which are present in assembly-level transport problems), and TSA presents several challenges. The authors devise methods for overcoming each of them in a computationally efficient way. Since the boundary angular data exist on different grids in the high- and low-order problems, they define restriction and prolongation operations specific to the method of long characteristics to map between the two grids. They implement the conjugate gradient (CG) method in the presence of opposing reflection boundary conditions to solve the TSA low-order equations. The CG iteration may be applied only to symmetric positive definite (SPD) matrices; they prove that the long-characteristics discretization yields an SPD matrix. They present results of the acceleration scheme on a simple test problem, a typical pressurized water reactor assembly, and a typical boiling water reactor assembly

Transport Synthetic Acceleration for Long-Characteristics Assembly-Level Transport Problems

International Nuclear Information System (INIS)

Zika, Michael R.; Adams, Marvin L.

2000-01-01

We apply the transport synthetic acceleration (TSA) scheme to the long-characteristics spatial discretization for the two-dimensional assembly-level transport problem. This synthetic method employs a simplified transport operator as its low-order approximation. Thus, in the acceleration step, we take advantage of features of the long-characteristics discretization that make it particularly well suited to assembly-level transport problems. Our main contribution is to address difficulties unique to the long-characteristics discretization and produce a computationally efficient acceleration scheme.The combination of the long-characteristics discretization, opposing reflecting boundary conditions (which are present in assembly-level transport problems), and TSA presents several challenges. We devise methods for overcoming each of them in a computationally efficient way. Since the boundary angular data exist on different grids in the high- and low-order problems, we define restriction and prolongation operations specific to the method of long characteristics to map between the two grids. We implement the conjugate gradient (CG) method in the presence of opposing reflection boundary conditions to solve the TSA low-order equations. The CG iteration may be applied only to symmetric positive definite (SPD) matrices; we prove that the long-characteristics discretization yields an SPD matrix. We present results of our acceleration scheme on a simple test problem, a typical pressurized water reactor assembly, and a typical boiling water reactor assembly

Supplementary neutron flux calculations for the ORNL pool critical assembly pressure vessel facility

Energy Technology Data Exchange (ETDEWEB)

Maerker, R.E.; Maudlin, P.J.

1981-02-01

A three-dimensional Monte Carlo calculation was performed to estimate the neutron flux in the 8/7 configuration of the ORNL Pool Critical Assembly Pressure Vessel Facility. The calculational tool was the multigroup transport code MORSE operated in the adjoint mode. The MORSE flux results compared well with those using a previously adopted procedure for constructing a three-dimensional flux from one- and two-dimensional discrete ordinates calculations using the DOT-IV code. This study concluded that use of these discrete ordinates constructions in previous calculations is sufficiently accurate and does not account for the existing discrepancies between calculation and experiment.

Supplementary neutron flux calculations for the ORNL pool critical assembly pressure vessel facility

International Nuclear Information System (INIS)

Maerker, R.E.; Maudlin, P.J.

1981-02-01

A three-dimensional Monte Carlo calculation was performed to estimate the neutron flux in the 8/7 configuration of the ORNL Pool Critical Assembly Pressure Vessel Facility. The calculational tool was the multigroup transport code MORSE operated in the adjoint mode. The MORSE flux results compared well with those using a previously adopted procedure for constructing a three-dimensional flux from one- and two-dimensional discrete ordinates calculations using the DOT-IV code. This study concluded that use of these discrete ordinates constructions in previous calculations is sufficiently accurate and does not account for the existing discrepancies between calculation and experiment

Development of 2-D/1-D fusion method for three-dimensional whole-core heterogeneous neutron transport calculations

International Nuclear Information System (INIS)

Lee, Gil Soo

2006-02-01

To describe power distribution and multiplication factor of a reactor core accurately, it is necessary to perform calculations based on neutron transport equation considering heterogeneous geometry and scattering angles. These calculations require very heavy calculations and were nearly impossible with computers of old days. From the limitation of computing power, traditional approach of reactor core design consists of heterogeneous transport calculation in fuel assembly level and whole core diffusion nodal calculation with assembly homogenized properties, resulting from fuel assembly transport calculation. This approach may be effective in computation time, but it gives less accurate results for highly heterogeneous problems. As potential for whole core heterogeneous transport calculation became more feasible owing to rapid development of computing power during last several years, the interests in two and three dimensional whole core heterogeneous transport calculations by deterministic method are increased. For two dimensional calculation, there were several successful approaches using even parity transport equation with triangular meshes, S N method with refined rectangular meshes, the method of characteristics (MOC) with unstructured meshes, and so on. The work in this thesis originally started from the two dimensional whole core heterogeneous transport calculation by using MOC. After successful achievement in two dimensional calculation, there were efforts in three-dimensional whole-core heterogeneous transport calculation using MOC. Since direct extension to three dimensional calculation of MOC requires too much computing power, indirect approach to three dimensional calculation was considered.Thus, 2D/1D fusion method for three dimensional heterogeneous transport calculation was developed and successfully implemented in a computer code. The 2D/1D fusion method is synergistic combination of the MOC for radial 2-D calculation and S N -like methods for axial 1

An iterative homogenization technique that preserves assembly core exchanges

International Nuclear Information System (INIS)

Mondot, Ph.; Sanchez, R.

2003-01-01

A new interactive homogenization procedure for reactor core calculations is proposed that requires iterative transport assembly and diffusion core calculations. At each iteration the transport solution of every assembly type is used to produce homogenized cross sections for the core calculation. The converged solution gives assembly fine multigroup transport fluxes that preserve macro-group assembly exchanges in the core. This homogenization avoids the periodic lattice-leakage model approximation and gives detailed assembly transport fluxes without need of an approximated flux reconstruction. Preliminary results are given for a one-dimensional core model. (authors)

Detailed channel thermal-hydraulic calculation of nuclear reactor fuel assemblies

International Nuclear Information System (INIS)

Zhukov, A.V.; Sorokin, A.P.; Ushakov, P.A.; Yur'ev, Yu.S.

1981-01-01

The system of equations of mass balance, quantity of motion and energy used in calculation of nuclear reactor fuel assemblies is obtained. The equation system is obtained on the base of integral equations of hydrodynamics interaction in assemblies of smooth fuel elements and fuel elements with wire packing. The calculation results of coolant heating distributions by the fast reactor assembly channels are presented. The analysis of the results obtained shows that interchannel exchange essentially uniforms the coolant heating distribution in the peripheral range of the assembly but it does not remove non-uniformity caused by power distribution non-uniformity in the cross section. Geometry of the peripheral assembly range plays an essential role in the heating distribution. Change of the calculation gap between the peripheral fuel elements and assembly shells can result either in superheating or in subcooling in the peripheral channels relatively to joint internal channels of the assembly. Heat supply to the coolant passing through interassembly gaps decreases temperature in the assembly periphery and results in the increase of temperature non-uniformity by the perimeter of peripheral fuel elements. It is concluded that the applied method of the channel-by-channel calculation is ef-- fective in thermal-physical calculation of nuclear reactor fuel assemblies and it permits to solve a wide range of problems [ru

Improvement of the efficiency of two-dimensional multigroup transport calculations assuming isotropic reflection with multilevel spatial discretisation

International Nuclear Information System (INIS)

Stankovski, Z.; Zmijarevic, I.

1987-06-01

This paper presents two approximations used in multigroup two-dimensional transport calculations in large, very homogeneous media: isotropic reflection together with recently proposed group-dependent spatial representations. These approximations are implemented as standard options in APOLLO 2 assembly transport code. Presented example calculations show that significant savings in computational costs are obtained while preserving the overall accuracy

Transportation of part supply improvement in agricultural machinery assembly plant

Science.gov (United States)

Saysaman, Anusit; Chutima, Parames

2018-02-01

This research focused on the problem caused by the transportation of part supply in agricultural machinery assembly plant in Thailand, which is one of the processes that are critical to the whole production process. If poorly managed, it will affect transportation of part supply, the emergence of sink cost, quality problems, and the ability to respond to the needs of the customers in time. Since the competition in the agricultural machinery market is more intense, the efficiency of part transportation process has to be improved. In this study, the process of transporting parts of the plant was studied and it was found that the efficiency of the process of transporting parts from the sub assembly line to its main assembly line was 83%. The approach to the performance improvement is done by using the Lean tool to limit wastes based on the ECRS principle and applying pull production system by changing the transportation method to operate as milkrun for transportation of parts to synchronize with the part demands of the main assembly line. After the transportation of parts from sub-assembly line to the main assembly line was improved, the efficiency raised to 98% and transportation process cost was saved to 540,000 Baht per year.

Transport of fresh MOX fuel assemblies for the Monju initial core

International Nuclear Information System (INIS)

Kurakami, J.; Ouchi, Y.; Usami, M.

1997-01-01

Transport of fresh MOX fuel assemblies for the prototype FBR MONJU initial core started in July 1992 and ended in March 1994. As many as 205 fresh MOX fuel assemblies for an inner core, 91 assemblies for an outer core and 5 assemblies for testing) were transported in nine transport missions. The packaging for fuel assemblies, which has shielding and shock absorbing material inside, meets IAEA regulatory requirements for Type B(U) packaging including hypothetical accident conditions such as the 9 m drop test, fire test, etc. Moreover, this package design feature such advanced technologies as high performance neutron shielding material and an automatic hold-down mechanism for the fuel assemblies. Every effort was made to carry out safe transport in conjunction with the cooperation of every competent organisation. This effort includes establishment of the transport control centre, communication training, and accompanying of the radiation monitoring expert. No transport accident occurred during the transport and all the transport missions were successfully completed on schedule. (Author)

Microfluidic device for the assembly and transport of microparticles

Science.gov (United States)

James, Conrad D [Albuquerque, NM; Kumar, Anil [Framingham, MA; Khusid, Boris [New Providence, NJ; Acrivos, Andreas [Stanford, CA

2010-06-29

A microfluidic device comprising independently addressable arrays of interdigitated electrodes can be used to assembly and transport large-scale microparticle structures. The device and method uses collective phenomena in a negatively polarized suspension exposed to a high-gradient strong ac electric field to assemble the particles into predetermined locations and then transport them collectively to a work area for final assembly by sequentially energizing the electrode arrays.

Multigroup cross section collapsing optimization of a He-3 detector assembly model using deterministic transport techniques

International Nuclear Information System (INIS)

Huang, Mi; Yi, Ce; Manalo, Kevin L.; Sjoden, Glenn E.

2011-01-01

Multigroup optimization is performed on a neutron detector assembly to examine the validity of transport response in forward and adjoint modes. For SN transport simulations, we discuss the multigroup collapse of an 80 group library to 40, 30, and 16 groups, constructed from using the 3-D parallel PENTRAN and macroscopic cross section collapsing with YGROUP contribution weighting. The difference in using P_1 and P_3 Legendre order in scattering cross sections is investigated; also, associated forward and adjoint transport responses are calculated. We conclude that for the block analyzed, a 30 group cross section optimizes both computation time and accuracy relative to the 80 group transport calculations. (author)

Search for a transport method for the calculation of the PWR control and safety clusters

International Nuclear Information System (INIS)

Bruna, G.B.; Van Frank, C.; Vergain, M.L.; Chauvin, J.P.; Palmiotti, G.; Nobile, M.

1990-01-01

The project studies of power reactors rely mainly on diffusion calculations, but transport ones are often needed for assessing fine effects, intimately linked to geometry and spectrum heterogeneities. Accurate transport computations are necessary, in particular, for shielded cross section generation, and when homogenization and dishomogenization processes are involved. The transport codes, generally, offer the user a variety of computational options, related to different approximation levels. In every case, it is obviously desirable to be able to choose the reliable degree of approximation to be accepted in any particular computational circumstance of the project. The search for such adapted procedures is to be made on the basis of critical experiments. In our studies, this task was made possible by the availability of suitable results of the CAMELEON critical experiment, carried on in the EOLE facility at CEA's Center of Cadarache. In this paper, we summarize some of the work in progress at FRAMATOME on the definition of an assembly based transport calculation scheme to be used for PWR control and safety cluster computations. Two main items, devoted to the search of the optimum computational procedures, are presented here: - a parametrical study on computational options, made in an infinite medium assembly geometry, - a series of comparisons between calculated and experimental values of pin power distribution

Neutron transport study based on assembly modular ray tracing MOC method

International Nuclear Information System (INIS)

Tian Chao; Zheng Youqi; Li Yunzhao; Li Shuo; Chai Xiaoming

2015-01-01

It is difficulty for the MOC method based on Cell Modular Ray Tracing to deal with the irregular geometry such as the water gap between the PWR lattices. Hence, the neutron transport code NECP-Medlar based on Assembly Modular Ray Tracing is developed. CMFD method is used to accelerate the transport calculation. The numerical results of the 2D C5G7 benchmark and typical PWR lattice prove that NECP-Medlar has an excellent performance in terms of accuracy and efficiency. Besides, NECP-Medlar can describe clearly the flux distribution of the lattice with water gap. (authors)

TEMP-M program for thermal-hydraulic calculation of fast reactor fuel assemblies

International Nuclear Information System (INIS)

Bogoslovskaya, C.P.; Sorokin, A.P.; Tikhomirov, B.B.; Titov, P.A.; Ushakov, P.A.

1983-01-01

TEMP-M program (Fortran, BESM-6 computer) for thermal-hydraulic calculation of fast reactor fuel assemblies is described. Results of calculation of temperature field in a 127 fuel element assembly of BN-600, reactor accomplished according to TEMP-N program are considered as an example. Algorithm, realized in the program, enables to calculate the distributions of coolant heating, fuel element temperature (over perimeter and length) and assembly shell temperature. The distribution of coolant heating in assembly channels is determined from a solution of the balance equation system which accounts for interchannel exchange, nonadiabatic conditions on the assembly shell. The TEMP-M program gives necessary information for calculation of strength, seviceability of fast reactor core elements, serves an effective instrument for calculations when projecting reactor cores and analyzing thermal-hydraulic characteristics of operating reactor fuel assemblies

An improved assembly for the transport of fuel elements

International Nuclear Information System (INIS)

Myers, G.

1979-01-01

An improved assembly for the transport and storage of radioactive nuclear fuel elements is described. The fuel element transport canister is of the type in which the fuel elements are submerged in liquid with a self regulating ullage system, so that the fuel elements are always submerged in the liquid even when the assembly is used in one orientation during loading and another orientation during transportation. (UK)

Guideline of Monte Carlo calculation. Neutron/gamma ray transport simulation by Monte Carlo method

CERN Document Server

2002-01-01

This report condenses basic theories and advanced applications of neutron/gamma ray transport calculations in many fields of nuclear energy research. Chapters 1 through 5 treat historical progress of Monte Carlo methods, general issues of variance reduction technique, cross section libraries used in continuous energy Monte Carlo codes. In chapter 6, the following issues are discussed: fusion benchmark experiments, design of ITER, experiment analyses of fast critical assembly, core analyses of JMTR, simulation of pulsed neutron experiment, core analyses of HTTR, duct streaming calculations, bulk shielding calculations, neutron/gamma ray transport calculations of the Hiroshima atomic bomb. Chapters 8 and 9 treat function enhancements of MCNP and MVP codes, and a parallel processing of Monte Carlo calculation, respectively. An important references are attached at the end of this report.

Calculation of neutron importance function in fissionable assemblies using Monte Carlo method

International Nuclear Information System (INIS)

Feghhi, S. A. H.; Afarideh, H.; Shahriari, M.

2007-01-01

The purpose of the present work is to develop an efficient solution method to calculate neutron importance function in fissionable assemblies for all criticality conditions, using Monte Carlo Method. The neutron importance function has a well important role in perturbation theory and reactor dynamic calculations. Usually this function can be determined by calculating adjoint flux through out solving the Adjoint weighted transport equation with deterministic methods. However, in complex geometries these calculations are very difficult. In this article, considering the capabilities of MCNP code in solving problems with complex geometries and its closeness to physical concepts, a comprehensive method based on physical concept of neutron importance has been introduced for calculating neutron importance function in sub-critical, critical and supercritical conditions. For this means a computer program has been developed. The results of the method has been benchmarked with ANISN code calculations in 1 and 2 group modes for simple geometries and their correctness has been approved for all three criticality conditions. Ultimately, the efficiency of the method for complex geometries has been shown by calculation of neutron importance in MNSR research reactor

Calculation of Permeability inside the Basket including one Fuel Assembly

Energy Technology Data Exchange (ETDEWEB)

Yu, Seung Hwan; Bang, Kyung Sik; Lee, Ju an; Choi, Woo Seok [KAERI, Daejeon (Korea, Republic of)

2016-05-15

In general, the porous media model and the effective thermal conductivity were used to simply the fuel assembly. The methods of calculating permeability were compared considering the flow inside a basket which includes a nuclear fuel. Detailed fuel assembly was a computational modeling and the flow characteristics were investigated. The flow inside the basket which included a fuel assembly is analyzed by CFD. As the height of the fuel assembly increases, the pressure drop linearly increased. The inertia resistance could be neglected. Three methods to calculate the permeability were compared. The permeability by the friction factor is 50% less than the permeability by wall shear stress and pressure drop.

Tailoring Quantum Dot Assemblies to Extend Exciton Coherence Times and Improve Exciton Transport

Science.gov (United States)

Seward, Kenton; Lin, Zhibin; Lusk, Mark

2012-02-01

The motion of excitons through nanostructured assemblies plays a central role in a wide range of physical phenomena including quantum computing, molecular electronics, photosynthetic processes, excitonic transistors and light emitting diodes. All of these technologies are severely handicapped, though, by quasi-particle lifetimes on the order of a nanosecond. The movement of excitons must therefore be as efficient as possible in order to move excitons meaningful distances. This is problematic for assemblies of small Si quantum dots (QDs), where excitons quickly localize and entangle with dot phonon modes. Ensuing exciton transport is then characterized by a classical random walk reduced to very short distances because of efficient recombination. We use a combination of master equation (Haken-Strobl) formalism and density functional theory to estimate the rate of decoherence in Si QD assemblies and its impact on exciton mobility. Exciton-phonon coupling and Coulomb interactions are calculated as a function of dot size, spacing and termination to minimize the rate of intra-dot phonon entanglement. This extends the time over which more efficient exciton transport, characterized by partial coherence, can be maintained.

Optimized design for TWR assembly by CFD calculations

International Nuclear Information System (INIS)

Lu Jianchao; Lu Chuan; Yan Mingyu

2013-01-01

High temperature difference in travelling wave reactor bundle was found in the previous work. It could not be used in bundle design. Various analysis focused on helical wrapped wires and assembly housing was carried out by CFD calculation which found that the helical wrapped wires could influence the temperature differences while the effect was not obvious. Adding the strips and fillets on the assembly housing could optimize the thermal characteristics greatly, which can be used in the TWR assembly design. (authors)

Calculation of neutron importance function in fissionable assemblies using Monte Carlo method

International Nuclear Information System (INIS)

Feghhi, S.A.H.; Shahriari, M.; Afarideh, H.

2007-01-01

The purpose of the present work is to develop an efficient solution method for the calculation of neutron importance function in fissionable assemblies for all criticality conditions, based on Monte Carlo calculations. The neutron importance function has an important role in perturbation theory and reactor dynamic calculations. Usually this function can be determined by calculating the adjoint flux while solving the adjoint weighted transport equation based on deterministic methods. However, in complex geometries these calculations are very complicated. In this article, considering the capabilities of MCNP code in solving problems with complex geometries and its closeness to physical concepts, a comprehensive method based on the physical concept of neutron importance has been introduced for calculating the neutron importance function in sub-critical, critical and super-critical conditions. For this propose a computer program has been developed. The results of the method have been benchmarked with ANISN code calculations in 1 and 2 group modes for simple geometries. The correctness of these results has been confirmed for all three criticality conditions. Finally, the efficiency of the method for complex geometries has been shown by the calculation of neutron importance in Miniature Neutron Source Reactor (MNSR) research reactor

Calculation of drop course of control rod assembly in PWR

International Nuclear Information System (INIS)

Zhou Xiaojia; Mao Fei; Min Peng; Lin Shaoxuan

2013-01-01

The validation of control rod drop performance is an important part of safety analysis of nuclear power plant. Development of computer code for calculating control rod drop course will be useful for validating and improving the design of control rod drive line. Based on structural features of the drive line, the driving force on moving assembly was analyzed and decomposed, the transient value of each component of the driving force was calculated by choosing either theoretical method or numerical method, and the simulation code for calculating rod cluster control assembly (RCCA) drop course by time step increase was achieved. The analysis results of control rod assembly drop course calculated by theoretical model and numerical method were validated by comparing with RCCA drop test data of Qinshan Phase Ⅱ 600 MW PWR. It is shown that the developed RCCA drop course calculation code is suitable for RCCA in PWR and can correctly simulate the drop course and the stress of RCCA. (authors)

New calculations for critical assemblies using MCNP4B

International Nuclear Information System (INIS)

Adams, A.A.; Frankle, S.C.; Little, R.C.

1997-07-01

A suite of 41 criticality benchmarks has been modeled using MCNP trademark (version 4B). Most of the assembly specifications were obtained from the Cross Section Evaluation Working Group (CSEWG) and the International Criticality Safety Benchmark Evaluation Project (ICSBEP) compendiums of experimental benchmarks. A few assembly specifications were obtained from experimental papers. The suite contains thermal and fast assemblies, bare and reflected assemblies, and emphasizes 233 U, 235 U, 238 U, and 239 Pu. The values of k eff for each assembly in the suite were calculated using MCNP libraries derived primarily from release 2 of ENDF/B-V and release 2 of ENDF/B-VI. The results show that the new ENDF/B-VI.2 evaluations for H, O, N, B, 235 U, 238 U, and 239 Pu can have a significant impact on the values of k eff . In addition to the integral quantity k eff , several additional experimental measurements were performed and documented. These experimental measurements include central fission and reaction-rate ratios for various isotopes, and neutron leakage and flux spectra. They provide more detailed information about the accuracy of the nuclear data than can k eff . Comparison calculations were performed using both ENDF/B-V.2 and ENDF/B-VI.2-based data libraries. The purpose of this paper is to compare the results of these additional calculations with experimental data, and to use these results to assess the quality of the nuclear data

On the adequacy of Cartesian geometry discrete ordinates solutions for assembly calculations

International Nuclear Information System (INIS)

Schunert, S.; Azmy, Y. Y.

2009-01-01

The current generation of lattice codes employs the method of Collision Probabilities (CP), the Method of Characteristics (MOC) or methods derived thereof to solve the two-dimensional multigroup transport equation on the assembly level. We compare the attainable solution accuracy of the lattice code DRAGON to the accuracy of the Discrete Ordinates (DO) code DORT on the basis of the two-dimensional GE-13 assembly in order to determine if the DO on Cartesian meshes is suitable as flux solver in future lattice codes. If DO exhibits high accuracy for assembly configurations, the next question is at what computational expense compared to traditional assembly codes. For this purpose DORT and DRAGON are required to converge to a reference solution, obtained by a multigroup MCNP calculation, with increasing angular quadrature order and decreasing spatial cell size; additionally for DRAGON the reference solution must be approached with increasing tracking density. The convergence of the two codes is judged via the multiplication factor, the pin wise relative error in the fission production rate, it's RMS and the maximum of it's absolute value over all pins. Additionally the computational cost of the obtained solutions is judged via the user CPU time. Although the multiplication factor computed by both codes converges with refinement of the employed meshes, the maximum deviation error of the fission production rate in the central region of the assembly remains unsatisfactorily high for CP and MOC. (authors)

Burnup credit calculations for criticality safety justification for RBMK-1000 spent fuel of transport and storage systems

Directory of Open Access Journals (Sweden)

V. V. Galchenko

2010-12-01

Full Text Available In present paper the burnup credit calculations for TK-8 transport container and SVJP-1 spent fuel storage fa-cility of pool type with RBMK-1000 spent fuel during 100-years of cooling time were performed for criticality safety analysis purpose using MCNP and SCALE codes. Only actinides were taken into account for these critical systems. Two approaches were analyzed with isotopes distribution calculations along fuel assembly height and without it. The results show that subcriticality margin is increased considerably using burnup credit and isotopes distribution along fuel assembly height made this value more reasonable.

Calculation of Savannah River K Reactor Mark-22 assembly LOCA/ECS power limits

International Nuclear Information System (INIS)

Fischer, S.R.; Farman, R.F.; Birdsell, S.A.

1992-01-01

This paper summarizes the results of TRAC-PF1/MOD3 calculations of Mark-22 fuel assembly of loss-of-coolant accident/emergency cooling system (LOCA/ECS) power limits for the Savannah River Site (SRS) K Reactor. This effort was part of a larger effort undertaken by the Los Alamos National Laboratory for the US Department of Energy to perform confirmatory power limits calculations for the SRS K Reactor. A method using a detailed three-dimensional (3D) TRAC model of the Mark-22 fuel assembly was developed to compute LOCA/ECS power limits. Assembly power was limited to ensure that no point on the fuel assembly walls would exceed the local saturation temperature. The detailed TRAC model for the Mark-22 assembly consisted of three concentric 3D vessel components which simulated the two targets, two fuel tubes, and three main flow channels of the fuel assembly. The model included 100% eccentricity between the assembly annuli and a 20% power tilt. Eccentricity in the radial alignment of the assembly annuli arises because axial spacer ribs that run the length of the fuel and targets are used. Wall-shear, interfacial-shear, and wall heat-transfer correlations were developed and implemented in TRAC-PF1/MOD3 specifically for modeling flow and heat transfer in the narrow ribbed annuli encountered in the Mark-22 fuel assembly design. We established the validity of these new constitutive models using separate-effects benchmarks. TRAC system calculations of K Reactor indicated that the limiting ECS-phase accident is a double-ended guillonite break in a process water line at the pump discharge (i.e., a PDLOCA). The fuel assembly with the minimum cooling potential is identified from this system calculation. Detailed assembly calculations then were performed using appropriate boundary conditions obtained from this limiting system LOCA. Coolant flow rates and pressure boundary conditions were obtained from this system calculation and applied to the detailed assembly model

VERA Pin and Fuel Assembly Depletion Benchmark Calculations by McCARD and DeCART

Energy Technology Data Exchange (ETDEWEB)

Park, Ho Jin; Cho, Jin Young [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2016-10-15

Monte Carlo (MC) codes have been developed and used to simulate a neutron transport since MC method was devised in the Manhattan project. Solving the neutron transport problem with the MC method is simple and straightforward to understand. Because there are few essential approximations for the 6- dimension phase of a neutron such as the location, energy, and direction in MC calculations, highly accurate solutions can be obtained through such calculations. In this work, the VERA pin and fuel assembly (FA) depletion benchmark calculations are performed to examine the depletion capability of the newly generated DeCART multi-group cross section library. To obtain the reference solutions, MC depletion calculations are conducted using McCARD. Moreover, to scrutinize the effect by stochastic uncertainty propagation, uncertainty propagation analyses are performed using a sensitivity and uncertainty (S/U) analysis method and stochastic sampling (S.S) method. It is still expensive and challenging to perform a depletion analysis by a MC code. Nevertheless, many studies and works for a MC depletion analysis have been conducted to utilize the benefits of the MC method. In this study, McCARD MC and DeCART MOC transport calculations are performed for the VERA pin and FA depletion benchmarks. The DeCART depletion calculations are conducted to examine the depletion capability of the newly generated multi-group cross section library. The DeCART depletion calculations give excellent agreement with the McCARD reference one. From the McCARD results, it is observed that the MC depletion results depend on how to split the burnup interval. First, only to quantify the effect of the stochastic uncertainty propagation at 40 DTS, the uncertainty propagation analyses are performed using the S/U and S.S. method.

Electron transport within transparent assemblies of tin-doped indium oxide colloidal nanocrystals

Science.gov (United States)

Grisolia, J.; Decorde, N.; Gauvin, M.; Sangeetha, N. M.; Viallet, B.; Ressier, L.

2015-08-01

Stripe-like compact assemblies of tin-doped indium oxide (ITO) colloidal nanocrystals (NCs) are fabricated by stop-and-go convective self-assembly (CSA). Systematic evaluation of the electron transport mechanisms in these systems is carried out by varying the length of carboxylate ligands protecting the NCs: butanoate (C4), octanoate (C8) and oleate (C18). The interparticle edge-to-edge distance L0, along with a number of carbon atoms in the alkyl chain of the coating ligand, are deduced from small-angle x-ray scattering (SAXS) measurements and exhibit a linear relationship with a slope of 0.11 nm per carbon pair unit. Temperature-dependent resistance characteristics are analyzed using several electron transport models: Efros-Shklovskii variable range hopping (ES-VRH), inelastic cotunneling (IC), regular island array and percolation. The analysis indicated that the first two models (ES-VRH and IC) fail to explain the observed behavior, and that only simple activated transport takes place in these systems under the experimental conditions studied (T = 300 K to 77 K). Related transport parameters were then extracted using the regular island array and percolation models. The effective tunneling decay constant βeff of the ligands and the Coulomb charging energy EC are found to be around 5.5 nm-1 and 25 meV, respectively, irrespective of ligand lengths. The theoretical tunneling decay constant β calculated using the percolation model is in the range 9 nm-1. Electromechanical tests on the ITO nanoparticle assemblies indicate that their sensitivities are as high as ˜30 and remain the same regardless of ligand lengths, which is in agreement with the constant effective βeff extracted from regular island array and percolation models.

Range calculations using multigroup transport methods

International Nuclear Information System (INIS)

Hoffman, T.J.; Robinson, M.T.; Dodds, H.L. Jr.

1979-01-01

Several aspects of radiation damage effects in fusion reactor neutron and ion irradiation environments are amenable to treatment by transport theory methods. In this paper, multigroup transport techniques are developed for the calculation of particle range distributions. These techniques are illustrated by analysis of Au-196 atoms recoiling from (n,2n) reactions with gold. The results of these calculations agree very well with range calculations performed with the atomistic code MARLOWE. Although some detail of the atomistic model is lost in the multigroup transport calculations, the improved computational speed should prove useful in the solution of fusion material design problems

Criticality coefficient calculation for a small PWR using Monte Carlo Transport Code

Energy Technology Data Exchange (ETDEWEB)

Trombetta, Debora M.; Su, Jian, E-mail: dtrombetta@nuclear.ufrj.br, E-mail: sujian@nuclear.ufrj.br [Coordenacao dos Programas de Pos-Graduacao em Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil); Chirayath, Sunil S., E-mail: sunilsc@tamu.edu [Department of Nuclear Engineering and Nuclear Security Science and Policy Institute, Texas A and M University, TX (United States)

2015-07-01

Computational models of reactors are increasingly used to predict nuclear reactor physics parameters responsible for reactivity changes which could lead to accidents and losses. In this work, preliminary results for criticality coefficient calculation using the Monte Carlo transport code MCNPX were presented for a small PWR. The computational modeling developed consists of the core with fuel elements, radial reflectors, and control rods inside a pressure vessel. Three different geometries were simulated, a single fuel pin, a fuel assembly and the core, with the aim to compare the criticality coefficients among themselves.The criticality coefficients calculated were: Doppler Temperature Coefficient, Coolant Temperature Coefficient, Coolant Void Coefficient, Power Coefficient, and Control Rod Worth. The coefficient values calculated by the MCNP code were compared with literature results, showing good agreement with reference data, which validate the computational model developed and allow it to be used to perform more complex studies. Criticality Coefficient values for the three simulations done had little discrepancy for almost all coefficients investigated, the only exception was the Power Coefficient. Preliminary results presented show that simple modelling as a fuel assembly can describe changes at almost all the criticality coefficients, avoiding the need of a complex core simulation. (author)

Experience feedback from the transportation of Framatome fuel assemblies

International Nuclear Information System (INIS)

Robin, M.E.; Gaillard, G.; Aubin, C.

1998-01-01

Framatome, the foremost world nuclear fuel manufacturer, has for 25 years been delivering fuel elements from its three factories (Dessel, Romans, Pierrelatte) to the various sites in France and abroad (Germany, Sweden, Belgium, China, Korea, South Africa, Switzerland). During this period, Framatome has built up experience and expertise in fuel element transportation by road, rail and sea. In this filed, the range of constraints is very wide: safety and environmental protection constraints; constraints arising from the control and protection of nuclear materials, contractual and financial constraints, media watchdogs. Through the experience feedback from the transportation of FRAMATOME assemblies, this paper addresses all the phases in the transportation of fresh fuel assemblies. (authors)

Development of 3-D FBR heterogeneous core calculation method based on characteristics method

International Nuclear Information System (INIS)

Takeda, Toshikazu; Maruyama, Manabu; Hamada, Yuzuru; Nishi, Hiroshi; Ishibashi, Junichi; Kitano, Akihiro

2002-01-01

A new 3-D transport calculation method taking into account the heterogeneity of fuel assemblies has been developed by combining the characteristics method and the nodal transport method. In the axial direction the nodal transport method is applied, and the characteristics method is applied to take into account the radial heterogeneity of fuel assemblies. The numerical calculations have been performed to verify 2-D radial calculations of FBR assemblies and partial core calculations. Results are compared with the reference Monte-Carlo calculations. A good agreement has been achieved. It is shown that the present method has an advantage in calculating reaction rates in a small region

Estimation of reactor core calculation by HELIOS/MASTER at power generating condition through DeCART, whole-core transport code

International Nuclear Information System (INIS)

Kim, H. Y.; Joo, H. G.; Kim, K. S.; Kim, G. Y.; Jang, M. H.

2003-01-01

The reactivity and power distribution errors of the HELIOS/MASTER core calculation under power generating conditions are assessed using a whole core transport code DeCART. For this work, the cross section tablesets were generated for a medium sized PWR following the standard procedure and two group nodal core calculations were performed. The test cases include the HELIOS calculations for 2-D assemblies at constant thermal conditions, MASTER 3D assembly calculations at power generating conditions, and the core calculations at HZP, HFP, and an abnormal power conditions. In all these cases, the results of the DeCART code in which pinwise thermal feedback effects are incorporated are used as the reference. The core reactivity, assemblywise power distribution, axial power distribution, peaking factor, and thermal feedback effects are then compared. The comparison shows that the error of the HELIOS/MASTER system in the core reactivity, assembly wise power distribution, pin peaking factor are only 100∼300 pcm, 3%, and 2%, respectively. As far as the detailed pinwise power distribution is concerned, however, errors greater than 15% are observed

Criticality calculations for a critical assembly, graphite moderate, using 20% enriched uranium

International Nuclear Information System (INIS)

Almeida Ferreira, A.C. de; Hukai, R.Y.

1975-01-01

The construction of a Zero Power Reactor (ZPR) at the Instituto de Energia Atomica in order to measure the neutron characteristics (parameters) of HTGR reactors is proposed. The necessary quantity fissile uranium for these measurements has been calculed. Criticality studies of graphite moderated critical assemblies containing thorium have been made and the critical mass of each of several typical commercial HTGR compositions has been calculated using computer codes HAMMER and CITATION. Assemblies investigated contained a central cylindrical core region, simulating a typical commercial HTGR composition, a uranium-graphite driver region and a outer pure graphite reflector region. It is concluded that a 10Kg inventory of fissile uranium will be required for a program of measurements utilizing each of the several calculated assemblies

Calculation of transport coefficients in an axisymmetric plasma

International Nuclear Information System (INIS)

Shumaker, D.E.

1977-01-01

A method of calculating the transport coefficient in an axisymmetric toroidal plasma is presented. This method is useful in calculating the transport coefficients in a Tokamak plasma confinement device. The particle density and temperature are shown to be a constant on a magnetic flux surface. Transport equations are given for the total particle flux and total energy flux crossing a closed toroidal surface. Also transport equations are given for the toroidal magnetic flux. A computer code was written to calculate the transport coefficients for a three species plasma, electrons and two species of ions. This is useful for calculating the transport coefficients of a plasma which contains impurities. It was found that the particle and energy transport coefficients are increased by a large amount, and the transport coefficients for the toroidal magnetic field are reduced by a small amount

An immersed body method for coupled neutron transport and thermal hydraulic simulations of PWR assemblies

International Nuclear Information System (INIS)

Jewer, S.; Buchan, A.G.; Pain, C.C.; Cacuci, D.G.

2014-01-01

Highlights: • A new method of coupled radiation transport, heat and momentum exchanges on fluids, and heat transfer simulations. • Simulation of the thermal hydraulics and radiative properties within whole PWR assemblies. • An immersed body method for modelling complex solid domains on practical computational meshes. - Abstract: A recently developed immersed body method is adapted and used to model a typical pressurised water reactor (PWR) fuel assembly. The approach is implemented with the numerical framework of the finite element, transient criticality code, FETCH which is composed of the neutron transport code, EVENT, and the CFD code, FLUIDITY. Within this framework the neutron transport equation, Navier–Stokes equations and a fluid energy conservation equation are solved in a coupled manner on a coincident structured or unstructured mesh. The immersed body method has been used to model the solid fuel pins. The key feature of this method is that the fluid/neutronic domain and the solid domain are represented by overlapping and non-conforming meshes. The main difficulty of this approach, for which a solution is proposed in this work, is the conservative mapping of the energy and momentum exchange between the fluid/neutronic mesh and the solid fuel pin mesh. Three numerical examples are presented which include a validation of the fuel pin submodel against an analytical solution; an uncoupled (no neutron transport solution) PWR fuel assembly model with a specified power distribution which was validated against the COBRA-EN subchannel analysis code; and finally a coupled model of a PWR fuel assembly with reflective neutron boundary conditions. Coupling between the fluid and neutron transport solutions is through the nuclear cross sections dependence on Doppler fuel temperature, coolant density and temperature, which was taken into account by using pre-calculated cross-section lookup tables generated using WIMS9a. The method was found to show good agreement

Tailorable Exciton Transport in Doped Peptide–Amphiphile Assemblies

Energy Technology Data Exchange (ETDEWEB)

Solomon, Lee A. [Center; Sykes, Matthew E. [Center; Wu, Yimin A. [Center; Schaller, Richard D. [Center; Department; Wiederrecht, Gary P. [Center; Fry, H. Christopher [Center

2017-08-29

Light-harvesting biomaterials are an attractive target in photovoltaics, photocatalysis, and artificial photosynthesis. Through peptide self-assembly, complex nanostructures can be engineered to study the role of chromophore organization during light absorption and energy transport. To this end, we demonstrate the one-dimensional transport of excitons along naturally occurring, light-harvesting, Zn-protoporphyrin IX chromophores within self-assembled peptide-amphiphile nanofibers. The internal structure of the nanofibers induces packing of the porphyrins into linear chains. We find that this peptide assembly can enable long-range exciton diffusion, yet it also induces the formation of excimers between adjacent molecules, which serve as exciton traps. Electronic coupling between neighboring porphyrin molecules is confirmed by various spectroscopic methods. The exciton diffusion process is then probed through transient photoluminescence and absorption measurements and fit to a model for one-dimensional hopping. Because excimer formation impedes exciton hopping, increasing the interchromophore spacing allows for improved diffusivity, which we control through porphyrin doping levels. We show that diffusion lengths of over 60 nm are possible at low porphyrin doping, representing an order of magnitude improvement over the highest doping fractions.

Three dimensions transport calculations for PWR core; Calcul de coeur R.E.P. en transport 3D

Energy Technology Data Exchange (ETDEWEB)

Richebois, E

2000-07-01

The objective of this work is to define improved 3-D core calculation methods based on the transport theory. These methods can be particularly useful and lead to more precise computations in areas of the core where anisotropy and steep flux gradients occur, especially near interface and boundary conditions and in regions of high heterogeneity (bundle with absorbent rods). In order to apply the transport theory a new method for calculating reflector constants has been developed, since traditional methods were only suited for 2-group diffusion core calculations and could not be extrapolated to transport calculations. In this thesis work, the new method for obtaining reflector constants is derived regardless of the number of energy groups and of the operator used. The core calculations results using the reflector constants thereof obtained have been validated on the EDF's power reactor Saint Laurent B1 with MOX loading. The advantages of a 3-D core transport calculation scheme have been highlighted as opposed to diffusion methods; there are a considerable number of significant effects and potential advantages to be gained in rod worth calculations for instance. These preliminary results obtained with on particular cycle will have to be confirmed by more systematic analysis. Accidents like MSLB (main steam line break) and LOCA (loss of coolant accident) should also be investigated and constitute challenging situations where anisotropy is high and/or flux gradients are steep. This method is now being validated for others EDF's PWRs' reactors, as well as for experimental reactors and other types of commercial reactors. (author)

Three dimensions transport calculations for PWR core; Calcul de coeur R.E.P. en transport 3D

Energy Technology Data Exchange (ETDEWEB)

Richebois, E

2000-07-01

The objective of this work is to define improved 3-D core calculation methods based on the transport theory. These methods can be particularly useful and lead to more precise computations in areas of the core where anisotropy and steep flux gradients occur, especially near interface and boundary conditions and in regions of high heterogeneity (bundle with absorbent rods). In order to apply the transport theory a new method for calculating reflector constants has been developed, since traditional methods were only suited for 2-group diffusion core calculations and could not be extrapolated to transport calculations. In this thesis work, the new method for obtaining reflector constants is derived regardless of the number of energy groups and of the operator used. The core calculations results using the reflector constants thereof obtained have been validated on the EDF's power reactor Saint Laurent B1 with MOX loading. The advantages of a 3-D core transport calculation scheme have been highlighted as opposed to diffusion methods; there are a considerable number of significant effects and potential advantages to be gained in rod worth calculations for instance. These preliminary results obtained with on particular cycle will have to be confirmed by more systematic analysis. Accidents like MSLB (main steam line break) and LOCA (loss of coolant accident) should also be investigated and constitute challenging situations where anisotropy is high and/or flux gradients are steep. This method is now being validated for others EDF's PWRs' reactors, as well as for experimental reactors and other types of commercial reactors. (author)

EPRI-LATTICE: a multigroup neutron transport code for light water reactor lattice physics calculations

International Nuclear Information System (INIS)

Jones, D.B.

1986-01-01

EPRI-LATTICE is a multigroup neutron transport computer code for the analysis of light water reactor fuel assemblies. It can solve the two-dimensional neutron transport problem by two distinct methods: (a) the method of collision probabilities and (b) the method of discrete ordinates. The code was developed by S. Levy Inc. as an account of work sponsored by the Electric Power Research Institute (EPRI). The collision probabilities calculation in EPRI-LATTICE (L-CP) is based on the same methodology that exists in the lattice codes CPM-2 and EPRI-CPM. Certain extensions have been made to the data representations of the CPM programs to improve the overall accuracy of the calculation. The important extensions include unique representations of scattering matrices and fission fractions (chi) for each composition in the problem. A new capability specifically developed for the EPRI-LATTICE code is a discrete ordinates methodology. The discrete ordinates calculation in EPRI-LATTICE (L-SN) is based on the discrete S/sub n/ methodology that exists in the TWODANT program. In contrast to TWODANT, which utilizes synthetic diffusion acceleration and supports multiple geometries, only the transport equations are solved by L-SN and only the data representations for the two-dimensional geometry are treated

Calculation of transport coefficients in an axisymmetric plasma

International Nuclear Information System (INIS)

Shumaker, D.E.

1976-01-01

A method of calculating the transport coefficient in an axisymmetric toroidal plasma is presented. This method is useful in calculating the transport coefficients in a Tokamak plasma confinement device. The particle density and temperature are shown to be a constant on a magnetic flux surface. Transport equations are given for the total particle flux and total energy flux crossing a closed toroidal surface. Also transport equations are given for the toroidal magnetic flux. A computer code was written to calculate the transport coefficients for a three species plasma, electrons and two species of ions. This is useful for calculating the transport coefficients of a plasma which contains impurities. It was found that the particle and energy transport coefficients are increased by a large amount, and the transport coefficients for the toroidal magnetic field are reduced by a small amount. For example, a deuterium plasma with 1.3 percent oxygen, one of the particle transport coefficients is increased by a factor of about four. The transport coefficients for the toroidal magnetic flux are reduced by about 20 percent. The increase in the particle transport coefficient is due to the collisional scattering of the deuterons by the heavy oxygen ions which is larger than the deuteron electron scattering, the normal process for particle transport in a two species plasma. The reduction in the toroidal magnetic flux transport coefficients are left unexplained

Storage, transportation and disposal system for used nuclear fuel assemblies

Science.gov (United States)

Scaglione, John M.; Wagner, John C.

2017-01-10

An integrated storage, transportation and disposal system for used fuel assemblies is provided. The system includes a plurality of sealed canisters and a cask sized to receive the sealed canisters in side by side relationship. The plurality of sealed canisters include an internal basket structure to receive a plurality of used fuel assemblies. The internal basket structure includes a plurality of radiation-absorbing panels and a plurality of hemispherical ribs generally perpendicular to the canister sidewall. The sealed canisters are received within the cask for storage and transportation and are removed from the cask for disposal at a designated repository. The system of the present invention allows the handling of sealed canisters separately or collectively, while allowing storage and transportation of high burnup fuel and damaged fuel to the designated repository.

Criticality safety evaluation for TWR-S fuel assembly transportation using TK-S16 containers

International Nuclear Information System (INIS)

Pesic, M.P.; Steljic, M.M.; Antic, D.P.

2002-01-01

Criticality safety issues, concerning transportation of fresh high-enriched uranium fuel elements (TWR-S fuel assembly type) with Russian containers TK-S16, are objects of study in this paper. Three-dimensional (3D) models of fuel element and container were made, based upon their well-known geometry and material structure. The way to pack fuel elements in a bundle inside of the container is proposed. Calculations were done by MCNP4B2 computer code. This Monte Carlo criticality code determined the effective multiplication factor from the cross-section data and specific geometry data. This evaluation demonstrated the subcriticality of a single package and an array of packages during normal conditions of transport and various hypothetical accident conditions. (author)

Assembly work and transport of JT-60SA cryostat base

International Nuclear Information System (INIS)

Okano, Fuminori; Masaki, Kei; Yagyu, Jun-ichi; Shibama, Yusuke; Sakasai, Akira; Miyo, Yasuhiko; Kaminaga, Atsushi; Nishiyama, Tomokazu; Suzuki, Sadaaki; Nakamura, Shigetoshi; Shibanuma, Kiyoshi

2013-11-01

Japan Atomic Energy Agency started to construct a fully superconducting tokamak experiment device, JT-60SA, to support the ITER since January, 2013 at the Fusion Research and Development Directorate in Naka, Japan. The JT-60SA will be constructed with enhancing the previous JT-60 infrastructures, in the JT-60 torus hall, where the ex-JT-60 machine was disassembled. The JT-60SA Cryostat Base, for base of the entire tokamak structure, were assembly as the first step of this construction. The Cryostat Base (CB, 250tons) is consists of 7 main components made of stainless steel, in 12 m diameter and 3 m height. The CB was built in the Spain and transported to the Naka site, via Hitachi port. After pre-assembly work including preliminary measurements and sole plate adjustments of its height/flatness, the JT-60SA CB was carefully set on the sole plate. JT-60SA CB was assembled with high accuracy by using a laser tracker. The CB was adjusted in the height and flatness against the assembly reference position and determined by the absolute coordinates. This report introduces the concrete result of assembly work and transport of JT-60SA CB. (author)

Packaging and transport case of test fuel assembly irradiated in the Creys-Malville reactor

International Nuclear Information System (INIS)

Geffroy, J.; Vivien, J.; Pouard, M.; Dujardin, G.N.; Veron, B.; Michoux, H.

1986-06-01

Some irradiated fuel assemblies from the fast neutron Creys Malville reactor will be sent to hot laboratories to follow fuel behavior. These test assemblies will be examined after a limited cooling time and transport is realized at high residual power (about 10kW) and cladding temperature should not rise over 500deg C. The fuel assemblies are not dismantled and transported into sodium. The assembly is placed into a case containing sodium plugged and put into a packaging. Dimensioning, thermal behavior, radiation protection and containment are examined [fr

Pulse superimposition calculational methodology for estimating the subcriticality level of nuclear fuel assemblies

International Nuclear Information System (INIS)

Talamo, Alberto; Gohar, Y.; Rabiti, C.; Aliberti, G.; Kondev, F.; Smith, D.; Zhong, Z.; Kiyavitskaya, H.; Bournos, V.; Fokov, Y.; Routkovskaya, C.; Serafimovich, I.

2009-01-01

One of the most reliable experimental methods for measuring the subcriticality level of a nuclear fuel assembly is the Sjoestrand method applied to the reaction rate generated from a pulsed neutron source. This study developed a new analytical methodology simulating the Sjoestrand method, which allows comparing the experimental and analytical reaction rates and the obtained subcriticality levels. In this methodology, the reaction rate is calculated due to a single neutron pulse using MCNP/MCNPX computer code or any other neutron transport code that explicitly simulates the delayed fission neutrons. The calculation simulates a single neutron pulse over a long time period until the delayed neutron contribution to the reaction rate is vanished. The obtained reaction rate is then superimposed to itself, with respect to the time, to simulate the repeated pulse operation until the asymptotic level of the reaction rate, set by the delayed neutrons, is achieved. The superimposition of the pulse to itself was calculated by a simple C computer program. A parallel version of the C program is used due to the large amount of data being processed, e.g. by the Message Passing Interface (MPI). The analytical results of this new calculation methodology have shown an excellent agreement with the experimental data available from the YALINA-Booster facility of Belarus. This methodology can be used to calculate Bell and Glasstone spatial correction factor.

Pulse superimposition calculational methodology for estimating the subcriticality level of nuclear fuel assemblies

Energy Technology Data Exchange (ETDEWEB)

Talamo, Alberto [Argonne National Laboratory, 9700 South Cass Avenue, Argonne, IL 60439 (United States)], E-mail: atalamo@anl.gov; Gohar, Y. [Argonne National Laboratory, 9700 South Cass Avenue, Argonne, IL 60439 (United States); Rabiti, C. [Idaho National Laboratory, P.O. Box 2528, Idaho Falls, ID 83403 (United States); Aliberti, G.; Kondev, F.; Smith, D.; Zhong, Z. [Argonne National Laboratory, 9700 South Cass Avenue, Argonne, IL 60439 (United States); Kiyavitskaya, H.; Bournos, V.; Fokov, Y.; Routkovskaya, C.; Serafimovich, I. [Joint Institute for Power and Nuclear Research-Sosny, National Academy of Sciences (Belarus)

2009-07-21

One of the most reliable experimental methods for measuring the subcriticality level of a nuclear fuel assembly is the Sjoestrand method applied to the reaction rate generated from a pulsed neutron source. This study developed a new analytical methodology simulating the Sjoestrand method, which allows comparing the experimental and analytical reaction rates and the obtained subcriticality levels. In this methodology, the reaction rate is calculated due to a single neutron pulse using MCNP/MCNPX computer code or any other neutron transport code that explicitly simulates the delayed fission neutrons. The calculation simulates a single neutron pulse over a long time period until the delayed neutron contribution to the reaction rate is vanished. The obtained reaction rate is then superimposed to itself, with respect to the time, to simulate the repeated pulse operation until the asymptotic level of the reaction rate, set by the delayed neutrons, is achieved. The superimposition of the pulse to itself was calculated by a simple C computer program. A parallel version of the C program is used due to the large amount of data being processed, e.g. by the Message Passing Interface (MPI). The analytical results of this new calculation methodology have shown an excellent agreement with the experimental data available from the YALINA-Booster facility of Belarus. This methodology can be used to calculate Bell and Glasstone spatial correction factor.

Fuel cell assembly with electrolyte transport

Science.gov (United States)

Chi, Chang V.

1983-01-01

A fuel cell assembly wherein electrolyte for filling the fuel cell matrix is carried via a transport system comprising a first passage means for conveying electrolyte through a first plate and communicating with a groove in a second plate at a first point, the first and second plates together sandwiching the matrix, and second passage means acting to carry electrolyte exclusively through the second plate and communicating with the groove at a second point exclusive of the first point.

Trial evaluation on criticality safety of the fuel assemblies at falling accident as spent fuel transport and storage cask

International Nuclear Information System (INIS)

Tadano, Tomoaki

2016-01-01

The authors conducted critical safety assessment on the supposed event at the time of a fall accident of cask, and examined the influence on criticality safety. If the spacer of fuel assembly is sound, it is assumed that the pitch of fuel rod interval changes, and if the spacer is broken, it is assumed that the fuel rod is unevenly distributed in the basket. For the critical calculation of fuel assembly basket system, they performed it using a calculation code. For both of the single cell and assembly, calculation results showed an increase in the effective multiplication factor of reactivity of 2-3%. When this reactivity is applied to the criticality analysis result of PWR fuel assembly, the value approaches to the limit 0.95 of the neutron effective multiplication factor keff. However, the keff when new fuel is loaded is sufficiently lower than 0.93. Therefore, it is unlikely that the criticality analysis result approaches to 0.95 at all burnups, and the possibility to become criticality is very low in actual spent fuel transport. When considering the reactivity of this research, it is possible that the design condition for the assumption of novel fuel loading becomes severer. Furthermore, criticality analysis under non-uniform pitch will become necessary, and criticality safety analysis for BWR fuel with heterogeneous enrichment degree and burnup degree will become also necessary. (A.O.)

Analytical model for calculation of the thermo hydraulic parameters in a fuel rod assembly

Energy Technology Data Exchange (ETDEWEB)

Cesna, B., E-mail: benas@mail.lei.l [Lithuanian Energy Institute, Laboratory of Nuclear Installation Safety, Breslaujos g. 3, LT-44403 Kaunas (Lithuania)

2010-11-15

Research highlights: {yields} Proposed calculation model can be used for rapid calculation of the bundles with rods spaced by wire wrapping or honey type spacer grids. {yields} Model estimate three flow cross mixture mechanisms. {yields} Program DARS is enable to analyses experimental results. - Abstract: The paper presents the procedure of the cellular calculation of thermo hydraulic parameters of a single-phase gas flow in a fuel rod assembly. The procedure is implemented in the DARS program. The program is intended for calculation of the distribution of the gaseous coolant parameters and wall temperatures in case of arbitrary, geometrically specified, arrangement of the rods in fuel assembly and in case of arbitrary, functionally specified in space, heat release in the rods. In mathematical model the flow cross-section of the channel of intricate shape is conventionally divided to elementary cells formed by straight lines, which connect the centers of rods. Within the limits of a single cell the coolant parameters and the temperature of the corresponding part of the rod surface are assumed constant. The entire fuel assembly is viewed as a system of parallel interconnected channels. Program DARS is illustrated by calculation of a temperature mode of 85-rod assembly with spacers of wire wrapping on the rods.

Electron stopping powers for transport calculations

International Nuclear Information System (INIS)

Berger, M.J.

1988-01-01

The reliability of radiation transport calculations depends on the accuracy of the input cross sections. Therefore, it is essential to review and update the cross sections from time to time. Even though the main interest of the author's group at NBS is in transport calculations and their applications, the group spends almost as much time on the analysis and preparation of cross sections as on the development of transport codes. Stopping powers, photon attenuation coefficients, bremsstrahlung cross sections, and elastic-scattering cross sections in recent years have claimed attention. This chapter deals with electron stopping powers (with emphasis on collision stopping powers), and reviews the state of the art as reflected by Report 37 of the International Commission on Radiation Units and Measurements

Three dimensions transport calculations for PWR core

International Nuclear Information System (INIS)

Richebois, E.

2000-01-01

The objective of this work is to define improved 3-D core calculation methods based on the transport theory. These methods can be particularly useful and lead to more precise computations in areas of the core where anisotropy and steep flux gradients occur, especially near interface and boundary conditions and in regions of high heterogeneity (bundle with absorbent rods). In order to apply the transport theory a new method for calculating reflector constants has been developed, since traditional methods were only suited for 2-group diffusion core calculations and could not be extrapolated to transport calculations. In this thesis work, the new method for obtaining reflector constants is derived regardless of the number of energy groups and of the operator used. The core calculations results using the reflector constants thereof obtained have been validated on the EDF's power reactor Saint Laurent B1 with MOX loading. The advantages of a 3-D core transport calculation scheme have been highlighted as opposed to diffusion methods; there are a considerable number of significant effects and potential advantages to be gained in rod worth calculations for instance. These preliminary results obtained with on particular cycle will have to be confirmed by more systematic analysis. Accidents like MSLB (main steam line break) and LOCA (loss of coolant accident) should also be investigated and constitute challenging situations where anisotropy is high and/or flux gradients are steep. This method is now being validated for others EDF's PWRs' reactors, as well as for experimental reactors and other types of commercial reactors. (author)

Two-group k-eigenvalue benchmark calculations for planar geometry transport in a binary stochastic medium

International Nuclear Information System (INIS)

Davis, I.M.; Palmer, T.S.

2005-01-01

Benchmark calculations are performed for neutron transport in a two material (binary) stochastic multiplying medium. Spatial, angular, and energy dependence are included. The problem considered is based on a fuel assembly of a common pressurized water reactor. The mean chord length through the assembly is determined and used as the planar geometry system length. According to assumed or calculated material distributions, this system length is populated with alternating fuel and moderator segments of random size. Neutron flux distributions are numerically computed using a discretized form of the Boltzmann transport equation employing diffusion synthetic acceleration. Average quantities (group fluxes and k-eigenvalue) and variances are calculated from an ensemble of realizations of the mixing statistics. The effects of varying two parameters in the fuel, two different boundary conditions, and three different sets of mixing statistics are assessed. A probability distribution function (PDF) of the k-eigenvalue is generated and compared with previous research. Atomic mix solutions are compared with these benchmark ensemble average flux and k-eigenvalue solutions. Mixing statistics with large standard deviations give the most widely varying ensemble solutions of the flux and k-eigenvalue. The shape of the k-eigenvalue PDF qualitatively agrees with previous work. Its overall shape is independent of variations in fuel cross-sections for the problems considered, but its width is impacted by these variations. Statistical distributions with smaller standard deviations alter the shape of this PDF toward a normal distribution. The atomic mix approximation yields large over-predictions of the ensemble average k-eigenvalue and under-predictions of the flux. Qualitatively correct flux shapes are obtained in some cases. These benchmark calculations indicate that a model which includes higher statistical moments of the mixing statistics is needed for accurate predictions of binary

Radiation transport calculation methods in BNCT

International Nuclear Information System (INIS)

Koivunoro, H.; Seppaelae, T.; Savolainen, S.

2000-01-01

Boron neutron capture therapy (BNCT) is used as a radiotherapy for malignant brain tumours. Radiation dose distribution is necessary to determine individually for each patient. Radiation transport and dose distribution calculations in BNCT are more complicated than in conventional radiotherapy. Total dose in BNCT consists of several different dose components. The most important dose component for tumour control is therapeutic boron dose D B . The other dose components are gamma dose D g , incident fast neutron dose D f ast n and nitrogen dose D N . Total dose is a weighted sum of the dose components. Calculation of neutron and photon flux is a complex problem and requires numerical methods, i.e. deterministic or stochastic simulation methods. Deterministic methods are based on the numerical solution of Boltzmann transport equation. Such are discrete ordinates (SN) and spherical harmonics (PN) methods. The stochastic simulation method for calculation of radiation transport is known as Monte Carlo method. In the deterministic methods the spatial geometry is partitioned into mesh elements. In SN method angular integrals of the transport equation are replaced with weighted sums over a set of discrete angular directions. Flux is calculated iteratively for all these mesh elements and for each discrete direction. Discrete ordinates transport codes used in the dosimetric calculations are ANISN, DORT and TORT. In PN method a Legendre expansion for angular flux is used instead of discrete direction fluxes, land the angular dependency comes a property of vector function space itself. Thus, only spatial iterations are required for resulting equations. A novel radiation transport code based on PN method and tree-multigrid technique (TMG) has been developed at VTT (Technical Research Centre of Finland). Monte Carlo method solves the radiation transport by randomly selecting neutrons and photons from a prespecified boundary source and following the histories of selected particles

RECOVERY ACT - Thylakoid Assembly and Folded Protein Transport by the Tat Pathway

Energy Technology Data Exchange (ETDEWEB)

Dabney-Smith, Carole [Miami Univ., Oxford, OH (United States)

2016-07-18

Assembly of functional photosystems complete with necessary intrinsic (membrane-bound) and extrinsic proteins requires the function of at least 3 protein transport pathways in thylakoid membranes. Our research focuses on one of those pathways, a unique and essential protein transport pathway found in the chloroplasts of plants, bacteria, and some archaebacteria, the Twin arginine translocation (Tat) system. The chloroplast Tat (cpTat) system is thought to be responsible for the proper location of ~50% of thylakoid lumen proteins, several of which are necessary for proper photosystem assembly, maintenance, and function. Specifically, cpTat systems are unique because they transport fully folded and assembled proteins across ion tight membranes using only three membrane components, Tha4, Hcf106, and cpTatC, and the protonmotive force generated by photosynthesis. Despite the importance of the cpTat system in plants, the mechanism of transport of a folded precursor is not well known. Our long-term goal is to investigate the role protein transport systems have on organelle biogenesis, particularly the assembly of membrane protein complexes in thylakoids of chloroplasts. The objective of this proposal is to correlate structural changes in the membrane-bound cpTat component, Tha4, to the mechanism of translocation of folded-precursor substrates across the membrane bilayer by using a cysteine accessibility and crosslinking approach. Our central hypothesis is that the precursor passes through a proteinaceous pore of assembled Tha4 protomers that have undergone a conformational or topological change in response to transport. This research is predicated upon the observations that Tha4 exists in molar excess in the membrane relative to the other cpTat components; its regulated assembly to the precursor-bound receptor; and our data showing oligomerization of Tha4 into very large complexes in response to transport. Our rationale for these studies is that understanding cp

Two-level MOC calculation scheme in APOLLO2 for cross-section library generation for LWR hexagonal assemblies

International Nuclear Information System (INIS)

Petrov, Nikolay; Todorova, Galina; Kolev, Nikola; Damian, Frederic

2011-01-01

The accurate and efficient MOC calculation scheme in APOLLO2, developed by CEA for generating multi-parameterized cross-section libraries for PWR assemblies, has been adapted to hexagonal assemblies. The neutronic part of this scheme is based on a two-level calculation methodology. At the first level, a multi-cell method is used in 281 energy groups for cross-section definition and self-shielding. At the second level, precise MOC calculations are performed in a collapsed energy mesh (30-40 groups). In this paper, the application and validation of the two-level scheme for hexagonal assemblies is described. Solutions for a VVER assembly are compared with TRIPOLI4® calculations and direct 281g MOC solutions. The results show that the accuracy is close to that of the 281g MOC calculation while the CPU time is substantially reduced. Compared to the multi-cell method, the accuracy is markedly improved. (author)

OECD/NEA burnup credit criticality benchmarks phase IIIA: Criticality calculations of BWR spent fuel assemblies in storage and transport

Energy Technology Data Exchange (ETDEWEB)

Okuno, Hiroshi; Naito, Yoshitaka [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Ando, Yoshihira [Toshiba Corp., Kawasaki, Kanagawa (Japan)

2000-09-01

The report describes the final results of Phase IIIA Benchmarks conducted by the Burnup Credit Criticality Calculation Working Group under the auspices of the Nuclear Energy Agency of the Organization for Economic Cooperation and Development (OECD/NEA). The benchmarks are intended to confirm the predictive capability of the current computer code and data library combinations for the neutron multiplication factor (k{sub eff}) of a layer of irradiated BWR fuel assembly array model. In total 22 benchmark problems are proposed for calculations of k{sub eff}. The effects of following parameters are investigated: cooling time, inclusion/exclusion of FP nuclides and axial burnup profile, and inclusion of axial profile of void fraction or constant void fractions during burnup. Axial profiles of fractional fission rates are further requested for five cases out of the 22 problems. Twenty-one sets of results are presented, contributed by 17 institutes from 9 countries. The relative dispersion of k{sub eff} values calculated by the participants from the mean value is almost within the band of {+-}1%{delta}k/k. The deviations from the averaged calculated fission rate profiles are found to be within {+-}5% for most cases. (author)

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

Energy Technology Data Exchange (ETDEWEB)

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

2006-07-01

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

Development of Monte Carlo decay gamma-ray transport calculation system

Energy Technology Data Exchange (ETDEWEB)

Sato, Satoshi [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment; Kawasaki, Nobuo [Fujitsu Ltd., Tokyo (Japan); Kume, Etsuo [Japan Atomic Energy Research Inst., Center for Promotion of Computational Science and Engineering, Tokai, Ibaraki (Japan)

2001-06-01

In the DT fusion reactor, it is critical concern to evaluate the decay gamma-ray biological dose rates after the reactor shutdown exactly. In order to evaluate the decay gamma-ray biological dose rates exactly, three dimensional Monte Carlo decay gamma-ray transport calculation system have been developed by connecting the three dimensional Monte Carlo particle transport calculation code and the induced activity calculation code. The developed calculation system consists of the following four functions. (1) The operational neutron flux distribution is calculated by the three dimensional Monte Carlo particle transport calculation code. (2) The induced activities are calculated by the induced activity calculation code. (3) The decay gamma-ray source distribution is obtained from the induced activities. (4) The decay gamma-rays are generated by using the decay gamma-ray source distribution, and the decay gamma-ray transport calculation is conducted by the three dimensional Monte Carlo particle transport calculation code. In order to reduce the calculation time drastically, a biasing system for the decay gamma-ray source distribution has been developed, and the function is also included in the present system. In this paper, the outline and the detail of the system, and the execution example are reported. The evaluation for the effect of the biasing system is also reported. (author)

DRAGON and CORD-2 nuclear calculation of the NPP Krško fuel assembly

International Nuclear Information System (INIS)

Kromar, Marjan; Kurinčič, Bojan

2012-01-01

Highlights: ▶ Comparison of the DRAGON 4 and CORD-2 calculation of the NPP Krško 16 × 16 fuel assembly has been performed. ▶ Two different enrichments (4.60% and 4.95%), three IFBA configurations (0, 64 and 116 IFBAs) and the burnup up to 60,000 MWd/tU were considered. ▶ Comparison shows that the agreement in the results of both codes is very good. ▶ Regarding efficiency, one CORD-2 assembly depletion calculation takes about 3 min of the CPU time on the typical PC, while the DRAGON calculation runs more than 25 h. - Abstract: The geometry of the reactor core is usually too complex to be solved in one step. Therefore, a solution for the whole core in 3-D geometry is sought in several steps, where some kind of homogenization procedure of neutron few-group cross sections is applied. Usually, assembly-homogenized effective two-group cross sections are determined, which are suitable for solving the diffusion equation for the whole core by a coarse mesh nodal methods. In this paper DRAGON 4 and CORD-2 codes are used for the calculation of NPP Krško 16 × 16 fuel assemblies without and with IFBA rods. DRAGON code was selected, since it can use the same cross-section library as the WIMS-D5 code employed in the CORD-2 system. Different results arise therefore solely from the different models used in the calculations. The heterogeneous depletion calculation was performed up to burnup of 60,000 MWd/tU. Results of both codes are compared for the infinite multiplication factor, fast to thermal spectral ratio and pin power distributions.

Lateral hydraulic forces calculation on PWR fuel assemblies with computational fluid dynamics codes

International Nuclear Information System (INIS)

Corpa Masa, R.; Jimenez Varas, G.; Moreno Garcia, B.

2016-01-01

To be able to simulate the behavior of nuclear fuel under operating conditions, it is required to include all the representative loads, including the lateral hydraulic forces which were not included traditionally because of the difficulty of calculating them in a reliable way. Thanks to the advance in CFD codes, now it is possible to assess them. This study calculates the local lateral hydraulic forces, caused by the contraction and expansion of the flow due to the bow of the surrounding fuel assemblies, on of fuel assembly under typical operating conditions from a three loop Westinghouse PWR reactor. (Author)

Comparison of calculated and experimental results of fuel bunching in VERA assemblies 1B, 3A and 5A

International Nuclear Information System (INIS)

Matthews, J.D.

1964-12-01

Comparison is made of experimental and calculated reactivity changes when the fuel plates of the VERA assemblies are bunched from 1/8 in. thickness to 1/2 in. In addition, U238 fine structure reaction rates are studied. The method used is that investigated in detail by James and Matthews ('A Perturbation Method for Multigroup Neutron Transport Calculations in Plane Lattices', AEEW - R 219), while the multigroup data sets used are both those of Yiftah, Okrent and Moldauer, and Roach. The method and data give only rough agreement with experimental reactivity changes, with a tendency to over-predict the results. U238 fission rates are in better agreement, though the method predicts too small a variation for the U238 capture rate across a cell. (author)

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

International Nuclear Information System (INIS)

Wilson, Michael L.

2001-01-01

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

Calculation of local characteristics of velocity field in turbulent coolant flow in fast reactor fuel assembly

International Nuclear Information System (INIS)

Muehlbauer, P.

1981-08-01

Experience is described gained with the application of computer code VELASCO in calculating the velocity field in fast reactor fuel assemblies taking into account configuration disturbances due to fuel pin displacement. Theoretical results are compared with the results of experiments conducted by UJV on aerodynamic models HEM-1 (model of the fuel assembly central part) and HEM-2 (model of the fuel assembly peripheral part). The results are reported of calculating the distribution of shear stress in wetted rod surfaces and in the assembly wall (model HEM-2) and the corresponding experimental results are shown. The shear stress distribution in wetted surfaces obtained using the VELASCO code allowed forming an opinion on the code capability of comprising local parameters of turbulent flow through a fuel rod bundle. The applicability was also tested of the code for calculating mean velocities in the individual zones, eg., in elementary cells. (B.S.)

Surrogate fuel assembly multi-axis shaker tests to simulate normal conditions of rail and truck transport

Energy Technology Data Exchange (ETDEWEB)

McConnell, Paul E. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Koenig, Greg John [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Uncapher, William Leonard [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Grey, Carissa [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Engelhardt, Charles [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Saltzstein, Sylvia J. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Sorenson, Ken B. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

2016-05-01

This report describes the third set of tests (the “DCLa shaker tests”) of an instrumented surrogate PWR fuel assembly. The purpose of this set of tests was to measure strains and accelerations on Zircaloy-4 fuel rods when the PWR assembly was subjected to rail and truck loadings simulating normal conditions of transport when affixed to a multi-axis shaker. This is the first set of tests of the assembly simulating rail normal conditions of transport.

Statistics of Monte Carlo methods used in radiation transport calculation

International Nuclear Information System (INIS)

Datta, D.

2009-01-01

Radiation transport calculation can be carried out by using either deterministic or statistical methods. Radiation transport calculation based on statistical methods is basic theme of the Monte Carlo methods. The aim of this lecture is to describe the fundamental statistics required to build the foundations of Monte Carlo technique for radiation transport calculation. Lecture note is organized in the following way. Section (1) will describe the introduction of Basic Monte Carlo and its classification towards the respective field. Section (2) will describe the random sampling methods, a key component of Monte Carlo radiation transport calculation, Section (3) will provide the statistical uncertainty of Monte Carlo estimates, Section (4) will describe in brief the importance of variance reduction techniques while sampling particles such as photon, or neutron in the process of radiation transport

Surrogate fuel assembly multi-axis shaker tests to simulate normal conditions of rail and truck transport

Energy Technology Data Exchange (ETDEWEB)

McConnell, Paul E. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Koenig, Greg John [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Uncapher, William Leonard [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Grey, Carissa [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Engelhardt, Charles [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Saltzstein, Sylvia J. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Sorenson, Ken B. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

2016-05-12

This report describes the third set of tests (the “DCL^a shaker tests”) of an instrumented surrogate PWR fuel assembly. The purpose of this set of tests was to measure strains and accelerations on Zircaloy-4 fuel rods when the PWR assembly was subjected to rail and truck loadings simulating normal conditions of transport when affixed to a multi-axis shaker. This is the first set of tests of the assembly simulating rail normal conditions of transport.

Effect of Structure and Disorder on the Charge Transport in Defined Self-Assembled Monolayers of Organic Semiconductors.

Science.gov (United States)

Schmaltz, Thomas; Gothe, Bastian; Krause, Andreas; Leitherer, Susanne; Steinrück, Hans-Georg; Thoss, Michael; Clark, Timothy; Halik, Marcus

2017-09-26

Self-assembled monolayer field-effect transistors (SAMFETs) are not only a promising type of organic electronic device but also allow detailed analyses of structure-property correlations. The influence of the morphology on the charge transport is particularly pronounced, due to the confined monolayer of 2D-π-stacked organic semiconductor molecules. The morphology, in turn, is governed by relatively weak van-der-Waals interactions and is thus prone to dynamic structural fluctuations. Accordingly, combining electronic and physical characterization and time-averaged X-ray analyses with the dynamic information available at atomic resolution from simulations allows us to characterize self-assembled monolayer (SAM) based devices in great detail. For this purpose, we have constructed transistors based on SAMs of two molecules that consist of the organic p-type semiconductor benzothieno[3,2-b][1]benzothiophene (BTBT), linked to a C 11 or C 12 alkylphosphonic acid. Both molecules form ordered SAMs; however, our experiments show that the size of the crystalline domains and the charge-transport properties vary considerably in the two systems. These findings were confirmed by molecular dynamics (MD) simulations and semiempirical molecular-orbital electronic-structure calculations, performed on snapshots from the MD simulations at different times, revealing, in atomistic detail, how the charge transport in organic semiconductors is influenced and limited by dynamic disorder.

Sn transport calculations on vector and parallel processors

International Nuclear Information System (INIS)

Rhoades, W.A.; Childs, R.L.

1987-01-01

The transport of radiation from the source to the location of people or equipment gives rise to some of the most challenging of calculations. A problem may involve as many as a billion unknowns, each evaluated several times to resolve interdependence. Such calculations run many hours on a Cray computer, and a typical study involves many such calculations. This paper will discuss the steps taken to vectorize the DOT code, which solves transport problems in two space dimensions (2-D); the extension of this code to 3-D; and the plans for extension to parallel processors

Transport equivalent diffusion constants for reflector region in PWRs

International Nuclear Information System (INIS)

Tahara, Yoshihisa; Sekimoto, Hiroshi

2002-01-01

The diffusion-theory-based nodal method is widely used in PWR core designs for reason of its high computing speed in three-dimensional calculations. The baffle/reflector (B/R) constants used in nodal calculations are usually calculated based on a one-dimensional transport calculation. However, to achieve high accuracy of assembly power prediction, two-dimensional model is needed. For this reason, the method for calculating transport equivalent diffusion constants of reflector material was developed so that the neutron currents on the material boundaries could be calculated exactly in diffusion calculations. Two-dimensional B/R constants were calculated using the transport equivalent diffusion constants in the two-dimensional diffusion calculation whose geometry reflected the actual material configuration in the reflector region. The two-dimensional B/R constants enabled us to predict assembly power within an error of 1.5% at hot full power conditions. (author)

Capacity analysis of automatic transport systems in an assembly factory

NARCIS (Netherlands)

Zijm, W.H.M.; Lenstra, J.K.; Tijms, H.C.; Volgenant, A.

1989-01-01

We describe a case study concerning the capacity analysis of a completely automated transport system in a flexible assembly environment. Basically, the system is modelled as a network of queues, however, due to its complex nature, product-form network theory is not applicable. Instead, we present an

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

International Nuclear Information System (INIS)

Leroyer, H.; Brosselard, C.; Girardi, E.

2012-01-01

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)

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

Energy Technology Data Exchange (ETDEWEB)

Kenneth D. Wright

1997-09-03

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

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

International Nuclear Information System (INIS)

Wright, Kenneth D.

1997-01-01

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

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

International Nuclear Information System (INIS)

Wright, Kenneth D.

1997-01-01

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

Discrete-ordinates electron transport calculations using standard neutron transport codes

International Nuclear Information System (INIS)

Morel, J.E.

1979-01-01

The primary purpose of this work was to develop a method for using standard neutron transport codes to perform electron transport calculations. The method is to develop approximate electron cross sections which are sufficiently well-behaved to be treated with standard S/sub n/ methods, but which nonetheless yield flux solutions which are very similar to the exact solutions. The main advantage of this approach is that, once the approximate cross sections are constructed, their multigroup Legendre expansion coefficients can be calculated and input to any standard S/sub n/ code. Discrete-ordinates calculations were performed to determine the accuracy of the flux solutions for problems corresponding to 1.0-MeV electrons incident upon slabs of aluminum and gold. All S/sub n/ calculations were compared with similar calculations performed with an electron Monte Carlo code, considered to be exact. In all cases, the discrete-ordinates solutions for integral flux quantities (i.e., scalar flux, energy deposition profiles, etc.) are generally in agreement with the Monte Carlo solutions to within approximately 5% or less. The central conclusion is that integral electron flux quantities can be efficiently and accurately calculated using standard S/sub n/ codes in conjunction with approximate cross sections. Furthermore, if group structures and approximate cross section construction are optimized, accurate differential flux energy spectra may also be obtainable without having to use an inordinately large number of energy groups. 1 figure

OECD/NEA burnup credit criticality benchmarks phase IIIB: Burnup calculations of BWR fuel assemblies for storage and transport

International Nuclear Information System (INIS)

Okuno, Hiroshi; Naito, Yoshitaka; Suyama, Kenya

2002-02-01

The report describes the final results of the Phase IIIB Benchmark conducted by the Expert Group on Burnup Credit Criticality Safety under the auspices of the Nuclear Energy Agency (NEA) of the Organization for Economic Cooperation and Development (OECD). The Benchmark was intended to compare the predictability of current computer code and data library combinations for the atomic number densities of an irradiated PWR fuel assembly model. The fuel assembly was irradiated under specific power of 25.6 MW/tHM up to 40 GWd/tHM and cooled for five years. The void fraction was assumed to be uniform throughout the channel box and constant, at 0, 40 and 70%, during burnup. In total, 16 results were submitted from 13 institutes of 7 countries. The calculated atomic number densities of 12 actinides and 20 fission product nuclides were found to be for the most part within a range of ±10% relative to the average, although some results, esp. 155 Eu and gadolinium isotopes, exceeded the band, which will require further investigation. Pin-wise burnup results agreed well among the participants. The results in the infinite neutron multiplication factor k ∞ also accorded well with each other for void fractions of 0 and 40%; however some results deviated from the averaged value noticeably for the void fraction of 70%. (author)

OECD/NEA burnup credit criticality benchmarks phase IIIB. Burnup calculations of BWR fuel assemblies for storage and transport

Energy Technology Data Exchange (ETDEWEB)

Okuno, Hiroshi; Naito, Yoshitaka; Suyama, Kenya [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

2002-02-01

The report describes the final results of the Phase IIIB Benchmark conducted by the Expert Group on Burnup Credit Criticality Safety under the auspices of the Nuclear Energy Agency (NEA) of the Organization for Economic Cooperation and Development (OECD). The Benchmark was intended to compare the predictability of current computer code and data library combinations for the atomic number densities of an irradiated PWR fuel assembly model. The fuel assembly was irradiated under specific power of 25.6 MW/tHM up to 40 GWd/tHM and cooled for five years. The void fraction was assumed to be uniform throughout the channel box and constant, at 0, 40 and 70%, during burnup. In total, 16 results were submitted from 13 institutes of 7 countries. The calculated atomic number densities of 12 actinides and 20 fission product nuclides were found to be for the most part within a range of {+-}10% relative to the average, although some results, esp. {sup 155}Eu and gadolinium isotopes, exceeded the band, which will require further investigation. Pin-wise burnup results agreed well among the participants. The results in the infinite neutron multiplication factor k{sub {infinity}} also accorded well with each other for void fractions of 0 and 40%; however some results deviated from the averaged value noticeably for the void fraction of 70%. (author)

Determination of uncertainties in the calculation of dose rates at transport and storage casks; Unsicherheiten bei der Berechnung von Dosisleistungen an Transport- und Lagerbehaeltern

Energy Technology Data Exchange (ETDEWEB)

Schloemer, Luc Laurent Alexander

2014-12-17

The compliance with the dose rate limits for transport and storage casks (TLB) for spent nuclear fuel from pressurised water reactors can be proved by calculation. This includes the determination of the radioactive sources and the shielding-capability of the cask. In this thesis the entire computational chain, which extends from the determination of the source terms to the final Monte-Carlo-transport-calculation is analysed and the arising uncertainties are quantified not only by benchmarks but also by variational calculi. The background of these analyses is that the comparison with measured dose rates at different TLBs shows an overestimation by the values calculated. Regarding the studies performed, the overestimation can be mainly explained by the detector characteristics for the measurement of the neutron dose rate and additionally in case of the gamma dose rates by the energy group structure, which the calculation is based on. It turns out that the consideration of the uncertainties occurring along the computational chain can lead to even greater overestimation. Concerning the dose rate calculation at cask loadings with spent uranium fuel assemblies an uncertainty of (({sup +21}{sub -28}) ±2) % (rel.) for the total gamma dose rate and of ({sup +28±23}{sub -55±4}) % (rel.) for the total neutron dose rate are estimated. For mixed-loadings with spent uranium and MOX fuel assemblies an uncertainty of ({sup +24±3}{sub -27±2}) % (rel.) for the total gamma dose rate and of ({sup +28±23}{sub -55±4}) % (rel.) for the total neutron dose rate are quantified. The results show that the computational chain has not to be modified, because the calculations performed lead to conservative dose rate predictions, even if high uncertainties at neutron dose rate measurements arise. Thus at first the uncertainties of the neutron dose rate measurement have to be decreased to enable a reduction of the overestimation of the calculated dose rate afterwards. In the present thesis

Optimization of the Spent Fuel Attribute Tester using radiation transport calculations

International Nuclear Information System (INIS)

Laub, T.W.; Dupree, S.A.; Arlt, R.

1993-01-01

The International Atomic Energy Agency uses the Spent Fuel Attribute Tester (SFAT) to measure gamma signatures from fuel assemblies stored in spent fuel pools. It consists of a shielded, collimated NaI(Tl) detector attached to an air-filled pipe. The purpose of the present study was to define design changes, within operational constraints, that would improve the target assembly 137 Cs signal relative to the background signals from adjacent assemblies. This improvement is essential to reducing to an acceptable level the measurement time during an inspection. Monte Carlo calculations of the entire geometry were impractical, therefore, a hybrid method was developed that combined one-dimensional discrete ordinates models of the spent fuel pool, three-dimensional Monte Carlo calculations of the SFAT, and detector response calculations. The method compared well with measurements taken with the existing baseline SFAT. Calculations predicted significant improvements in signal-to-noise ratio. Recommended changes included shortening the pipe and increasing its wall thickness, placing low-Z filters in the crystal line of sight, reducing the thickness of shielding around the collimator aperture and adding shielding around the crystal, and reducing the diameter of the crystal. An instrument incorporating these design changes is being fabricated in Finland and will be tested this year

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

Energy Technology Data Exchange (ETDEWEB)

Kenneth D. Wright

1997-07-30

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

Charged-particle calculations using Boltzmann transport methods

International Nuclear Information System (INIS)

Hoffman, T.J.; Dodds, H.L. Jr.; Robinson, M.T.; Holmes, D.K.

1981-01-01

Several aspects of radiation damage effects in fusion reactor neutron and ion irradiation environments are amenable to treatment by transport theory methods. In this paper, multigroup transport techniques are developed for the calculation of charged particle range distributions, reflection coefficients, and sputtering yields. The Boltzmann transport approach can be implemented, with minor changes, in standard neutral particle computer codes. With the multigroup discrete ordinates code, ANISN, determination of ion and target atom distributions as functions of position, energy, and direction can be obtained without the stochastic error associated with atomistic computer codes such as MARLOWE and TRIM. With the multigroup Monte Carlo code, MORSE, charged particle effects can be obtained for problems associated with very complex geometries. Results are presented for several charged particle problems. Good agreement is obtained between quantities calculated with the multigroup approach and those obtained experimentally or by atomistic computer codes

Reference calculations on critical assemblies with Apollo2 code working with a fine multigroup mesh

International Nuclear Information System (INIS)

Aggery, A.

1999-12-01

The objective of this thesis is to add to the multigroup transport code APOLLO2 the capability to perform deterministic reference calculations, for any type of reactor, using a very fine energy mesh of several thousand groups. This new reference tool allows us to validate the self-shielding model used in industrial applications, to perform depletion calculations, differential effects calculations, critical buckling calculations or to evaluate precisely data required by the self shielding model. At its origin, APOLLO2 was designed to perform routine calculations with energy meshes around one hundred groups. That is why, in the current format of cross sections libraries, almost each value of the multigroup energy transfer matrix is stored. As this format is not convenient for a high number of groups (concerning memory size), we had to search out a new format for removal matrices and consequently to modify the code. In the new format we found, only some values of removal matrices are kept (these values depend on a reconstruction precision choice), the other ones being reconstructed by a linear interpolation, what reduces the size of these matrices. Then we had to show that APOLLO2 working with a fine multigroup mesh had the capability to perform reference calculations on any assembly geometry. For that, we successfully carried out the validation with several calculations for which we compared APOLLO2 results (obtained with the universal mesh of 11276 groups) to results obtained with Monte Carlo codes (MCNP, TRIPOLI4). Physical analysis led with this new tool have been very fruitful and show a great potential for such an R and D tool. (author)

Multiscale modeling of transport of grains through granular assemblies

Directory of Open Access Journals (Sweden)

Tejada Ignacio G

2017-01-01

Full Text Available We investigate the transport of moderately large passive particles through granular assemblies caused by seeping flows. This process can only be described by highly nonlinear continuum models, since the local permeability, the advection and dispersion mechanisms are strongly determined by the concentration of transported particles. Particles may sometimes get temporally trapped and thus proper kinetic mass transfer models are required. The mass transfer depends on the complexity of the porous medium, the kind of interaction forces and the concentration of transported particles. We study these two issues by means of numerical and laboratory experiments. In the laboratory we use an oedo-permeameter to force sand grains to move through a gravel bed under conditions of constant hydraulic pressure drop. To understand the process, numerical experiments were performed to approach particle transport at the grain scale with a fully coupled method. The DEM-PFV combines the discrete element method with a pore scale finite volume formulation to solve the interstitial fluid flow and particle transport problems. These experiments help us to set up a continuum transport model that can be used in a boundary value problem.

Refinements to temperature calculations of spent fuel assemblies when in a stagnant gas environment

International Nuclear Information System (INIS)

Rhodes, C.A.; Haire, M.J.

1984-01-01

Undesirably high temperatures are possible in irradiated fuel assemblies because of the radioactive decay of fission products formed while in the reactor. The COXPRO computer code has been used for some time to calculate temperatures in spent fuel when the fuel is suspended in a stagnant gas environment. This code assumed radiation to be the only mode of heat dissipation within the fuel pin bundle. Refinements have been made to include conduction as well as radiation heat transfer within this code. Comparison of calculated and measured temperatures in four separate and independent tests indicate that maximum fuel assembly temperatures can be predicted to within about 6%. 2 references, 5 figures

Neutronic calculations with transport and diffusion computer codes for light water moderated critical with UO2 enriched at 4,75% as fuel

International Nuclear Information System (INIS)

Sabundjian, G.; Nakata, H.

1983-02-01

The neutronic calculational procedure in a 4,75% w/O enriched UO 2 fueled light water moderated critical assembly was tested, using the transport codes and diffusin code available at the Instituto de Pesquisas Energeticas e Nucleares. The results of the tested codes, LEOPARD, CITHAMMER, LASER, GELS and CITATION, were found to be satisfatory and only a slight advantage is presented by CITHAMMER code. (Author) [pt

Critical and shielding parametric studies with the Monte Carlo code TRIPOLI to identify the key points to take into account during the transportation of blanket assemblies with high ratio of americium

International Nuclear Information System (INIS)

Gosmain, Cecile-Aline

2011-01-01

In the framework of French research program on Generation IV sodium cooled fast reactor, one possible option consists in burning minor actinides in this kind of Advanced Sodium Technological Reactor. Two types of transmutation mode are studied in the world : the homogeneous mode of transmutation where actinides are scattered with very low enrichment ratio in fissile assemblies and the heterogeneous mode where fissile core is surrounded by blanket assemblies filled with minor actinides with ratio of incorporated actinides up to 20%. Depending on which element is considered to be burnt and on its content, these minor actinides contents imply constraints on assemblies' transportation between Nuclear Power Plants and fuel cycle facilities. In this study, we present some academic studies in order to identify some key constraints linked to the residual power and neutron/gamma load of such kind of blanket assemblies. To simplify the approach, we considered a modeling of a 'model cask' dedicated to the transportation of a unique irradiated blanket assembly loaded with 20% of Americium and basically inspired from an existent cask designed initially for the damaged fissile Superphenix assembly transport. Thermal calculations performed with EDF-SYRTHES code have shown that due to thermal limitations on cladding temperature, the decay time to be considered before transportation is 20 years. This study is based on explicit 3D representations of the cask and the contained blanket assembly with the Monte Carlo code TRIPOLI/JEFF3.1.1 library and concludes that after such a decay time, the transportation of a unique Americium radial blanket is feasible only if the design of our model cask is modified in order to comply with the dose limitation criterion. (author)

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

International Nuclear Information System (INIS)

Wright, Kenneth D.

1997-01-01

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

Calculation of the local power peaking near WWER-440 control assemblies with Hf plates

International Nuclear Information System (INIS)

Hegyi, Gy.; Hordosy, G.; Kereszturi, A.; Maraszy, Cs.; Temesvari, E.

2003-01-01

The original coupler design of the WWER-440 assemblies had the following well known deficiency: The relatively large amount of water in the coupler between the absorber and fuel port of the control assembly can cause undesirably sharp power peaking in the fuel rods next to the coupler. The power peaking can be especially high after control rod withdrawal when the coupler reached low burnup level region of the adjacent assembly. The modernized coupler design overcomes the original problem by applying a thin Hf plate in the critical region. The very complicated structure of the coupler requires the verification of the core design methods by high precision 3D Monte Carlo calculations. The paper presents an MCNP reference calculation on the control rod coupler benchmark with Hf absorber plates. The benchmark solution with the KARATE-440 code system is also presented. The need for treating the Hf burnout in the reflector region is investigated (Authors)

The structure of the COPII transport-vesicle coat assembled on membranes.

Science.gov (United States)

Zanetti, Giulia; Prinz, Simone; Daum, Sebastian; Meister, Annette; Schekman, Randy; Bacia, Kirsten; Briggs, John A G

2013-09-17

Coat protein complex II (COPII) mediates formation of the membrane vesicles that export newly synthesised proteins from the endoplasmic reticulum. The inner COPII proteins bind to cargo and membrane, linking them to the outer COPII components that form a cage around the vesicle. Regulated flexibility in coat architecture is essential for transport of a variety of differently sized cargoes, but structural data on the assembled coat has not been available. We have used cryo-electron tomography and subtomogram averaging to determine the structure of the complete, membrane-assembled COPII coat. We describe a novel arrangement of the outer coat and find that the inner coat can assemble into regular lattices. The data reveal how coat subunits interact with one another and with the membrane, suggesting how coordinated assembly of inner and outer coats can mediate and regulate packaging of vesicles ranging from small spheres to large tubular carriers. DOI:http://dx.doi.org/10.7554/eLife.00951.001.

Calculation and analysis for a series of enriched uranium bare sphere critical assemblies

International Nuclear Information System (INIS)

Yang Shunhai

1994-12-01

The imported reactor fuel assembly MARIA program system is adapted to CYBER 825 computer in China Institute of Atomic Energy, and extensively used for a series of enriched uranium bare sphere critical assemblies. The MARIA auxiliary program of resonance modification MA is designed for taking account of the effects of resonance fission and absorption on calculated results. By which, the multigroup constants in the library attached to MARIA program are revised based on the U.S. Evaluated Nuclear Data File ENDF/B-IV, the related nuclear data files are replaced. And then, the reactor geometry buckling and multiplication factor are given in output tapes. The accuracy of calculated results is comparable with those of Monte Carlo and Sn method, and the agreement with experiment result is in 1%. (5 refs., 4 figs., 3 tabs.)

Development of hybrid core calculation system using 2-D full-core heterogeneous transport calculation and 3-D advanced nodal calculation

International Nuclear Information System (INIS)

Sugimura, Naoki; Mori, Masaaki; Hijiya, Masayuki; Ushio, Tadashi; Arakawa, Yasushi

2004-01-01

This paper presents the Hybrid Core Calculation System which is a very rigorous but a practical calculation system applicable to best estimate core design calculations taking advantage of the recent remarkable progress of computers. The basic idea of this system is to generate the correction factors for assembly homogenized cross sections, discontinuity factors, etc. by comparing the CASMO-4 and SIMULATE-3 2-D core calculation results under the consistent calculation condition and then apply them for SIMULATE-3 3-D calculation. The CASMO-4 2-D heterogeneous core calculation is performed for each depletion step with the core conditions previously determined by ordinary SIMULATE-3 core calculation to avoid time consuming iterative calculations searching for the critical boron concentrations while treating the thermal hydraulic feedback. The final SIMULATE-3 3-D calculation using the correction factors is performed with iterative calculations searching for the critical boron concentrations while treating the thermal hydraulic feedback. (author)

NEPTUNE: a modular system for light-water reactor calculation

International Nuclear Information System (INIS)

Bouchard, J.; Kanevoky, A.; Reuss, P.

1975-01-01

A complete modular system of light water reactor calculations has been designed. It includes basic nuclear data processing, the APOLLO phase: transport calculations for cells, multicells, fuel assemblies or reactors, the NEPTUNE phase: reactor calculations. A fuel management module, devoted to the automatic determination of the best shuffling strategy is included in NEPTUNE [fr

Generalized diffusion theory for calculating the neutron transport scalar flux

International Nuclear Information System (INIS)

Alcouffe, R.E.

1975-01-01

A generalization of the neutron diffusion equation is introduced, the solution of which is an accurate approximation to the transport scalar flux. In this generalization the auxiliary transport calculations of the system of interest are utilized to compute an accurate, pointwise diffusion coefficient. A procedure is specified to generate and improve this auxiliary information in a systematic way, leading to improvement in the calculated diffusion scalar flux. This improvement is shown to be contingent upon satisfying the condition of positive calculated-diffusion coefficients, and an algorithm that ensures this positivity is presented. The generalized diffusion theory is also shown to be compatible with conventional diffusion theory in the sense that the same methods and codes can be used to calculate a solution for both. The accuracy of the method compared to reference S/sub N/ transport calculations is demonstrated for a wide variety of examples. (U.S.)

Calculation of three-dimensional groundwater transport using second-order moments

International Nuclear Information System (INIS)

Pepper, D.W.; Stephenson, D.E.

1987-01-01

Groundwater transport of contaminants from the F-Area seepage basin at the Savannah River Plant (SRP) was calculated using a three-dimensional, second-order moment technique. The numerical method calculates the zero, first, and second moment distributions of concentration within a cell volume. By summing the moments over the entire solution domain, and using a Lagrangian advection scheme, concentrations are transported without numerical dispersion errors. Velocities obtained from field tests are extrapolated and interpolated to all nodal points; a variational analysis is performed over the three-dimensional velocity field to ensure mass consistency. Transport predictions are calculated out to 12,000 days. 28 refs., 9 figs

Molecular transport calculations with Wannier Functions

DEFF Research Database (Denmark)

Thygesen, Kristian Sommer; Jacobsen, Karsten Wedel

2005-01-01

We present a scheme for calculating coherent electron transport in atomic-scale contacts. The method combines a formally exact Green's function formalism with a mean-field description of the electronic structure based on the Kohn-Sham scheme of density functional theory. We use an accurate plane...

Supplementary neutron-flux calculations for the ORNL Pool Critical Assembly Pressure Vessel Facility

International Nuclear Information System (INIS)

Maudlin, P.J.; Maerker, R.E.

1982-01-01

A three-dimensional Monte Carlo calculation using the MORSE code was performed to validate a procedure previously adopted in the ORNL discrete ordinate analysis of measurements made in the ORNL Pool Critical Assembly Pressure Vessel Facility. The results of these flux calculations agree, within statistical undertainties of about 5%, with those obtained from a discrete ordinate analysis employing the same procedure. This study therefore concludes that the procedure for combining several one- and two-dimensional discrete ordinate calculations into a three-dimensional flux is sufficiently accurate that it does not account for the existing discrepancies observed between calculations and measurements in this facility

Supplementary neutron-flux calculations for the ORNL Pool Critical Assembly Pressure Vessel Facility

Energy Technology Data Exchange (ETDEWEB)

Maudlin, P.J.; Maerker, R.E.

1982-01-01

A three-dimensional Monte Carlo calculation using the MORSE code was performed to validate a procedure previously adopted in the ORNL discrete ordinate analysis of measurements made in the ORNL Pool Critical Assembly Pressure Vessel Facility. The results of these flux calculations agree, within statistical undertainties of about 5%, with those obtained from a discrete ordinate analysis employing the same procedure. This study therefore concludes that the procedure for combining several one- and two-dimensional discrete ordinate calculations into a three-dimensional flux is sufficiently accurate that it does not account for the existing discrepancies observed between calculations and measurements in this facility.

On the calculation of multi-group fission spectrum vectors

International Nuclear Information System (INIS)

Mueller, E.Z.

1984-05-01

In this report, the problem of calculating fission spectrum vectors in a consistent manner is formulated. The practical implications of using fission spectrum vectors in multi-group transport calculations are also addressed. The significance of the weighting spectra used for the calculation of fission spectrum vectors is illustrated for the case of a simple neutronic assembly

Viscosity, granular-temperature, and stress calculations for shearing assemblies of inelastic, frictional disks

International Nuclear Information System (INIS)

Walton, O.R.; Braun, R.L.

1986-01-01

Employing nonequilibrium molecular-dynamics methods the effects of two energy loss mechanisms on viscosity, stress, and granular-temperature in assemblies of nearly rigid, inelastic frictional disks undergoing steady-state shearing are calculated. Energy introduced into the system through forced shearing is dissipated by inelastic normal forces or through frictional sliding during collisions resulting in a natural steady-state kinetic energy density (granular-temperature) that depends on the density and shear rate of the assembly and on the friction and inelasticity properties of the disks. The calculations show that both the mean deviatoric particle velocity and the effective viscosity of a system of particles with fixed friction and restitution coefficients increase almost linearly with strain rate. Particles with a velocity-dependent coefficient of restitution show a less rapid increase in both deviatoric velocity and viscosity as strain rate increases. Particles with highly dissipative interactions result in anisotropic pressure and velocity distributions in the assembly, particularly at low densities. At very high densities the pressure also becomes anisotropic due to high contact forces perpendicular to the shearing direction. The mean rotational velocity of the frictional disks is nearly equal to one-half the shear rate. The calculated ratio of shear stress to normal stress varies significantly with density while the ratio of shear stress to total pressure shows much less variation. The inclusion of surface friction (and thus particle rotation) decreases shear stress at low density but increases shear stress under steady shearing at higher densities

DRAGON, Reactor Cell Calculation System with Burnup

International Nuclear Information System (INIS)

2007-01-01

1 - Description of program or function: DRAGON is a collection of models to simulate the neutronic behavior of a unit cell or a fuel assembly in a nuclear reactor. It includes all of the functions that characterize a lattice cell code, namely: interpolation of microscopic cross sections supplied by means of standard libraries; resonance self-shielding calculations in multidimensional geometries; multigroup and multidimensional neutron flux calculations which can take into account neutron leakage; transport-transport or transport-diffusion equivalence calculations as well as editing of condensed and homogenized nuclear properties for reactor calculations; and finally isotopic depletion calculations. The user must supply cross sections. DRAGON can access directly standard microscopic cross-section libraries in the following formats: DRAGON, MATXS (TRANSX-CTR), WIMSD4, WIMS-AECL, and APOLLO. It has the capability of exchanging macroscopic and microscopic cross-section libraries with a code such as PSR-0206/TRANSX-CTR or PSR-0317/TRANSX-2 by the use of the GOXS and ISOTXS format files. Macroscopic cross sections can also be read in DRAGON via the input data stream. 2 - Method of solution: DRAGON contains a multigroup iterator conceived to control a number of different algorithms for the solution of the neutron transport equation. Each of these algorithms is presented in the form of a one-group solution procedure where the contributions from other energy groups are included in a source term. The current version, DRAGON 9 71124 (Release 3.02), which was released in January 1998, contains three such algorithms. The JPM option solves the integral transport equation using the interface current method applied to homogeneous blocks; the SYBIL option solves the integral transport equation using the collision probability method for simple one-dimensional (1-D) or two-dimensional (2-D) geometries and the interface current method for 2-D Cartesian or hexagonal assemblies; and the

Benchmarks with diffusion theory and transport theory

International Nuclear Information System (INIS)

Cunha Menezes Filho, A. da; Souza, A.L. de.

1984-01-01

The multiplication factor and some spectral indices for five critical assemblies (ZPR-6-7, ZPR-3-11, GODIVA, BIG-TEN and FLATTOP) are calculated by Diffusion and Transport Theory, with group constants generated by MC 2 (for diffusion calculations) and by NJOY (for transport calculations). The discrepancies encountered in the ZPR-6-7 spectra, can be tracked to the large differences in the elastic cross section for Iron, calculated by MC 2 and NJOY. (Author) [pt

Thermal-hydraulic calculations for a fuel assembly in a European Pressurized Reactor using the RELAP5 code

Directory of Open Access Journals (Sweden)

Skrzypek Maciej

2015-09-01

Full Text Available The main object of interest was a typical fuel assembly, which constitutes a core of the nuclear reactor. The aim of the paper is to describe the phenomena and calculate thermal-hydraulic characteristic parameters in the fuel assembly for a European Pressurized Reactor (EPR. To perform thermal-hydraulic calculations, the RELAP5 code was used. This code allows to simulate steady and transient states for reactor applications. It is also an appropriate calculation tool in the event of a loss-of-coolant accident in light water reactors. The fuel assembly model with nodalization in the RELAP5 (Reactor Excursion and Leak Analysis Program code was presented. The calculations of two steady states for the fuel assembly were performed: the nominal steady-state conditions and the coolant flow rate decreased to 60% of the nominal EPR flow rate. The calculation for one transient state for a linearly decreasing flow rate of coolant was simulated until a new level was stabilized and SCRAM occurred. To check the correctness of the obtained results, the authors compared them against the reactor technical documentation available in the bibliography. The obtained results concerning steady states nearly match the design data. The hypothetical transient showed the importance of the need for correct cooling in the reactor during occurrences exceeding normal operation. The performed analysis indicated consequences of the coolant flow rate limitations during the reactor operation.

Parallel SN transport calculations on a transputer network

International Nuclear Information System (INIS)

Kim, Yong Hee; Cho, Nam Zin

1994-01-01

A parallel computing algorithm for the neutron transport problems has been implemented on a transputer network and two reactor benchmark problems (a fixed-source problem and an eigenvalue problem) are solved. We have shown that the parallel calculations provided significant reduction in execution time over the sequential calculations

The transport of fuel assemblies. New containers for transport the used nuclear material in Juzbado factory

International Nuclear Information System (INIS)

2005-01-01

Juzbado Manufacturing Facility is designed to be versatile and flexible. It is manufactured different kind of fuel assemblies PWR, BWR and VVER, beginning by the uranium oxide coming from the conversion facilities. The transport of these products (radioactive material fissile) requires the availability of different kind of packages; our models variety is similar to the big manufacturers. It is required a depth knowledge of the licensing process, approvals, manufacturing and handling instruction to be confident. Moreover, the recently changes on the Transport Regulations and the demands for the approval by the Competent Authorities have required the renovation of most of the package designs for the transport of radioactive material fissile worldwide. ENUSA assumed time ago this renovation and it is nowadays in the pick moment of this process. If we also consider the complexity on the management of multimodal international transportations, the Logistic task for the transport of nuclear material associated to the Juzbado factory results in a real changeling area. (Author)

Analysis of core calculation schemes for advanced water reactors

International Nuclear Information System (INIS)

Nicolas, Anne

1989-01-01

This research thesis addresses the analysis of the core control of sub-moderated water reactors with plutonium fuel and varying spectrum. Firstly, a calculation scheme is defined, based on transport theory for the three existing assembly configurations. It is based on the efficiency analysis of the control cluster and of the flow sheet shape in the assembly. Secondly, studies of the assembly with control cluster and within a theory of diffusion with homogenization or detailed assembly representation are performed by taking the environment into account in order to assess errors. Thirdly, due to the presence of a very efficient absorbent in control clusters, a deeper physical analysis requires the study of the flow gradient existing at the interface between assemblies. A parameter is defined to assess this gradient, and theoretically calculated by using finite elements. Developed software is validated [fr

ASOP, Shield Calculation, 1-D, Discrete Ordinates Transport

International Nuclear Information System (INIS)

1993-01-01

1 - Nature of physical problem solved: ASOP is a shield optimization calculational system based on the one-dimensional discrete ordinates transport program ANISN. It has been used to design optimum shields for space applications of SNAP zirconium-hydride-uranium- fueled reactors and uranium-oxide fueled thermionic reactors and to design beam stops for the ORELA facility. 2 - Method of solution: ASOP generates coefficients of linear equations describing the logarithm of the dose and dose-weight derivatives as functions of position from data obtained in an automated sequence of ANISN calculations. With the dose constrained to a design value and all dose-weight derivatives required to be equal, the linear equations may be solved for a new set of shield dimensions. Since changes in the shield dimensions may cause the linear functions to change, the entire procedure is repeated until convergence is obtained. The detailed calculations of the radiation transport through shield configurations for every step in the procedure distinguish ASOP from other shield optimization computer code systems which rely on multiple component sources and attenuation coefficients to describe the transport. 3 - Restrictions on the complexity of the problem: Problem size is limited only by machine size

The nuclear heating calculation scheme for material testing in the future Jules Horowitz Reactor

International Nuclear Information System (INIS)

Huot, N.; Aggery, A.; Blanchet, D.; Courcelle, A.; Czernecki, S.; Di-Salvo, J.; Doederlein, C.; Serviere, H.; Willermoz, G.

2004-01-01

An innovative nuclear heating calculation scheme for materials testing carried out in in the future Jules Horowitz reactor (JHR) is described. A heterogeneous gamma source calculation is first performed at assembly level using the deterministic code APOLLO2. This is followed by a Monte Carlo gamma transport calculation in the whole core using the TRIPOLI4 code. The calculated gamma sources at the assembly level are applied in the whole core simulation using a weighting based on power distribution obtained from the neutronic core calculation. (authors)

Burnup calculations using Monte Carlo method

International Nuclear Information System (INIS)

Ghosh, Biplab; Degweker, S.B.

2009-01-01

In the recent years, interest in burnup calculations using Monte Carlo methods has gained momentum. Previous burn up codes have used multigroup transport theory based calculations followed by diffusion theory based core calculations for the neutronic portion of codes. The transport theory methods invariably make approximations with regard to treatment of the energy and angle variables involved in scattering, besides approximations related to geometry simplification. Cell homogenisation to produce diffusion, theory parameters adds to these approximations. Moreover, while diffusion theory works for most reactors, it does not produce accurate results in systems that have strong gradients, strong absorbers or large voids. Also, diffusion theory codes are geometry limited (rectangular, hexagonal, cylindrical, and spherical coordinates). Monte Carlo methods are ideal to solve very heterogeneous reactors and/or lattices/assemblies in which considerable burnable poisons are used. The key feature of this approach is that Monte Carlo methods permit essentially 'exact' modeling of all geometrical detail, without resort to ene and spatial homogenization of neutron cross sections. Monte Carlo method would also be better for in Accelerator Driven Systems (ADS) which could have strong gradients due to the external source and a sub-critical assembly. To meet the demand for an accurate burnup code, we have developed a Monte Carlo burnup calculation code system in which Monte Carlo neutron transport code is coupled with a versatile code (McBurn) for calculating the buildup and decay of nuclides in nuclear materials. McBurn is developed from scratch by the authors. In this article we will discuss our effort in developing the continuous energy Monte Carlo burn-up code, McBurn. McBurn is intended for entire reactor core as well as for unit cells and assemblies. Generally, McBurn can do burnup of any geometrical system which can be handled by the underlying Monte Carlo transport code

Minaret, a deterministic neutron transport solver for nuclear core calculations

Energy Technology Data Exchange (ETDEWEB)

Moller, J-Y.; Lautard, J-J., E-mail: jean-yves.moller@cea.fr, E-mail: jean-jacques.lautard@cea.fr [CEA - Centre de Saclay , Gif sur Yvette (France)

2011-07-01

We present here MINARET a deterministic transport solver for nuclear core calculations to solve the steady state Boltzmann equation. The code follows the multi-group formalism to discretize the energy variable. It uses discrete ordinate method to deal with the angular variable and a DGFEM to solve spatially the Boltzmann equation. The mesh is unstructured in 2D and semi-unstructured in 3D (cylindrical). Curved triangles can be used to fit the exact geometry. For the curved elements, two different sets of basis functions can be used. Transport solver is accelerated with a DSA method. Diffusion and SPN calculations are made possible by skipping the transport sweep in the source iteration. The transport calculations are parallelized with respect to the angular directions. Numerical results are presented for simple geometries and for the C5G7 Benchmark, JHR reactor and the ESFR (in 2D and 3D). Straight and curved finite element results are compared. (author)

Minaret, a deterministic neutron transport solver for nuclear core calculations

International Nuclear Information System (INIS)

Moller, J-Y.; Lautard, J-J.

2011-01-01

We present here MINARET a deterministic transport solver for nuclear core calculations to solve the steady state Boltzmann equation. The code follows the multi-group formalism to discretize the energy variable. It uses discrete ordinate method to deal with the angular variable and a DGFEM to solve spatially the Boltzmann equation. The mesh is unstructured in 2D and semi-unstructured in 3D (cylindrical). Curved triangles can be used to fit the exact geometry. For the curved elements, two different sets of basis functions can be used. Transport solver is accelerated with a DSA method. Diffusion and SPN calculations are made possible by skipping the transport sweep in the source iteration. The transport calculations are parallelized with respect to the angular directions. Numerical results are presented for simple geometries and for the C5G7 Benchmark, JHR reactor and the ESFR (in 2D and 3D). Straight and curved finite element results are compared. (author)

Electrical transport properties in Fe-Cr nanocluster-assembled granular films

Science.gov (United States)

Wang, Xiong-Zhi; Wang, Lai-Sen; Zhang, Qin-Fu; Liu, Xiang; Xie, Jia; Su, A.-Mei; Zheng, Hong-Fei; Peng, Dong-Liang

2017-09-01

The Fe100-xCrx nanocluster-assembled granular films with Cr atomic fraction (x) ranging from 0 to 100 were fabricated by using a plasma-gas-condensation cluster deposition system. The TEM characterization revealed that the uniform Fe clusters were coated with a Cr layer to form a Fe-Cr core-shell structure. Then, the as-prepared Fe100-xCrx nanoclusters were randomly assembled into a granular film in vacuum environments with increasing the deposition time. Because of the competition between interfacial resistance and shunting effect of Cr layer, the room temperature resistivity of the Fe100-xCrx nanocluster-assembled granular films first increased and then decreased with increasing the Cr atomic fraction (x), and revealed a maximum of 2 × 104 μΩ cm at x = 26 at.%. The temperature-dependent longitudinal resistivity (ρxx), magnetoresistance (MR) effect and anomalous Hall effect (AHE) of these Fe100-xCrx nanocluster-assembled granular films were also studied systematically. As the x increased from 0 to 100, the ρxx of all samples firstly decreased and then increased with increasing the measuring temperature. The dependence of ρxx on temperature could be well addressed by a mechanism incorporated for the fluctuation-induced-tunneling (FIT) conduction process and temperature-dependent scattering effect. It was found that the anomalous Hall effect (AHE) had no legible scaling relation in Fe100-xCrx nanocluster-assembled granular films. However, after deducting the contribution of tunneling effect, the scaling relation was unambiguous. Additionally, the Fe100-xCrx nanocluster-assembled granular films revealed a small negative magnetoresistance (MR), which decreased with the increase of x. The detailed physical mechanism of the electrical transport properties in these Fe100-xCrx nanocluster-assembled granular films was also studied.

Transport losses in single and assembled coated conductors with textured-metal substrate with reduced magnetism

International Nuclear Information System (INIS)

Amemiya, N.; Jiang, Z.; Li, Z.; Nakahata, M.; Kato, T.; Ueyama, M.; Kashima, N.; Nagaya, S.; Shiohara, S.

2008-01-01

Transport losses in a coated conductor with a textured-metal substrate with reduced magnetism were studied experimentally. The substrate is with a clad structure, and HoBCO superconductor layer is deposited on the substrate with buffer layers. The measured transport loss of a sample whose critical current is 126.0 A falls between Norris's strip value and Norris's ellipse value. The increase in the measured transport loss from Norris's strip value can be attributed to its non-uniform lateral J c distribution. The same buffered clad tape was placed under an IBAD-MOCVD coated conductor with a non-magnetic substrate, and its transport loss was measured. The comparison between the measured transport loss of this sample and that of the identical IBAD-MOCVD coated conductor without the buffered clad tape indicates that the increase in the transport loss due to this buffered clad tape is small. The transport losses of hexagonal assemblies of IBAD-MOCVD coated conductors, whose structure simulates that of superconducting power transmission cables, were also measured where the buffered clad tapes were under-lied or over-lied on the coated conductors. The increase in the transport loss of hexagonal assemblies of coated conductors due to the buffered clad tapes is at an allowable level

On the mixing model for calculating the temperature fields in nuclear reactor fuel assemblies

International Nuclear Information System (INIS)

Mikhin, V.I.; Zhukov, A.V.

1985-01-01

One of the alternatives of the mixing model applied for calculating temperature fields in nuclear reactor fuel assemblies,including the fuel assemblies with nonequilibrium energy-release in fuel element cross section, is consistently described. The equations for both constant and variable values of coolant density and heat capacity are obtained. The mixing model is based on a set of mass, heat and longitudinal momentum balance equations. This set is closed by the ratios connecting the unknown values for gaps between fuel elements with the averaged values for neighbouring channels. The ratios to close momentum and heat balance equations, explaining, in particular, the nonequivalent heat and mass, momentum and mass transfer coefficients, are suggested. The balance equations with variable coolant density and heat capacity are reduced to the form coinciding with those of the similar equations with constant values of these parameters. Application of one of the main ratios of the mixing model relating the coolant transverse overflow in the gaps between fuel elements to the averaged coolant rates (flow rates) in the neighbouring channels is mainly limited by the coolant stabilized flow in the fuel assemblies with regular symmetrical arrangement of elements. Mass transfer coefficients for these elements are experimentally determined. The ratio in the paper is also applicable for calculation of fuel assembly temperature fields with a small relative shift of elements

Analysis of time-of-flight experiment on lithium-oxide assemblies by a two-dimensional transport code DOT3.5

International Nuclear Information System (INIS)

Oyama, Yukio; Yamaguchi, Seiya; Maekawa, Hiroshi

1985-03-01

Calculational analyses were made on the time-of-flight experiment of neutron leakage spectra from lithium-oxide slabs. The uncertainties in the calculation due to modelling were examined and it was estimated to be 1-2 %. The calculational results were compared with the experimental ones. The calculations were carried out by a two-dimensional transport code DOT3.5 using ENDF/B-4 nuclear data file. The comparison of energy-integrated fluxes in C/E from made it clear that the tendency of discrepancy between both results depended on the thickness of assembly and leaking angle. The discrepancy of C/E was about 40 % at the maximum. The effect due to the cross section change to a new data of 7 Li(n,n't) 4 He was also examined. This type of comparison is useful for the systematic assesments. From the comparison, it was suggested that the angular distribution of secondary neutron should be improved in the calculation, and the correct differential data of cross section are required. (author)

Application of a numerical transport correction in diffusion calculations

International Nuclear Information System (INIS)

Tomatis, Daniele; Dall'Osso, Aldo

2011-01-01

Full core calculations by ordinary transport methods can demand considerable computational time, hardly acceptable in the industrial work frame. However, the trend of next generation nuclear cores goes toward more heterogeneous systems, where transport phenomena of neutrons become very important. On the other hand, using diffusion solvers is more practical allowing faster calculations, but a specific formulation of the diffusion coefficient is requested to reproduce the scalar flux with reliable physical accuracy. In this paper, the Ronen method is used to evaluate numerically the diffusion coefficient in the slab reactor. The new diffusion solution is driven toward the solution of the integral neutron transport equation by non linear iterations. Better estimates of currents are computed and diffusion coefficients are corrected at node interfaces, still assuming Fick's law. This method enables obtaining closer results to the transport solution by a common solver in multigroup diffusion. (author)

Efficient calculation of dissipative quantum transport properties in semiconductor nanostructures

Energy Technology Data Exchange (ETDEWEB)

Greck, Peter

2012-11-26

We present a novel quantum transport method that follows the non-equilibrium Green's function (NEGF) framework but side steps any self-consistent calculation of lesser self-energies by replacing them by a quasi-equilibrium expression. We termed this method the multi-scattering Buettiker-Probe (MSB) method. It generalizes the so-called Buettiker-Probe model but takes into account all relevant individual scattering mechanisms. It is orders of magnitude more efficient than a fully selfconsistent non-equilibrium Green's function calculation for realistic devices, yet accurately reproduces the results of the latter method as well as experimental data. This method is fairly easy to implement and opens the path towards realistic three-dimensional quantum transport calculations. In this work, we review the fundamentals of the non-equilibrium Green's function formalism for quantum transport calculations. Then, we introduce our novel MSB method after briefly reviewing the original Buettiker-Probe model. Finally, we compare the results of the MSB method to NEGF calculations as well as to experimental data. In particular, we calculate quantum transport properties of quantum cascade lasers in the terahertz (THz) and the mid-infrared (MIR) spectral domain. With a device optimization algorithm based upon the MSB method, we propose a novel THz quantum cascade laser design. It uses a two-well period with alternating barrier heights and complete carrier thermalization for the majority of the carriers within each period. We predict THz laser operation for temperatures up to 250 K implying a new temperature record.

Nuclear performance calculations for the ELMO Bumpy Torus Reactor (EBTR) reference design

International Nuclear Information System (INIS)

Santoro, R.T.; Barnes, J.M.

1977-12-01

The nuclear performance of the ELMO Bumpy Torus Reactor reference design has been calculated using the one-dimensional discrete ordinates code ANISN and the latest available ENDF/B-IV transport cross-section data and nuclear response functions. The calculated results include estimates of the spatial and integral heating rate with emphasis on the recovery of fusion neutron energy in the blanket assembly and minimization of the energy deposition rates in the cryogenic magnet coil assemblies. The tritium breeding ratio in the natural lithium-laden blanket was calculated to be 1.29 tritium nuclei per incident neutron. The radiation damage in the reactor structural material and in the magnet assembly is also given

Study on a transportation and emplacement system of pre-assembled EBS module for HLW geological disposal

International Nuclear Information System (INIS)

Awano, Toshihiko; Kanno, Takeshi; Katsumata, Syunsuke; Kosuge, Kazuhiro

2009-01-01

HLW disposal is one of the largest issue to utilize Nuclear power safely. In the past study, the concept, which buffer materials and Overpacked waste were transported into underground respectively, have shown. The concept of pre-assembled engineered barrier has advantage to simplify the logistics and emplacement procedure, however there are difficulties to support heavy weight of pre-assembled package by equipment under the condition of little clearance between tunnel and package. In this study, Combination of air bearing and two degree-of-freedom wheels were suggested for transportation, and air jack was suggested for unloading and emplacement system. Also, whole system for transportation and emplacement procedure was designed, and Scale model test was examined to evaluate the feasibility of these concept and functions. (author)

A meshless approach to radionuclide transport calculations

International Nuclear Information System (INIS)

Perko, J.; Sarler, B.

2005-01-01

Over the past thirty years numerical modelling has emerged as an interdisciplinary scientific discipline which has a significant impact in engineering and design. In the field of numerical modelling of transport phenomena in porous media, many commercial codes exist, based on different numerical methods. Some of them are widely used for performance assessment and safety analysis of radioactive waste repositories and groundwater modelling. Although they proved to be an accurate and reliable tool, they have certain limitations and drawbacks. Realistic problems often involve complex geometry which is difficult and time consuming to discretize. In recent years, meshless methods have attracted much attention due to their flexibility in solving engineering and scientific problems. In meshless methods the cumbersome polygonization of calculation domain is not necessary. By this the discretization time is reduced. In addition, the simulation is not as discretization density dependent as in traditional methods because of the lack of polygon interfaces. In this work fully meshless Diffuse Approximate Method (DAM) is used for calculation of radionuclide transport. Two cases are considered; First 1D comparison of 226 Ra transport and decay solved by the commercial Finite Volume Method (FVM) and Finite Element Method (FEM) based packages and DAM. This case shows the level of discretization density dependence. And second realistic 2D case of near-field modelling of radionuclide transport from the radioactive waste repository. Comparison is made again between FVM based code and DAM simulation for two radionuclides: Long-lived 14 C and short-lived 3 H. Comparisons indicate great capability of meshless methods to simulate complex transport problems and show that they should be seriously considered in future commercial simulation tools. (author)

Development and application of neutron transport methods and uncertainty analyses for reactor core calculations. Technical report; Entwicklung und Einsatz von Neutronentransportmethoden und Unsicherheitsanalysen fuer Reaktorkernberechnungen. Technischer Bericht

Energy Technology Data Exchange (ETDEWEB)

Zwermann, W.; Aures, A.; Bernnat, W.; and others

2013-06-15

This report documents the status of the research and development goals reached within the reactor safety research project RS1503 ''Development and Application of Neutron Transport Methods and Uncertainty Analyses for Reactor Core Calculations'' as of the 1{sup st} quarter of 2013. The superordinate goal of the project is the development, validation, and application of neutron transport methods and uncertainty analyses for reactor core calculations. These calculation methods will mainly be applied to problems related to the core behaviour of light water reactors and innovative reactor concepts. The contributions of this project towards achieving this goal are the further development, validation, and application of deterministic and stochastic calculation programmes and of methods for uncertainty and sensitivity analyses, as well as the assessment of artificial neutral networks, for providing a complete nuclear calculation chain. This comprises processing nuclear basis data, creating multi-group data for diffusion and transport codes, obtaining reference solutions for stationary states with Monte Carlo codes, performing coupled 3D full core analyses in diffusion approximation and with other deterministic and also Monte Carlo transport codes, and implementing uncertainty and sensitivity analyses with the aim of propagating uncertainties through the whole calculation chain from fuel assembly, spectral and depletion calculations to coupled transient analyses. This calculation chain shall be applicable to light water reactors and also to innovative reactor concepts, and therefore has to be extensively validated with the help of benchmarks and critical experiments.

DRAGON 3.05D, Reactor Cell Calculation System with Burnup

International Nuclear Information System (INIS)

2007-01-01

1 - Description of program or function: The computer code DRAGON contains a collection of models that can simulate the neutron behavior of a unit cell or a fuel assembly in a nuclear reactor. It includes all of the functions that characterize a lattice cell code, namely: the interpolation of microscopic cross sections supplied by means of standard libraries; resonance self-shielding calculations in multidimensional geometries; multigroup and multidimensional neutron flux calculations that can take into account neutron leakage; transport-transport or transport-diffusion equivalence calculations as well as editing of condensed and homogenized nuclear properties for reactor calculations; and finally isotopic depletion calculations. 2 - Methods: The code DRAGON contains a multigroup flux solver conceived that can use a various algorithms to solve the neutron transport equation for the spatial and angular distribution of the flux. Each of these algorithms is presented in the form of a one-group solution procedure where the contributions from other energy groups are considered as sources. The current release of DRAGON contains five such algorithms. The JPM option that solves the integral transport equation using the J+- method, (interface current method applied to homogeneous blocks); the SYBIL option that solves the integral transport equation using the collision probability method for simple one dimensional (1-D) or two dimensional (2-D) geometries and the interface current method for 2-D Cartesian or hexagonal assemblies; the EXCELL/NXT option to solve the integral transport equation using the collision probability method for more general 2-D geometries and for three dimensional (3-D) assemblies; the MOCC option to solve the transport equation using the method of cyclic characteristics in 2-D Cartesian, and finally the MCU option to solve the transport equation using the method of characteristics (non cyclic) for 3-D Cartesian geometries. The execution of DRAGON is

Buckling resistance calculation of Guide Thimbles for the mechanical design of fuel assembly type PWR under normal reactor operating conditions

International Nuclear Information System (INIS)

Cruz, C.B.L.

1990-01-01

The calculations demonstrate the fulfillment of one of the mechanical design criteria for the Fuel Assembly Structure under normal reactor operating conditions. The calculations of stresses in the Guide Thimbles are performed with the aid of the program ANSYS. This paper contains program parameters and modelling of a typical Fuel Assembly for a Reactor similar to ANGRA II. (author)

Calculation of transportation energy for biomass collection

Energy Technology Data Exchange (ETDEWEB)

Kanai, G.; Takekura, K.; Kato, H.; Kobayashi, Y.; Yakushido, K. [National Agricultural Research Center, Tsukuba, Ibaraki (Japan)

2010-07-01

This paper reported on a study at a rice straw facility in Japan that produces bioethanol. Simulation modeling and calculations methods were used to examine the characteristics of field-to-facility transportation. Fuel consumption was found to be influenced by the conversion rate from straw to ethanol, the quantity of straw collected, and the ratio of the field area to that around the facility. Standard conditions were assumed based on reported data and actual observations for 15 ML/yr ethanol production, 0.3 kL output of ethanol from 1 t dry straw, 53.6 day/yr working days, 2.7 t truck load capacity, and 0.128 as the ratio of field to the area around the facility. According to calculations, a quantity of 50 kt dry straw requires 2.78 L of fuel to transport 1 t of dry straw, 109.5 trucks, and a 19.1 km collection area radius. The fuel consumption for transportation was found to be proportional to the quantity of straw to the 0.5 power, but inversely proportional to the ratio of field to the 0.5 power. The rate of increase in the number of trucks needed to collect straw increases with the decrease in the ratio of the field to area surface around the facility.

Calculation of Selected Emissions from Transport Services in Road Public Transport

Directory of Open Access Journals (Sweden)

Konečný Vladimír

2017-01-01

Full Text Available The article deals with road public transport and its impact on the environment. According to the methodology given in EN 16258, CO2 emission value has been calculated. The input data for the calculation and the results are shown in the tables. The declaration is created according to STN CEN / TR 14310, which contains recommendations for compiling environmental reports. Finally, the comparison of the environmental impact of a bus and a passenger car, when converted to one passenger, bus has a lower CO2 emission than a passenger car in that section.

Optimal calculational schemes for solving multigroup photon transport problem

International Nuclear Information System (INIS)

Dubinin, A.A.; Kurachenko, Yu.A.

1987-01-01

A scheme of complex algorithm for solving multigroup equation of radiation transport is suggested. The algorithm is based on using the method of successive collisions, the method of forward scattering and the spherical harmonics method, and is realized in the FORAP program (FORTRAN, BESM-6 computer). As an example the results of calculating reactor photon transport in water are presented. The considered algorithm being modified may be used for solving neutron transport problems

Development of the model for the stress calculation of fuel assembly under accident load

International Nuclear Information System (INIS)

Kim, Il Kon

1993-01-01

The finite element model for the stress calculation in guide thimbles of a fuel assembly (FA) under seismic and loss-of-coolant-accident (LOCA) load is developed. For the stress calculation of FA under accident load, at first the program MAIN is developed to select the worst bending mode shaped FA from core model. And then the model for the stress calculation of FA is developed by means of the finite element code. The calculated results of program MAIN are used as the kinematic constraints of the finite element model of a FA. Compared the calculated results of the stiffness of the finite element model of FA with the test results they have good agreements. (Author)

The problems of calculation of heat transfer crisis in fuel assemblies of PW reactors based on modern versions of thermohydraulic codes

International Nuclear Information System (INIS)

Fialko, N.M.; Sharaevskij, G.I.; Sharaevskaya, E.I.; Babak, E.I.

2014-01-01

The article gives an analysis of the adequacy of computer software systems FASCICLE BM-DF and COBRA, which are designed to calculate the main parameters of the safety of water-cooled nuclear reactors. This calculation is based on determining the local thermal-hydraulic parameters of the flow of coolant in the fuel rod assembled elements. In this article introduced the results of the comparison of experiments performed to determine the distribution of the main thermal-hydraulic flow parameters characteristic of subchannels of fuel rod assembled elements with the data for calculating these parameters on the basis of declared computer codes. Particular attention is paid to the analysis of experimental and calculation data, by definition, burnout in rod fuel assembled elements

Test calculations of physical parameters of the TRX,BETTIS and MIT critical assemblies according to the TRIFON program

International Nuclear Information System (INIS)

Kochurov, B.P.

1980-01-01

Results of calculations of physical parameters characterizing the TRX, MIT and BETTIS critical assemblies obtained according to the program TRIFON are presented. The program TRIFON permits to calculate the space-energy neutron distribution in the multigroup approximation in a multizone cylindrical cell. Results of comparison of the TRX, BETTIS and MIT crytical assembly parameters with experimental data and calculational results according to the Monte Carlo method are presented as well. Deviations of the parameters are in the range of 1.5-2 of experimental errors. Data on the interference of uranium 238 levels in the resonant neutron absorption in the cell are given [ru

CALCULATION OF POLLUTION DYNAMICS NEAR RAILWAY TERRITORY DURING COAL TRANSPORTATION

Directory of Open Access Journals (Sweden)

M. M. Biliaiev

2017-02-01

Full Text Available Purpose. The article is aimed to develop 3D numerical model for the prediction of atmospheric pollution during transportation of bulk cargo in the railway car. Methodology.To solve this problem, it was developed three-dimensional numerical model, based on the use of the transport equation of dust pollution in the air by the wind and atmospheric turbulent diffusion. For the numerical integration of the simulating equation of the dust transport the implicit difference scheme was used. When constructing a difference scheme, it was carried out prior splitting of the original transport equation into the sequence of solutions of three equations. The first of them takes into account the transport of dust in paths, the second equation – dust transport under the influence of atmospheric turbulent diffusion, and the third equation –change of the dust concentration in the air due to its emissions from the cars.Unknown value of the pollutant concentration at every step of splitting is determined by the explicit scheme – the method of running account, which provides a simple numerical implementation of splitting equations. The developed numerical model is the basis for specialized computer program. On the basis of the constructed numerical model we carried out a computational experiment to assess the level of air pollution at the railway station during the motion of train with coal. Findings. Authors developed 3D numerical model, which belongs to the class of «screening models». This model takes into account the main physical factors affecting the process of dispersion of dust pollution in the atmosphere during coal transportation. The proposed numerical model requires low cost of computer time in the practical implementation on small and medium-power computers. This model can be used for rapid calculations of the dynamics of air pollution when transporting coal by rail. Calculations to determine the pollutant concentration and formation of the

Growth and anisotropic transport properties of self-assembled InAs nanostructures in InP

International Nuclear Information System (INIS)

Bierwagen, O.

2007-01-01

Self-assembled InAs nanostructures in InP, comprising quantum wells, quantum wires, and quantum dots, are studied in terms of their formation and properties. In particular, the structural, optical, and anisotropic transport properties of the nanostructures are investigated. The focus is a comprehending exploration of the anisotropic in-plane transport in large ensembles of laterally coupled InAs nanostructures. The self-assembled Stranski-Krastanov growth of InAs nanostructures is studied by gas-source molecular beam epitaxy on both nominally oriented and vicinal InP(001). Optical polarization of the interband transitions arising from the nanostructure type is demonstrated by photoluminescence and transmission spectroscopy. The experimentally convenient four-contact van der Pauw Hall measurement of rectangularly shaped semiconductors, usually applied to isotropic systems, is extended to yield the anisotropic transport properties. Temperature dependent transport measurements are performed in large ensembles of laterally closely spaced nanostructures. The transport of quantum wire-, quantum dash- and quantum dot containing samples is highly anisotropic with the principal axes of conductivity aligned to the directions. The direction of higher mobility is [ anti 110], which is parallel to the direction of the quantum wires. In extreme cases, the anisotropies exceed 30 for electrons, and 100 for holes. The extreme anisotropy for holes is due to diffusive transport through extended states in the [ anti 110], and hopping transport through laterally localized states in the [110] direction, within the same sample. A novel 5-terminal electronic switching device based on gate-controlled transport anisotropy is proposed. The gate-control of the transport anisotropy in modulation-doped, self-organized InAs quantum wires embedded in InP is demonstrated. (orig.)

Growth and anisotropic transport properties of self-assembled InAs nanostructures in InP

Energy Technology Data Exchange (ETDEWEB)

Bierwagen, O.

2007-12-20

Self-assembled InAs nanostructures in InP, comprising quantum wells, quantum wires, and quantum dots, are studied in terms of their formation and properties. In particular, the structural, optical, and anisotropic transport properties of the nanostructures are investigated. The focus is a comprehending exploration of the anisotropic in-plane transport in large ensembles of laterally coupled InAs nanostructures. The self-assembled Stranski-Krastanov growth of InAs nanostructures is studied by gas-source molecular beam epitaxy on both nominally oriented and vicinal InP(001). Optical polarization of the interband transitions arising from the nanostructure type is demonstrated by photoluminescence and transmission spectroscopy. The experimentally convenient four-contact van der Pauw Hall measurement of rectangularly shaped semiconductors, usually applied to isotropic systems, is extended to yield the anisotropic transport properties. Temperature dependent transport measurements are performed in large ensembles of laterally closely spaced nanostructures. The transport of quantum wire-, quantum dash- and quantum dot containing samples is highly anisotropic with the principal axes of conductivity aligned to the <110> directions. The direction of higher mobility is [ anti 110], which is parallel to the direction of the quantum wires. In extreme cases, the anisotropies exceed 30 for electrons, and 100 for holes. The extreme anisotropy for holes is due to diffusive transport through extended states in the [ anti 110], and hopping transport through laterally localized states in the [110] direction, within the same sample. A novel 5-terminal electronic switching device based on gate-controlled transport anisotropy is proposed. The gate-control of the transport anisotropy in modulation-doped, self-organized InAs quantum wires embedded in InP is demonstrated. (orig.)

Microwave emulations and tight-binding calculations of transport in polyacetylene

Energy Technology Data Exchange (ETDEWEB)

Stegmann, Thomas, E-mail: stegmann@icf.unam.mx [Instituto de Ciencias Físicas, Universidad Nacional Autónoma de México, Avenida Universidad s/n, 62210 Cuernavaca (Mexico); Franco-Villafañe, John A., E-mail: jofravil@fis.unam.mx [Instituto de Física, Benemérita Universidad Autónoma de Puebla, Apartado Postal J-48, 72570 Puebla (Mexico); Instituto de Ciencias Físicas, Universidad Nacional Autónoma de México, Avenida Universidad s/n, 62210 Cuernavaca (Mexico); Ortiz, Yenni P. [Instituto de Ciencias Físicas, Universidad Nacional Autónoma de México, Avenida Universidad s/n, 62210 Cuernavaca (Mexico); Kuhl, Ulrich [Université de Nice – Sophia Antipolis, Laboratoire de la Physique de la Matière Condensée, CNRS, Parc Valrose, 06108 Nice (France); Mortessagne, Fabrice, E-mail: fabrice.mortessagne@unice.fr [Université de Nice – Sophia Antipolis, Laboratoire de la Physique de la Matière Condensée, CNRS, Parc Valrose, 06108 Nice (France); Seligman, Thomas H. [Instituto de Ciencias Físicas, Universidad Nacional Autónoma de México, Avenida Universidad s/n, 62210 Cuernavaca (Mexico); Centro Internacional de Ciencias, 62210 Cuernavaca (Mexico)

2017-01-05

A novel approach to investigate the electron transport of cis- and trans-polyacetylene chains in the single-electron approximation is presented by using microwave emulation measurements and tight-binding calculations. In the emulation we take into account the different electronic couplings due to the double bonds leading to coupled dimer chains. The relative coupling constants are adjusted by DFT calculations. For sufficiently long chains a transport band gap is observed if the double bonds are present, whereas for identical couplings no band gap opens. The band gap can be observed also in relatively short chains, if additional edge atoms are absent, which cause strong resonance peaks within the band gap. The experimental results are in agreement with our tight-binding calculations using the nonequilibrium Green's function method. The tight-binding calculations show that it is crucial to include third nearest neighbor couplings to obtain the gap in the cis-polyacetylene. - Highlights: • Electronic transport in individual polyacetylene chains is studied. • Microwave emulation experiments and tight-binding calculations agree well. • In long chains a band-gap opens due the dimerization of the chain. • In short chains edge atoms cause strong resonance peaks in the center of the band-gap.

Microwave emulations and tight-binding calculations of transport in polyacetylene

International Nuclear Information System (INIS)

Stegmann, Thomas; Franco-Villafañe, John A.; Ortiz, Yenni P.; Kuhl, Ulrich; Mortessagne, Fabrice; Seligman, Thomas H.

2017-01-01

A novel approach to investigate the electron transport of cis- and trans-polyacetylene chains in the single-electron approximation is presented by using microwave emulation measurements and tight-binding calculations. In the emulation we take into account the different electronic couplings due to the double bonds leading to coupled dimer chains. The relative coupling constants are adjusted by DFT calculations. For sufficiently long chains a transport band gap is observed if the double bonds are present, whereas for identical couplings no band gap opens. The band gap can be observed also in relatively short chains, if additional edge atoms are absent, which cause strong resonance peaks within the band gap. The experimental results are in agreement with our tight-binding calculations using the nonequilibrium Green's function method. The tight-binding calculations show that it is crucial to include third nearest neighbor couplings to obtain the gap in the cis-polyacetylene. - Highlights: • Electronic transport in individual polyacetylene chains is studied. • Microwave emulation experiments and tight-binding calculations agree well. • In long chains a band-gap opens due the dimerization of the chain. • In short chains edge atoms cause strong resonance peaks in the center of the band-gap.

Load Balancing of Parallel Monte Carlo Transport Calculations

International Nuclear Information System (INIS)

Procassini, R J; O'Brien, M J; Taylor, J M

2005-01-01

The performance of parallel Monte Carlo transport calculations which use both spatial and particle parallelism is increased by dynamically assigning processors to the most worked domains. Since he particle work load varies over the course of the simulation, this algorithm determines each cycle if dynamic load balancing would speed up the calculation. If load balancing is required, a small number of particle communications are initiated in order to achieve load balance. This method has decreased the parallel run time by more than a factor of three for certain criticality calculations

Structural and Shielding Safety of a Transport Package for Radioisotope Sealed Source Assembly

Energy Technology Data Exchange (ETDEWEB)

Seo, Kiseog; Cho, Ilje; Kim, Donghak [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

2006-07-01

As some kinds of radioisotope (RI) sealed source are produced by HANARO research reactor, a demand of RI transport package is increasing gradually. Foreign countries, which produce the various RIs, have the intrinsic model of the RI transport package. It is necessary to develop a RI and its transport package simultaneously. It is difficult to design a shielding part for this transport package because the passage for this source assembly should be provided from the center of shielding part to the outside of the package. In order to endure the accident conditions such as a 9 m drop and puncture, this transport package consists of the guide tubes, a gamma shield and a shock absorber. This paper describe that a shielding and structural safety of RI sealed source transport package are evaluated under the accident conditions.

Structural and Shielding Safety of a Transport Package for Radioisotope Sealed Source Assembly

International Nuclear Information System (INIS)

Seo, Kiseog; Cho, Ilje; Kim, Donghak

2006-01-01

As some kinds of radioisotope (RI) sealed source are produced by HANARO research reactor, a demand of RI transport package is increasing gradually. Foreign countries, which produce the various RIs, have the intrinsic model of the RI transport package. It is necessary to develop a RI and its transport package simultaneously. It is difficult to design a shielding part for this transport package because the passage for this source assembly should be provided from the center of shielding part to the outside of the package. In order to endure the accident conditions such as a 9 m drop and puncture, this transport package consists of the guide tubes, a gamma shield and a shock absorber. This paper describe that a shielding and structural safety of RI sealed source transport package are evaluated under the accident conditions

Neutron Detector Signal Processing to Calculate the Effective Neutron Multiplication Factor of Subcritical Assemblies

International Nuclear Information System (INIS)

Talamo, Alberto; Gohar, Yousry

2016-01-01

This report describes different methodologies to calculate the effective neutron multiplication factor of subcritical assemblies by processing the neutron detector signals using MATLAB scripts. The subcritical assembly can be driven either by a spontaneous fission neutron source (e.g. californium) or by a neutron source generated from the interactions of accelerated particles with target materials. In the latter case, when the particle accelerator operates in a pulsed mode, the signals are typically stored into two files. One file contains the time when neutron reactions occur and the other contains the times when the neutron pulses start. In both files, the time is given by an integer representing the number of time bins since the start of the counting. These signal files are used to construct the neutron count distribution from a single neutron pulse. The built-in functions of MATLAB are used to calculate the effective neutron multiplication factor through the application of the prompt decay fitting or the area method to the neutron count distribution. If the subcritical assembly is driven by a spontaneous fission neutron source, then the effective multiplication factor can be evaluated either using the prompt neutron decay constant obtained from Rossi or Feynman distributions or the Modified Source Multiplication (MSM) method.

Neutron Detector Signal Processing to Calculate the Effective Neutron Multiplication Factor of Subcritical Assemblies

Energy Technology Data Exchange (ETDEWEB)

Talamo, Alberto [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Gohar, Yousry [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division

2016-06-01

This report describes different methodologies to calculate the effective neutron multiplication factor of subcritical assemblies by processing the neutron detector signals using MATLAB scripts. The subcritical assembly can be driven either by a spontaneous fission neutron source (e.g. californium) or by a neutron source generated from the interactions of accelerated particles with target materials. In the latter case, when the particle accelerator operates in a pulsed mode, the signals are typically stored into two files. One file contains the time when neutron reactions occur and the other contains the times when the neutron pulses start. In both files, the time is given by an integer representing the number of time bins since the start of the counting. These signal files are used to construct the neutron count distribution from a single neutron pulse. The built-in functions of MATLAB are used to calculate the effective neutron multiplication factor through the application of the prompt decay fitting or the area method to the neutron count distribution. If the subcritical assembly is driven by a spontaneous fission neutron source, then the effective multiplication factor can be evaluated either using the prompt neutron decay constant obtained from Rossi or Feynman distributions or the Modified Source Multiplication (MSM) method.

Discussion of electron cross sections for transport calculations

International Nuclear Information System (INIS)

Berger, M.J.

1983-01-01

This paper deals with selected aspects of the cross sections needed as input for transport calculations and for the modeling of radiation effects in biological materials. Attention is centered mainly on the cross sections for inelastic interactions between electrons and water molecules and the use of these cross sections for the calculation of energy degradation spectra and of ionization and excitation yields. 40 references, 3 figures, 1 table

RAGRAF: a computer code for calculating temperature distributions in multi-pin fuel assemblies in a stagnant gas atmosphere

International Nuclear Information System (INIS)

Eastham, A.

1979-02-01

A method of calculating the temperature distribution in a cross-section of a multi-pin nuclear reactor fuel assembly has been computerised. It utilises the thermal radiation interchange between individual fuel pins in either a square or triangular pitched lattice. A stagnant gas atmosphere within the fuel assembly is assumed which inhibits natural convection but permits thermal conduction between adjacent fuel pins. no restriction is placed upon the shape of wrapper used, but its temperature must always be uniform. RAGRAF has great flexibility because of the many options it provides. Although, essentially, it is a transient code, steady state solutions may be readily identified from successive temperature prints. An enclosure for the assembly wrapper is available, to be included or discarded at will during transient calculations. outside the limit of the assembly wrapper, any type or combination of heat transfer mode may be included. Transient variations in boundary temperature may be included if required. (author)

Accurate quantum calculations of translation-rotation eigenstates in electric-dipole-coupled H2O@C60 assemblies

Science.gov (United States)

Felker, Peter M.; Bačić, Zlatko

2017-09-01

We present methodology for variational calculation of the 6 n -dimensional translation-rotation (TR) eigenstates of assemblies of n H2O@C60 moieties coupled by dipole-dipole interactions. We show that the TR Hamiltonian matrix for any n can be constructed from dipole-dipole matrix elements computed for n = 2 . We present results for linear H2O@C60 assemblies. Two classes of eigenstates are revealed. One class comprises excitations of the 111 rotational level of H2O. The lowest-energy 111 -derived eigenstate for each assembly exhibits significant dipole ordering and shifts down in energy with the assembly size.

Entanglement of conjugated polymer chains influences molecular self-assembly and carrier transport

KAUST Repository

Zhao, Kui; Khan, Hadayat Ullah; Li, Ruipeng; Su, Yisong; Amassian, Aram

2013-01-01

The influence of polymer entanglement on the self-assembly, molecular packing structure, and microstructure of low-Mw (lightly entangled) and high-Mw (highly entangled) poly (3-hexylthiophene) (P3HT), and the carrier transport in thin-film transistors, are investigated. The polymer chains are gradually disentangled in a marginal solvent via ultrasonication of the polymer solution, and demonstrate improved diffusivity of precursor species (coils, aggregates, and microcrystallites), enhanced nucleation and crystallization of P3HT in solution, and self-assembly of well-ordered and highly textured fibrils at the solid-liquid interface. In low-Mw P3HT, reducing chain entanglement enhances interchain and intrachain ordering, but reduces the interconnectivity of ordered domains (tie molecules) due to the presence of short chains, thus deteriorating carrier transport even in the face of improving crystallinity. Reducing chain entanglement in high-Mw P3HT solutions increases carrier mobility up to ≈20-fold, by enhancing interchain and intrachain ordering while maintaining a sufficiently large number of tie molecules between ordered domains. These results indicate that charge carrier mobility is strongly governed by the balancing of intrachain and interchain ordering, on the one hand, and interconnectivity of ordered domains, on the other hand. In high-Mw P3HT, intrachain and interchain ordering appear to be the key bottlenecks to charge transport, whereas in low-Mw P3HT, the limited interconnectivity of the ordered domains acts as the primary bottleneck to charge transport. Conjugated polymer chains of poly(3-hexylthiophene) (P3HT) are gradually disentangled in solution and trends in carrier transport mechanisms in organic thin film transistors for low- and high-molecular weight P3HT are investigated. While intrachain and interchain ordering within ordered domains are the key bottlenecks to charge transport in high-Mw P3HT films, the limited interconnectivity of ordered

Entanglement of conjugated polymer chains influences molecular self-assembly and carrier transport

KAUST Repository

Zhao, Kui

2013-06-26

The influence of polymer entanglement on the self-assembly, molecular packing structure, and microstructure of low-Mw (lightly entangled) and high-Mw (highly entangled) poly (3-hexylthiophene) (P3HT), and the carrier transport in thin-film transistors, are investigated. The polymer chains are gradually disentangled in a marginal solvent via ultrasonication of the polymer solution, and demonstrate improved diffusivity of precursor species (coils, aggregates, and microcrystallites), enhanced nucleation and crystallization of P3HT in solution, and self-assembly of well-ordered and highly textured fibrils at the solid-liquid interface. In low-Mw P3HT, reducing chain entanglement enhances interchain and intrachain ordering, but reduces the interconnectivity of ordered domains (tie molecules) due to the presence of short chains, thus deteriorating carrier transport even in the face of improving crystallinity. Reducing chain entanglement in high-Mw P3HT solutions increases carrier mobility up to ≈20-fold, by enhancing interchain and intrachain ordering while maintaining a sufficiently large number of tie molecules between ordered domains. These results indicate that charge carrier mobility is strongly governed by the balancing of intrachain and interchain ordering, on the one hand, and interconnectivity of ordered domains, on the other hand. In high-Mw P3HT, intrachain and interchain ordering appear to be the key bottlenecks to charge transport, whereas in low-Mw P3HT, the limited interconnectivity of the ordered domains acts as the primary bottleneck to charge transport. Conjugated polymer chains of poly(3-hexylthiophene) (P3HT) are gradually disentangled in solution and trends in carrier transport mechanisms in organic thin film transistors for low- and high-molecular weight P3HT are investigated. While intrachain and interchain ordering within ordered domains are the key bottlenecks to charge transport in high-Mw P3HT films, the limited interconnectivity of ordered

Using of the Serpent code based on the Monte-Carlo method for calculation of the VVER-1000 fuel assembly characteristics

Directory of Open Access Journals (Sweden)

V. V. Galchenko

2016-12-01

Full Text Available The description of calculation scheme of fuel assembly for preparation of few-group characteristics is considered with help of Serpent code. This code uses the Monte-Carlo method and energy continuous microscopic data libraries. Serpent code is devoted for calculation of fuel assembly characteristics, burnup calculations and preparation of few-group homogenized macroscopic cross-sections. The results of verification simulations in comparison with other codes (WIMS, HELIOS, NESSEL etc., which are used for neutron-physical analysis of VVER type fuel, are presented.

Uncertainty analysis of neutron transport calculation

International Nuclear Information System (INIS)

Oka, Y.; Furuta, K.; Kondo, S.

1987-01-01

A cross section sensitivity-uncertainty analysis code, SUSD was developed. The code calculates sensitivity coefficients for one and two-dimensional transport problems based on the first order perturbation theory. Variance and standard deviation of detector responses or design parameters can be obtained using cross section covariance matrix. The code is able to perform sensitivity-uncertainty analysis for secondary neutron angular distribution(SAD) and secondary neutron energy distribution(SED). Covariances of 6 Li and 7 Li neutron cross sections in JENDL-3PR1 were evaluated including SAD and SED. Covariances of Fe and Be were also evaluated. The uncertainty of tritium breeding ratio, fast neutron leakage flux and neutron heating was analysed on four types of blanket concepts for a commercial tokamak fusion reactor. The uncertainty of tritium breeding ratio was less than 6 percent. Contribution from SAD/SED uncertainties are significant for some parameters. Formulas to estimate the errors of numerical solution of the transport equation were derived based on the perturbation theory. This method enables us to deterministically estimate the numerical errors due to iterative solution, spacial discretization and Legendre polynomial expansion of transfer cross-sections. The calculational errors of the tritium breeding ratio and the fast neutron leakage flux of the fusion blankets were analysed. (author)

LTRACK: Beam-transport calculation including wakefield effects

International Nuclear Information System (INIS)

Chan, K.C.D.; Cooper, R.K.

1988-01-01

LTRACK is a first-order beam-transport code that includes wakefield effects up to quadrupole modes. This paper will introduce the readers to this computer code by describing the history, the method of calculations, and a brief summary of the input/output information. Future plans for the code will also be described

Validation of the criticality calculation for fuel elements using the Gamtec 2 - Keno 2 and 4

International Nuclear Information System (INIS)

Teixeira, M.C.C.; Andrade, M.C. de

1990-01-01

For criticality safety in the fabrication, storage and transportation of fuel assemblies, subcriticality analysis must be done. The calculations are performed at CDTN with the GAMTEC computer code, to homogenize the fuel assembly in order to create 16 group cross-section library, and with KENO code, for determining the multiplication factor. To validate the calculational method, suitable Benchmark experiments have been done. The results show that the calculational model overestimates kef when kef+ 2 σ was considered. (author) [pt

Active Self-Assembled Spinners: dynamic crystals, transport and induced surface flows

Science.gov (United States)

Snezhko, Alexey; Kokot, Gasper

Strongly interacting colloids driven out-of-equilibrium by an external periodic forcing often develop nontrivial collective dynamics. Active magnetic colloids proved to be excellent model experimental systems to explore emergent behavior and active (out-of-equilibrium) self-assembly phenomena. Ferromagnetic micro-particles, suspended at a liquid interface and energized by a rotational homogeneous alternating magnetic field applied along the supporting interface, spontaneously form ensembles of synchronized self-assembled spinners with well-defined characteristic length. The size and the torque of an individual self-assembled spinner are controlled by the frequency of the driving magnetic field. Experiments reveal a rich collective dynamics in large ensembles of synchronized magnetic spinners that spontaneously form dynamic spinner lattices at the interface in a certain range of the excitation parameters. Non-trivial dynamics inside of the formed spinner lattices is observed. Transport of passive cargo particles and structure of the underlying self-induced surface flows is analyzed. The research was supported by the U.S. DOE, Office of Basic Energy Sciences, Division of Materials Science and Engineering.

Study on neutron streaming effect in large fast critical assembly

International Nuclear Information System (INIS)

Takeda, Toshikazu; Yamaoka, Mitsuaki; Sakurai, Shungo; Tanimoto, Koichi; Abe, Yuhei

1981-03-01

A cell calculation method taking into account the neutron leakage from a cell and a transport calculation method treating the neutron streaming have been developed, and their applicability has been investigated. In the cell calculation method, the neutron leakage in the perpendicular direction to plates was treated by introducing an albedo collision probability which is a first-flight collision probability incorporating albedos at cell boundaries, and that in the parallel direction was treated by the pseudo absorption method. The use of the albedo collision probability made it possible to calculate the flux tilt in a cell exactly. This cell calculation method was applied to two slab models where fuel drawers were stacked in perpendicular and parallel directions to plates. Cell averaged cross sections calculated by the proposed method agreed well with those obtained from exact transport calculations treating the plate-wise heterogeneity, while the infinite cell calculation and the conventional pseudo absorption method produced about 2% errors in the cell-averaged cross sections. The cell-averaging procedure for control-rod channels was also proposed, and this method was applied to the calculation of control-rod worths and control-rod position worths. A transport calculation method based on the response matrix method has been proposed to treat the neutron streaming in fast critical assemblies directly. A response matrix code in two dimensional XY geometry RES2D was made. The accuracy of response matrices obtained from the RES2D code was checked by applying it to a slab cell and by comparing cell-averaged cross sections and k-infinity with those from a reference cell calculation based on the collision probability. The agreement of the results was good, and it was found that the response matrix method is very promising for the treatment of the neutron streaming in fast critical assemblies. (author)

Prospects in deterministic three dimensional whole-core transport calculations

International Nuclear Information System (INIS)

Sanchez, Richard

2012-01-01

The point we made in this paper is that, although detailed and precise three-dimensional (3D) whole-core transport calculations may be obtained in the future with massively parallel computers, they would have an application to only some of the problems of the nuclear industry, more precisely those regarding multiphysics or for methodology validation or nuclear safety calculations. On the other hand, typical design reactor cycle calculations comprising many one-point core calculations can have very strict constraints in computing time and will not directly benefit from the advances in computations in large scale computers. Consequently, in this paper we review some of the deterministic 3D transport methods which in the very near future may have potential for industrial applications and, even with low-order approximations such as a low resolution in energy, might represent an advantage as compared with present industrial methodology, for which one of the main approximations is due to power reconstruction. These methods comprise the response-matrix method and methods based on the two-dimensional (2D) method of characteristics, such as the fusion method.

CALCULATING BEDLOAD TRANSPORT IN RIVERS: CONCEPTS, CALCULUS ROUTINES AND APPLICATION

Directory of Open Access Journals (Sweden)

Hudson de Azevedo Macedo

2017-10-01

Full Text Available Rivers are immensely important to human activities such as water supply, navigation, energy generation, and agriculture. They are also an important morphodynamic agent of erosion, transport and deposition. Their capacity to transport sediment depends on their hydraulic characteristics and can be predicted by mathematical models. Several mathematical models can be used to compute bed-load transport. Each one is appropriately better for certain conditions. In this paper, we present an application built in Microsoft Excel to compute the bed-load transport in rivers based on the Van Rijn mathematical model. The Van Rijn model is appropriate for rivers transporting sandy sediments in conditions of subcritical flow. Hydraulic parameters such as channel slope, stream power, and Reynolds and Froude numbers can be calculated using the application proposed here. The application was tested in the Paraná River and results from the calculations are consistent with data obtained from fieldwork surveys. The error of the application was only 20%, which shows good agreement of the model with survey values.

Calculations of the transport properties within the PAW formalism

Energy Technology Data Exchange (ETDEWEB)

Mazevet, S.; Torrent, M.; Recoules, V.; Jollet, F. [CEA Bruyeres-le-Chatel, DIF, 91 (France)

2010-07-01

We implemented the calculation of the transport properties within the PAW formalism in the ABINIT code. This feature allows the calculation of the electrical and optical properties, including the XANES spectrum, as well as the electronic contribution to the thermal conductivity. We present here the details of the implementation and results obtained for warm dense aluminum plasma. (authors)

Analysis of error in Monte Carlo transport calculations

International Nuclear Information System (INIS)

Booth, T.E.

1979-01-01

The Monte Carlo method for neutron transport calculations suffers, in part, because of the inherent statistical errors associated with the method. Without an estimate of these errors in advance of the calculation, it is difficult to decide what estimator and biasing scheme to use. Recently, integral equations have been derived that, when solved, predicted errors in Monte Carlo calculations in nonmultiplying media. The present work allows error prediction in nonanalog Monte Carlo calculations of multiplying systems, even when supercritical. Nonanalog techniques such as biased kernels, particle splitting, and Russian Roulette are incorporated. Equations derived here allow prediction of how much a specific variance reduction technique reduces the number of histories required, to be weighed against the change in time required for calculation of each history. 1 figure, 1 table

Contribution to gamma ray transport calculation in heterogeneous media

International Nuclear Information System (INIS)

Bourdet, L.

1985-04-01

This thesis presents the development of gamma transport calculation codes in three dimension heterogeneous geometries. These codes allow us to define the protection against gamma-rays or verify their efficiency. The laws that govern the interactions of gamma-rays with matters are briefly revised. A library with the all necessary constants for these codes is created. TRIPOLI-2, a code that treats in exact way the neutron transport in matters using Monte-Carlo method, has been adapted to deal with the transport of gamma-rays in matters as well. TRINISHI, a code which considers only one collision, has been realized to treat heterogeneous geometries containing voids. Elaborating a formula that calculates the albedo for gamma-ray reflection (the code ALBANE) allows us to solve the problem of gamma-ray reflection on plane surfaces. NARCISSE-2 deals with gamma-rays that suffer only one reflection on the inner walls of any closed volume (rooms, halls...) [fr

Verification of FA2D Prediction Capability Using Fuel Assembly Benchmark

International Nuclear Information System (INIS)

Jecmenica, R.; Pevec, D.; Grgic, D.; Konjarek, D.

2008-01-01

FA2D is 2D transport collision probability code developed at Faculty of Electrical Engineering and Computing, University Zagreb. It is used for calculation of cross section data at fuel assembly level. Main objective of its development was capability to generate cross section data to be used for fuel management and safety analyses of PWR reactors. Till now formal verification of code predictions capability is not performed at fuel assembly level, but results of fuel management calculations obtained using FA2D generated cross sections for NPP Krsko and IRIS reactor are compared against Westinghouse calculations. Cross section data were used within NRC's PARCS code and satisfactory preliminary results were obtained. This paper presents results of calculations performed for Nuclear Fuel Industries, Ltd., benchmark using FA2D, and SCALE5 TRITON calculation sequence (based on discrete ordinates code NEWT). Nuclear Fuel Industries, Ltd., Japan, released LWR Next Generation Fuels Benchmark with the aim to verify prediction capability in nuclear design for extended burnup regions. We performed calculations for two different Benchmark problem geometries - UO 2 pin cell and UO 2 PWR fuel assembly. The results obtained with two mentioned 2D spectral codes are presented for burnup dependency of infinite multiplication factor, isotopic concentration of important materials and for local peaking factor vs. burnup (in case of fuel assembly calculation).(author)

HAMMER, 1-D Multigroup Neutron Transport Infinite System Cell Calculation for Few-Group Diffusion Calculation

International Nuclear Information System (INIS)

Honeck, H.C.

1984-01-01

1 - Description of problem or function: HAMMER performs infinite lattice, one-dimensional cell multigroup calculations, followed (optionally) by one-dimensional, few-group, multi-region reactor calculations with neutron balance edits. 2 - Method of solution: Infinite lattice parameters are calculated by means of multigroup transport theory, composite reactor parameters by few-group diffusion theory. 3 - Restrictions on the complexity of the problem: - Cell calculations - maxima of: 30 thermal groups; 54 epithermal groups; 20 space points; 20 regions; 18 isotopes; 10 mixtures; 3 thermal up-scattering mixtures; 200 resonances per group; no overlap or interference; single level only. - Reactor calculations - maxima of : 40 regions; 40 mixtures; 250 space points; 4 groups

Transport calculation of medium-energy protons and neutrons by Monte Carlo method

International Nuclear Information System (INIS)

Ban, Syuuichi; Hirayama, Hideo; Katoh, Kazuaki.

1978-09-01

A Monte Carlo transport code, ARIES, has been developed for protons and neutrons at medium energy (25 -- 500 MeV). Nuclear data provided by R.G. Alsmiller, Jr. were used for the calculation. To simulate the cascade development in the medium, each generation was represented by a single weighted particle and an average number of emitted particles was used as the weight. Neutron fluxes were stored by the collisions density method. The cutoff energy was set to 25 MeV. Neutrons below the cutoff were stored to be used as the source for the low energy neutron transport calculation upon the discrete ordinates method. Then transport calculations were performed for both low energy neutrons (thermal -- 25 MeV) and secondary gamma-rays. Energy spectra of emitted neutrons were calculated and compared with those of published experimental and calculated results. The agreement was good for the incident particles of energy between 100 and 500 MeV. (author)

Investigation on the MOC with a linear source approximation scheme in three-dimensional assembly

International Nuclear Information System (INIS)

Zhu, Chenglin; Cao, Xinrong

2014-01-01

Method of characteristics (MOC) for solving neutron transport equation has already become one of the fundamental methods for lattice calculation of nuclear design code system. At present, MOC has three schemes to deal with the neutron source of the transport equation: the flat source approximation of the step characteristics (SC) scheme, the diamond difference (DD) scheme and the linear source (LS) characteristics scheme. The MOC for SC scheme and DD scheme need large storage space and long computing time when they are used to calculate large-scale three-dimensional neutron transport problems. In this paper, a LS scheme and its correction for negative source distribution were developed and added to DRAGON code. This new scheme was compared with the SC scheme and DD scheme which had been applied in this code. As an open source code, DRAGON could solve three-dimensional assembly with MOC method. Detailed calculation is conducted on two-dimensional VVER-1000 assembly under three schemes of MOC. The numerical results indicate that coarse mesh could be used in the LS scheme with the same accuracy. And the LS scheme applied in DRAGON is effective and expected results are achieved. Then three-dimensional cell problem and VVER-1000 assembly are calculated with LS scheme and SC scheme. The results show that less memory and shorter computational time are employed in LS scheme compared with SC scheme. It is concluded that by using LS scheme, DRAGON is able to calculate large-scale three-dimensional problems with less storage space and shorter computing time

Lagrangian Transport Calculations Using UARS Data. Part 2; Ozone

Science.gov (United States)

Manney, Gloria L.; Zurek, R. W.; Froidevaux, L.; Waters, J. W.; ONeill, A.; Swinbank, R.

1995-01-01

Trajectory calculations are used to examine ozone transport in the polar winter stratosphere during periods of the Upper Atmosphere Research Satellite (UARS) observations. The value of these calculations for determining mass transport was demonstrated previously using UARS observations of long-lived tracers, In the middle stratosphere, the overall ozone behavior observed by the Microwave Limb Sounder in the polar vortex is reproduced by this purely dynamical model. Calculations show the evolution of ozone in the lower stratosphere during early winter to be dominated by dynamics in December 1992 in the Arctic. Calculations for June 1992 in the Antarctic show evidence of chemical ozone destruction and indicate that approx. 50% of the chemical destruction may be masked by dynamical effects, mainly diabatic descent, which bring higher ozone into the lower-stratospheric vortex. Estimating differences between calculated and observed fields suggests that dynamical changes masked approx. 20% - 35% of chemical ozone loss during late February and early March 1993 in the Arctic. In the Antarctic late winter, in late August and early September 1992, below approx. 520 K, the evolution of vortex-averaged ozone is entirely dominated by chemical effects; above this level, however, chemical ozone depletion can be partially or completely masked by dynamical effects. Our calculations for 1992 showed that chemical loss was nearly completely compensated by increases due to diabatic descent at 655 K.

Initial Validation of Robotic Operations for In-Space Assembly of a Large Solar Electric Propulsion Transport Vehicle

Science.gov (United States)

Komendera, Erik E.; Dorsey, John T.

2017-01-01

Developing a capability for the assembly of large space structures has the potential to increase the capabilities and performance of future space missions and spacecraft while reducing their cost. One such application is a megawatt-class solar electric propulsion (SEP) tug, representing a critical transportation ability for the NASA lunar, Mars, and solar system exploration missions. A series of robotic assembly experiments were recently completed at Langley Research Center (LaRC) that demonstrate most of the assembly steps for the SEP tug concept. The assembly experiments used a core set of robotic capabilities: long-reach manipulation and dexterous manipulation. This paper describes cross-cutting capabilities and technologies for in-space assembly (ISA), applies the ISA approach to a SEP tug, describes the design and development of two assembly demonstration concepts, and summarizes results of two sets of assembly experiments that validate the SEP tug assembly steps.

Modeling Dynamic Objects in Monte Carlo Particle Transport Calculations

International Nuclear Information System (INIS)

Yegin, G.

2008-01-01

In this study, the Multi-Geometry geometry modeling technique was improved in order to handle moving objects in a Monte Carlo particle transport calculation. In the Multi-Geometry technique, the geometry is a superposition of objects not surfaces. By using this feature, we developed a new algorithm which allows a user to make enable or disable geometry elements during particle transport. A disabled object can be ignored at a certain stage of a calculation and switching among identical copies of the same object located adjacent poins during a particle simulation corresponds to the movement of that object in space. We called this powerfull feature as Dynamic Multi-Geometry technique (DMG) which is used for the first time in Brachy Dose Monte Carlo code to simulate HDR brachytherapy treatment systems. Our results showed that having disabled objects in a geometry does not effect calculated dose values. This technique is also suitable to be used in other areas such as IMRT treatment planning systems

Calculational modeling of fuel assemblies of WWER-1000 type with the use of burnable absorber Gadolinum; comparative analysis

International Nuclear Information System (INIS)

Yeremenko, M.L.; Kovbasenko, Yu.P.; Loetsch, T.

2001-01-01

In connection with the beginning of the use of fuel assemblies with burnable absorbers by integration of Gadolinum into the nuclear fuel at Ukrainian NPP the task of testing the code systems and the pertinent neutron cross section libraries for the new fuel arose. Taking into account the long term experience of German experts with calculations and evaluation of nuclear fuel containing Gadolinum it was decided to carry out a series of test calculations for fuel assembly lattices of PWR, WWER-440 and WWER-1000 types using the NESSEL/PYTHIA and CASMO/SIMULATE code systems (Authors)

One approach to accepting and transporting spent fuel from early-generation reactors with short fuel assemblies

International Nuclear Information System (INIS)

Peterson, R.W.; Bentz, E.J. Jr.; Bentz, C.B.

1993-01-01

In the early days of development of commercial nuclear power reactors in the U.S., the overall length and uranium loading of the fuel assemblies were considerably less than those of later generation facilities. In turn, some of these early facilities were designed for handling shorter casks than currently-certified casks. The spent fuel assemblies from these facilities are nearly all standard fuel within the definition in the Standard Contract (10 CFR 961) between the utilities and the U.S. Department of Energy (DOE) (the Big Rock Point fuel cross-section is outside the standard fuel dimension), and the utilities involved hold early delivery rights under DOE's oldest-fuel-first (OFF) allocation scenario. However, development of casks suitable for satisfying the acceptance and transportation requirements of some of these facilities is not currently underway in the DOE Cask System Development Program (CSDP). While the total MTU of these fuels is relatively small compared to the total program, the number of assemblies to be transported is significant, especially in the early years of operation according to the OFF allocation scenario. We therefore perceive a need for DOE to develop an approach and to implement plans to satisfy the unique acceptance and transportation requirements of these facilities. One such approach is outlined below. (author)

Transportation channels calculation method in MATLAB

International Nuclear Information System (INIS)

Averyanov, G.P.; Budkin, V.A.; Dmitrieva, V.V.; Osadchuk, I.O.; Bashmakov, Yu.A.

2014-01-01

Output devices and charged particles transport channels are necessary components of any modern particle accelerator. They differ both in sizes and in terms of focusing elements depending on particle accelerator type and its destination. A package of transport line designing codes for magnet optical channels in MATLAB environment is presented in this report. Charged particles dynamics in a focusing channel can be studied easily by means of the matrix technique. MATLAB usage is convenient because its information objects are matrixes. MATLAB allows the use the modular principle to build the software package. Program blocks are small in size and easy to use. They can be executed separately or commonly. A set of codes has a user-friendly interface. Transport channel construction consists of focusing lenses (doublets and triplets). The main of the magneto-optical channel parameters are total length and lens position and parameters of the output beam in the phase space (channel acceptance, beam emittance - beam transverse dimensions, particles divergence and image stigmaticity). Choice of the channel operation parameters is based on the conditions for satisfying mutually competing demands. And therefore the channel parameters calculation is carried out by using the search engine optimization techniques.

Calculations on safe storage and transportation of radioactive materials

Energy Technology Data Exchange (ETDEWEB)

Hathout, A M; El-Messiry, A M; Amin, E [National Center for Nuclear Safety and Radiation Control and AEA, Cairo (Egypt)

1997-12-31

In this work the safe storage and transportation of fresh fuel as a radioactive material studied. Egypt planned ET RR 2 reactor which is of relatively high power and would require adequate handling and transportation. Therefore, the present work is initiated to develop a procedure for safe handling and transportation of radioactive materials. The possibility of reducing the magnitude of radiation transmitted on the exterior of the packages is investigated. Neutron absorbers are used to decrease the neutron flux. Criticality calculations are carried out to ensure the achievement of subcriticality so that the inherent safety can be verified. The discrete ordinate transport code ANISN was used. The results show good agreement with other techniques. 2 figs., 2 tabs.

Accounting for chemical kinetics in field scale transport calculations

International Nuclear Information System (INIS)

Bryan, N.D.

2005-01-01

The modelling of column experiments has shown that the humic acid mediated transport of metal ions is dominated by the non-exchangeable fraction. Metal ions enter this fraction via the exchangeable fraction, and may transfer back again. However, in both directions these chemical reactions are slow. Whether or not a kinetic description of these processes is required during transport calculations, or an assumption of local equilibrium will suffice, will depend upon the ratio of the reaction half-time to the residence time of species within the groundwater column. If the flow rate is sufficiently slow or the reaction sufficiently fast then the assumption of local equilibrium is acceptable. Alternatively, if the reaction is sufficiently slow (or the flow rate fast), then the reaction may be 'decoupled', i.e. removed from the calculation. These distinctions are important, because calculations involving chemical kinetics are computationally very expensive, and should be avoided wherever possible. In addition, column experiments have shown that the sorption of humic substances and metal-humate complexes may be significant, and that these reactions may also be slow. In this work, a set of rules is presented that dictate when the local equilibrium and decoupled assumptions may be used. In addition, it is shown that in all cases to a first approximation, the behaviour of a kinetically controlled species, and in particular its final distribution against distance at the end of a calculation, depends only upon the ratio of the reaction first order rate to the residence time, and hence, even in the region where the simplifications may not be used, the behaviour is predictable. In this way, it is possible to obtain an estimate of the migration of these species, without the need for a complex transport calculation. (orig.)

Experience from transportation of irradiated WWER-440 fuel assemblies at Kozloduy NPP site after a short cooling time

International Nuclear Information System (INIS)

Stoyanova, I.; Kamenov, A.; Byrzev, L.; Christoskov, I.

2003-01-01

The presented results from the computation and analysis of the radiation characteristics of the irradiated fuel assemblies by the date of their transportation according to the selected loading patterns of the VSPOT cask and following the modified technology of transportation, i.e. without replacement of the pool solution by pure condensate, as well as the corresponding experimental results, confirm the applicability of the newly introduced safety criterion for the selection of a loading pattern of the cask with irradiated fuel assemblies after a short cooling time. The comparison between measured and computed surface dose rates shows that during the procedure of transfer of irradiated fuel assemblies from the pools of Units 1 and 2 to the pools of Kozloduy NPP Units 3 and 4 all safety limits, incl. the radiation protection requirements, were met

Uniform Gauss-Weight Quadratures for Discrete Ordinate Transport Calculations

International Nuclear Information System (INIS)

Carew, John F.; Hu, Kai; Zamonsky, Gabriel

2000-01-01

Recently, a uniform equal-weight quadrature set, UE n , and a uniform Gauss-weight quadrature set, UG n , have been derived. These quadratures have the advantage over the standard level-symmetric LQ n quadrature sets in that the weights are positive for all orders,and the transport solution may be systematically converged by increasing the order of the quadrature set. As the order of the quadrature is increased,the points approach a uniform continuous distribution on the unit sphere,and the quadrature is invariant with respect to spatial rotations. The numerical integrals converge for continuous functions as the order of the quadrature is increased.The numerical characteristics of the UE n quadrature set have been investigated previously. In this paper, numerical calculations are performed to evaluate the application of the UG n quadrature set in typical transport analyses. A series of DORT transport calculations of the >1-MeV neutron flux have been performed for a set of pressure-vessel fluence benchmark problems. These calculations employed the UG n (n = 8, 12, 16, 24, and 32) quadratures and indicate that the UG n solutions have converged to within ∼0.25%. The converged UG n solutions are found to be comparable to the UE n results and are more accurate than the level-symmetric S 16 predictions

Evaluation of dose equivalent rate distribution in JCO critical accident by radiation transport calculation

CERN Document Server

Sakamoto, Y

2002-01-01

In the prevention of nuclear disaster, there needs the information on the dose equivalent rate distribution inside and outside the site, and energy spectra. The three dimensional radiation transport calculation code is a useful tool for the site specific detailed analysis with the consideration of facility structures. It is important in the prediction of individual doses in the future countermeasure that the reliability of the evaluation methods of dose equivalent rate distribution and energy spectra by using of Monte Carlo radiation transport calculation code, and the factors which influence the dose equivalent rate distribution outside the site are confirmed. The reliability of radiation transport calculation code and the influence factors of dose equivalent rate distribution were examined through the analyses of critical accident at JCO's uranium processing plant occurred on September 30, 1999. The radiation transport calculations including the burn-up calculations were done by using of the structural info...

Dynamic Load Balancing of Parallel Monte Carlo Transport Calculations

International Nuclear Information System (INIS)

O'Brien, M; Taylor, J; Procassini, R

2004-01-01

The performance of parallel Monte Carlo transport calculations which use both spatial and particle parallelism is increased by dynamically assigning processors to the most worked domains. Since the particle work load varies over the course of the simulation, this algorithm determines each cycle if dynamic load balancing would speed up the calculation. If load balancing is required, a small number of particle communications are initiated in order to achieve load balance. This method has decreased the parallel run time by more than a factor of three for certain criticality calculations

A retrospective and prospective survey of three-dimensional transport calculations

International Nuclear Information System (INIS)

Nakahara, Yasuaki

1985-01-01

A retrospective survey is made on the three-dimensional radiation transport calculations. Introduction is given to computer codes based on the distinctive numerical methods such as the Monte Carlo, Direct Integration, Ssub(n) and Finite Element Methods to solve the three-dimensional transport equations. Prospective discussions are made on pros and cons of these methods. (author)

JNC results of BN-600 benchmark calculation (phase 4)

International Nuclear Information System (INIS)

Ishikawa, Makoto

2003-01-01

The present work is the results of JNC, Japan, for the Phase 4 of the BN-600 core benchmark problem (Hex-Z fully MOX fuelled core model) organized by IAEA. The benchmark specification is based on 1) the RCM report of IAEA CRP on 'Updated Codes and Methods to Reduce the Calculational Uncertainties of LMFR Reactivity Effects, Action 3.12' (Calculations for BN-600 fully fuelled MOX core for subsequent transient analyses). JENDL-3.2 nuclear data library was used for calculating 70 group ABBN-type group constants. Cell models for fuel assembly and control rod calculations were applied: homogeneous and heterogeneous (cylindrical supercell) model. Basic diffusion calculation was three-dimensional Hex-Z model, 18 group (Citation code). Transport calculations were 18 group, three-dimensional (NSHEC code) based on Sn-transport nodal method developed at JNC. The generated thermal power per fission was based on Sher's data corrected on the basis of ENDF/B-IV data library. Calculation results are presented in Tables for intercomparison

Reference calculations on critical assemblies with Apollo2 code working with a fine multigroup mesh; Calculs de reference avec un maillage multigroupe fin sur des assemblages critiques par Apollo2

Energy Technology Data Exchange (ETDEWEB)

Aggery, A

1999-12-01

The objective of this thesis is to add to the multigroup transport code APOLLO2 the capability to perform deterministic reference calculations, for any type of reactor, using a very fine energy mesh of several thousand groups. This new reference tool allows us to validate the self-shielding model used in industrial applications, to perform depletion calculations, differential effects calculations, critical buckling calculations or to evaluate precisely data required by the self shielding model. At its origin, APOLLO2 was designed to perform routine calculations with energy meshes around one hundred groups. That is why, in the current format of cross sections libraries, almost each value of the multigroup energy transfer matrix is stored. As this format is not convenient for a high number of groups (concerning memory size), we had to search out a new format for removal matrices and consequently to modify the code. In the new format we found, only some values of removal matrices are kept (these values depend on a reconstruction precision choice), the other ones being reconstructed by a linear interpolation, what reduces the size of these matrices. Then we had to show that APOLLO2 working with a fine multigroup mesh had the capability to perform reference calculations on any assembly geometry. For that, we successfully carried out the validation with several calculations for which we compared APOLLO2 results (obtained with the universal mesh of 11276 groups) to results obtained with Monte Carlo codes (MCNP, TRIPOLI4). Physical analysis led with this new tool have been very fruitful and show a great potential for such an R and D tool. (author)

Calculation of neutron and gamma transport at the FOA:type of problems and calculation methods

International Nuclear Information System (INIS)

Lefvert, T.

1975-11-01

Protection against the effects of nuclear warfare involves the analysis of the forms of results of a nuclear charge explosion producing neutron and gamma radiation. It brings out problems leading to the calculation of criticality, leakage, and deep transmission. Methods have been developed for various kinds of particle transport problems. Applications to radiation therapy, storage of fissile materials, and fast reactors are discussed. A list (with brief description) of all neutron and gamma transport programmes of the FOA is given. (J.S.)

Mechanical and thermomechanical calculations related to the storage of spent nuclear-fuel assemblies in granite

International Nuclear Information System (INIS)

Butkovich, T.R.

1980-05-01

A generic test of the geologic storage of spent-fuel assemblies is being made at Nevada Test Site. The spent-fuel assemblies were emplaced at a depth of 420 m (1370 ft) below the surface in a typical granite and will be retrieved at a later time. The early time, close-in thermal history of this type of repository is being simulated with spent-fuel and electrically heated canisters in a central drift, with auxiliary heaters in two parallel side drifts. Prior to emplacement of the spent-fuel canisters, preliminary calculations were made using a pair of existing finite-element codes, ADINA and ADINAT

The VEGA Assembly Spectrum Code

International Nuclear Information System (INIS)

Milosevic, M.

1997-01-01

The VEGA is assembly spectrum code, developed as a design tool for producing a few-group averaged cross section data for a wide range of reactor types including both thermal and fast reactors. It belongs to a class of codes, which may be characterized by the separate stages for micro group, spectrum and macro group assembly calculations. The theoretical foundation for the development of the VEGA code was integral transport theory in the first-flight collision probability formulation. Two versions of VEGA are now in use, VEGA-1 established on standard equivalence theory and VEGA-2 based on new subgroup method applicable for any geometry for which a flux solution is possible. This paper describes a features which are unique to the VEGA codes with concentration on the basic principles and algorithms used in the proposed subgroup method. Presented validation of this method, comprise the results for a homogenous uranium-plutonium mixture and a PWR cell containing a recycled uranium-plutonium oxide. Example application for a realistic fuel dissolver benchmark problem , which was extensive analyzed in the international calculations, is also included. (author)

Ultraviolet Light Generation and Transport in the Final Optics Assembly of the National Ignition Facility

Energy Technology Data Exchange (ETDEWEB)

Wegner, P. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Hackel, L. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Feit, M. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Parham, T. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Kozlowski, M. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Whitman, P. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

2015-02-12

The design of the National Ignition Facility (NIF) includes a Final Optics Assembly (FOA) subsystem for ultraviolet (UV) light generation and transport for each of the 192 beamlines. Analytical and experimental work has been done to help understand and predict the performance of FOA.

Effect of multicell DRAGON calculations depends on the environment on the DONJON predictions for the ACR-1000

International Nuclear Information System (INIS)

Duquette, J.-S.

2009-01-01

For understanding the behavior of a nuclear reactor core, it is necessary to make a full core calculation in order to compute the neutrons flux. To obtain the neutrons flux, solving the Boltzmann transport equation is required. That is not a simple task and it is impossible to analytically fend the solution of the neutrons transport equation on a complex core. Following a series of approximations, it is possible to numerically solve the neutrons transport equation. The solution of this equation is done step by step. Calculations will be performed over the ACR-1000 core. The Advanced CANDU Reactor (ACR-1000) is a generation III+ heavy water moderated and light water cooled reactor. It is a 1200 MW(e) power reactor. Amongst the ACR-1000 design parameters that differ from the CANDU 6, the reduced lattice pitch and the use of light water coolant and enriched fuel are the three most important. Those features modify the behavior of the neutrons in the ACR compared to the CANDU 6. The impact of the tight lattice is that a cell is more strongly coupled to its neighbor. The objective of this work is to determine the impact of the environment on the cell properties of the ACR-1000. Those properties will be used to perform full core calculations. The neutron transport calculations are performed with DRAGON whereas for the diffusion calculation on a full core. The code DONJON will be used. The DRAGON reference transport calculation will be made on a single cell. Then, a series of calculations will be performed using DRAGON over two types of assemblies, the first modelling the core interior and the second, modelling the core periphery. Moreover, the fuel age will sometimes be homogeneous, sometimes heterogeneous. The fuel will be burned during six hundred days. One thus obtains libraries of macroscopic cross sections over a six hundred days interval for various simulations. Thereafter, we will determine the effect of a neutrons transport multicell calculation on various DONJON

The APOLLO assembly spectrum code

International Nuclear Information System (INIS)

Kavenoky, A.; Sanchez, R.

1987-04-01

The APOLLO code was originally developed as a design tool for HTR's, later it was aimed at the calculation of PWR lattices. APOLLO is a general purpose assembly spectrum code based on the multigroup integral transport equation; refined collision probability modules allow the computation of 1D geometries with linearly anisotropic scattering and two term flux expansion. In 2D geometries modules based on the substructure method provide fast and accurate design calculations and a module based on a direct discretization is devoted to reference calculations. The SPH homogenization technique provides corrected cross sections performing an equivalence between coarse and refined calculations. The post processing module of APOLLO generate either APOLLIB to be used by APOLLO or NEPLIB for reactor diffusion calculation. The cross section library of APOLLO contains data and self-shielding data for more than 400 isotopes. APOLLO is able to compute the depletion of any medium accounting for any heavy isotope or fission product chain. 21 refs

Verification of homogenization in fast critical assembly analyses

International Nuclear Information System (INIS)

Chiba, Go

2006-01-01

In the present paper, homogenization procedures for fast critical assembly analyses are investigated. Errors caused by homogenizations are evaluated by the exact perturbation theory. In order to obtain reference solutions, three-dimensional plate-wise transport calculations are performed. It is found that the angular neutron flux along plate boundaries has a significant peak in the fission source energy range. To treat this angular dependence accurately, the double-Gaussian Chebyshev angular quadrature set with S 24 is applied. It is shown that the difference between the heterogeneous leakage theory and the homogeneous theory is negligible, and that transport cross sections homogenized with neutron flux significantly underestimate neutron leakage. The error in criticality caused by a homogenization is estimated at about 0.1%Δk/kk' in a small fast critical assembly. In addition, the neutron leakage is overestimated by both leakage theories when sodium plates in fuel lattices are voided. (author)

Comparison calculations of WWER-1000 fuel assemblies by using the MCNP 4.2 a KASSETA codes

International Nuclear Information System (INIS)

Trgina, M.

1993-12-01

The power multiplication and distribution factors are compared for various geometries and material configurations of WWER-1000 fuel assemblies. The calculations were performed in 2 ways: (i) using nuclear data, employing older and current data collections, and (ii) using the author's own model based on the KASSETA code. The comparison code MCNP 4.2 is described, intended for computerized simulation of the transport of neutrons, photons and electrons. This code uses its own cross section library. The methodology is outlined and a specification of the Monte Carlo method employed is given. The use of the refined data library gave rise to appreciable deviations of the multiplication factors in all variants. The use of the older data library led to identical criticality results for the variant with water holes. For inserted absorbers the discrepancies in criticality and in power distribution data are appreciable. The marked disagreement between the results of application of the MCNP 4.2 and KASSETA codes for the variants with inserted control elements is indicative of inappropriateness of the approximation procedure in the latter code. (J.B.). 2 tabs., 11 figs., 11 refs

Existing experimental criticality data applicable to nuclear-fuel-transportation systems

International Nuclear Information System (INIS)

Bierman, S.R.

1983-02-01

Analytical techniques are generally relied upon in making criticality evaluations involving nuclear material outside reactors. For these evaluations to be accepted the calculations must be validated by comparison with experimental data for a known set of conditions having physical and neutronic characteristics similar to those conditions being evaluated analytically. The purpose of this report is to identify those existing experimental data that are suitable for use in verifying criticality calculations on nuclear fuel transportation systems. In addition, near term needs for additional data in this area are identified. Of the considerable amount of criticality data currently existing, that are applicable to non-reactor systems, those particularly suitable for use in support of nuclear material transportation systems have been identified and catalogued into the following groups: (1) critical assemblies of fuel rods in water; (2) critical assemblies of fuel rods in water containing soluble neutron absorbers; (3) critical assemblies containing solid neutron absorber; (4) critical assemblies of fuel rods in water with heavy metal reflectors; and (5) critical assemblies of fuel rods in water with irregular features. A listing of the current near term needs for additional data in each of the groups has been developed for future use in planning criticality research in support of nuclear fuel transportation systems. The criticality experiments needed to provide these data are briefly described and identified according to priority and relative cost of performing the experiments

Rail-Cask Tests: Normal-Conditionsof- Transport Tests of Surrogate PWR Fuel Assemblies in an ENSA ENUN 32P Cask.

Energy Technology Data Exchange (ETDEWEB)

McConnell, Paul E. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Ross, Steven [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Grey, Carissa Ann [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Uncapher, William Leonard [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Arviso, Michael [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Garmendia, Rafael [Equipos Nucleares Sociedad Anonima, Madrid (Spain); Fernandez Perez, Ismael [Equipos Nucleares Sociedad Anonima, Madrid (Spain); Palacio, Alejandro [Equipos Nucleares Sociedad Anonima, Madrid (Spain); Calleja, Guillermo [Equipos Nucleares Sociedad Anonima, Madrid (Spain); Garrido, David [COORDINADORA, Madrid (Spain); Rodriguez Casas, Ana [COORDINADORA, Madrid (Spain); Gonzalez Garcia, Luis [COORDINADORA, Madrid (Spain); Chilton, Lyman Wes [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Ammerman, Douglas J. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Walz, Jacob [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Gershon, Sabina [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Saltzstein, Sylvia J. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Sorenson, Ken [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Klymyshyn, Nicholas [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Hanson, Brady [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Pena, Ruben [Transportation Technology Center, Inc., Pueblo, CO (United States); Walker, Russell [Transportation Technology Center, Inc., Pueblo, CO (United States)

2018-01-01

This report describes tests conducted using a full-size rail cask, the ENSA ENUN 32P, involving handling of the cask and transport of the cask via truck, ships, and rail. The purpose of the tests was to measure strains and accelerations on surrogate pressurized water reactor fuel rods when the fuel assemblies were subjected to Normal Conditions of Transport within the rail cask. In addition, accelerations were measured on the transport platform, the cask cradle, the cask, and the basket within the cask holding the assemblies. These tests were an international collaboration that included Equipos Nucleares S.A., Sandia National Laboratories, Pacific Northwest National Laboratory, Coordinadora Internacional de Cargas S.A., the Transportation Technology Center, Inc., the Korea Radioactive Waste Agency, and the Korea Atomic Energy Research Institute. All test results in this report are PRELIMINARY – complete analyses of test data will be completed and reported in FY18. However, preliminarily: The strains were exceedingly low on the surrogate fuel rods during the rail-cask tests for all the transport and handling modes. The test results provide a compelling technical basis for the safe transport of spent fuel.

Performing three-dimensional neutral particle transport calculations on tera scale computers

International Nuclear Information System (INIS)

Woodward, C.S.; Brown, P.N.; Chang, B.; Dorr, M.R.; Hanebutte, U.R.

1999-01-01

A scalable, parallel code system to perform neutral particle transport calculations in three dimensions is presented. To utilize the hyper-cluster architecture of emerging tera scale computers, the parallel code successfully combines the MPI message passing and paradigms. The code's capabilities are demonstrated by a shielding calculation containing over 14 billion unknowns. This calculation was accomplished on the IBM SP ''ASCI-Blue-Pacific computer located at Lawrence Livermore National Laboratory (LLNL)

Self-assembled peptide nanotubes as electronic materials: An evaluation from first-principles calculations

International Nuclear Information System (INIS)

Akdim, Brahim; Pachter, Ruth; Naik, Rajesh R.

2015-01-01

In this letter, we report on the evaluation of diphenylalanine (FF), dityrosine (YY), and phenylalanine-tryptophan (FW) self-assembled peptide nanotube structures for electronics and photonics applications. Realistic bulk peptide nanotube material models were used in density functional theory calculations to mimic the well-ordered tubular nanostructures. Importantly, validated functionals were applied, specifically by using a London dispersion correction to model intertube interactions and a range-separated hybrid functional for accurate bandgap calculations. Bandgaps were found consistent with available experimental data for FF, and also corroborate the higher conductance reported for FW in comparison to FF peptide nanotubes. Interestingly, the predicted bandgap for the YY tubular nanostructure was found to be slightly higher than that of FW, suggesting higher conductance as well. In addition, the band structure calculations along the high symmetry line of nanotube axis revealed a direct bandgap for FF. The results enhance our understanding of the electronic properties of these material systems and will pave the way into their application in devices

Optimized shielding calculation to the transport of 131I employed in nuclear medicine

International Nuclear Information System (INIS)

Sahyun, A.; Sordi, G.M.; Rodrigues, D.; Sanches, M.P.; Romero F, C.R.

1996-01-01

The objective of this paper is to present the basis for shielding calculation used in different situations that could occur during the transport of 131 I utilized in nuclear medicine for diagnostic and therapeutic purposes. The aim of these calculation is to optimize the shielding in order to satisfy the transport of radioactive material. These calculations were proposed for estimated activities around 1,85 GBq (50mCi), 3,7 GBq(100mCi) and 7,4 GBq(200mCi), considering the driver of the cargo company and his assistant as the critical group and the general people considered as effect of collective dose. The population density considered in the models is the one related to Sao Paulo city, because the transport is done by the highway across the city and the radioactive material is distributed from west to north and south, where the airports are located. This area ranges a perimeter of 40 km. For the collective dose calculation, it was considered a population dose of less than 1/100 of the annual limit dose for the public. Our main concern is related to the large volume of radioactive material that is transported per week, specially because 1/3 of this material has activities around 3,7 GBq (100mCi). During the calculations, we have figured out that the activities at the moment of transport are nearly 40% greater than the one related to the calibration date. As for the discrepancy of official alpha value of US$10000/man-Sv and the real value for our country of US$3000/man-Sv,a comparative study was performed. (authors). 3 refs., 2 figs., 2 tabs

Optical photon transport in powdered-phosphor scintillators. Part II. Calculation of single-scattering transport parameters

Energy Technology Data Exchange (ETDEWEB)

Poludniowski, Gavin G. [Joint Department of Physics, Division of Radiotherapy and Imaging, Institute of Cancer Research and Royal Marsden NHS Foundation Trust, Downs Road, Sutton, Surrey SM2 5PT, United Kingdom and Centre for Vision Speech and Signal Processing (CVSSP), Faculty of Engineering and Physical Sciences, University of Surrey, Guildford, Surrey GU2 7XH (United Kingdom); Evans, Philip M. [Centre for Vision Speech and Signal Processing (CVSSP), Faculty of Engineering and Physical Sciences, University of Surrey, Guildford, Surrey GU2 7XH (United Kingdom)

2013-04-15

Purpose: Monte Carlo methods based on the Boltzmann transport equation (BTE) have previously been used to model light transport in powdered-phosphor scintillator screens. Physically motivated guesses or, alternatively, the complexities of Mie theory have been used by some authors to provide the necessary inputs of transport parameters. The purpose of Part II of this work is to: (i) validate predictions of modulation transform function (MTF) using the BTE and calculated values of transport parameters, against experimental data published for two Gd{sub 2}O{sub 2}S:Tb screens; (ii) investigate the impact of size-distribution and emission spectrum on Mie predictions of transport parameters; (iii) suggest simpler and novel geometrical optics-based models for these parameters and compare to the predictions of Mie theory. A computer code package called phsphr is made available that allows the MTF predictions for the screens modeled to be reproduced and novel screens to be simulated. Methods: The transport parameters of interest are the scattering efficiency (Q{sub sct}), absorption efficiency (Q{sub abs}), and the scatter anisotropy (g). Calculations of these parameters are made using the analytic method of Mie theory, for spherical grains of radii 0.1-5.0 {mu}m. The sensitivity of the transport parameters to emission wavelength is investigated using an emission spectrum representative of that of Gd{sub 2}O{sub 2}S:Tb. The impact of a grain-size distribution in the screen on the parameters is investigated using a Gaussian size-distribution ({sigma}= 1%, 5%, or 10% of mean radius). Two simple and novel alternative models to Mie theory are suggested: a geometrical optics and diffraction model (GODM) and an extension of this (GODM+). Comparisons to measured MTF are made for two commercial screens: Lanex Fast Back and Lanex Fast Front (Eastman Kodak Company, Inc.). Results: The Mie theory predictions of transport parameters were shown to be highly sensitive to both grain size

Time-dependent Flow and Transport Calculations for Project Opalinus Clay (Entsorgungsnachweis)

International Nuclear Information System (INIS)

Kosakowski, G.

2004-07-01

This report describes two specific assessment cases used in the safety assessment for a proposed deep geological repository for spent fuel, high level waste and long-lived intermediate-level waste, sited in the Opalinus Clay of the Zuercher Weinland in northern Switzerland (Project Entsorgungsnachweis, NAG RA, 2002d). In this study the influence of time dependent flow processes on the radionuclide transport in the geosphere is investigated. In the Opalinus Clay diffusion dominates the transport of radionuclides, but processes exist that can locally increase the importance of the advective transport for some time. Two important cases were investigated: (1) glaciation-induced flow due to an additional overburden in the form of an ice shield of up to 400 m thickness and (2) fluid flow driven by tunnel convergence. For the calculations the code FRAC3DVS (Therrien and Sudicky, 1996) was used. FRAC3DVS solves the three-dimensional flow and transport equation in porous and fractured media. For the case of glaciation-induced flow (1) a two-dimensional reference model without glaciations was calculated. During the glaciations the geosphere release-rates are up to a factor of about 1.7 higher compared to the reference model. The influence of glaciations on the transport of cations or neutral species is less than for anions, since the importance of the advective transport for anions is higher due to the lower accessible porosity for anions. The increase in the release rates during glaciations is lower for sorbing compared to non-sorbing radionuclides. The influence of the tunnel convergence (2) on the transport of radionuclides in the geosphere is very small. Due to the higher source term the geosphere release rates are slightly higher if tunnel convergence is considered. In addition to the two assessment cases this report investigates the applicability of the one-dimensional approximation for modelling transport through the Opalinus Clay. For the reference case of the safety

Axial SPN and radial MOC coupled whole core transport calculation

International Nuclear Information System (INIS)

Cho, Jin-Young; Kim, Kang-Seog; Lee, Chung-Chan; Zee, Sung-Quun; Joo, Han-Gyu

2007-01-01

The Simplified P N (SP N ) method is applied to the axial solution of the two-dimensional (2-D) method of characteristics (MOC) solution based whole core transport calculation. A sub-plane scheme and the nodal expansion method (NEM) are employed for the solution of the one-dimensional (1-D) SP N equations involving a radial transverse leakage. The SP N solver replaces the axial diffusion solver of the DeCART direct whole core transport code to provide more accurate, transport theory based axial solutions. In the sub-plane scheme, the radial equivalent homogenization parameters generated by the local MOC for a thick plane are assigned to the multiple finer planes in the subsequent global three-dimensional (3-D) coarse mesh finite difference (CMFD) calculation in which the NEM is employed for the axial solution. The sub-plane scheme induces a much less nodal error while having little impact on the axial leakage representation of the radial MOC calculation. The performance of the sub-plane scheme and SP N nodal transport solver is examined by solving a set of demonstrative problems and the C5G7MOX 3-D extension benchmark problems. It is shown in the demonstrative problems that the nodal error reaching upto 1,400 pcm in a rodded case is reduced to 10 pcm by introducing 10 sub-planes per MOC plane and the transport error is reduced from about 150 pcm to 10 pcm by using SP 3 . Also it is observed, in the C5G7MOX rodded configuration B problem, that the eigenvalues and pin power errors of 180 pcm and 2.2% of the 10 sub-planes diffusion case are reduced to 40 pcm and 1.4%, respectively, for SP 3 with only about a 15% increase in the computing time. It is shown that the SP 5 case gives very similar results to the SP 3 case. (author)

ASSEMBLY TRANSFER SYSTEM DESCRIPTION DOCUMENT

International Nuclear Information System (INIS)

Gorpani, B.

2000-01-01

The Assembly Transfer System (ATS) receives, cools, and opens rail and truck transportation casks from the Carrier/Cask Handling System (CCHS). The system unloads transportation casks consisting of bare Spent Nuclear Fuel (SNF) assemblies, single element canisters, and Dual Purpose Canisters (DPCs). For casks containing DPCs, the system opens the DPCs and unloads the SNF. The system stages the assemblies, transfer assemblies to and from fuel-blending inventory pools, loads them into Disposal Containers (DCs), temporarily seals and inerts the DC, decontaminates the DC and transfers it to the Disposal Container Handling System. The system also prepares empty casks and DPCs for off-site shipment. Two identical Assembly Transfer System lines are provided in the Waste Handling Building (WHB). Each line operates independently to handle the waste transfer throughput and to support maintenance operations. Each system line primarily consists of wet and dry handling areas. The wet handling area includes a cask transport system, cask and DPC preparation system, and a wet assembly handling system. The basket transport system forms the transition between the wet and dry handling areas. The dry handling area includes the dry assembly handling system, assembly drying system, DC preparation system, and DC transport system. Both the wet and dry handling areas are controlled by the control and tracking system. The system operating sequence begins with moving transportation casks to the cask preparation area. The cask preparation operations consist of cask cavity gas sampling, cask venting, cask cool-down, outer lid removal, and inner shield plug lifting fixture attachment. Casks containing bare SNF (no DPC) are filled with water and placed in the cask unloading pool. The inner shield plugs are removed underwater. For casks containing a DPC, the cask lid(s) is removed, and the DPC is penetrated, sampled, vented, and cooled. A DPC lifting fixture is attached and the cask is placed

Nuclear data library in design calculation

International Nuclear Information System (INIS)

Hirano, Go; Kosaka, Shinya

2006-01-01

In core design calculation, nuclear data takes part as multi group cross section library during the assembly calculation, which is the first stage of a core design calculation. This report summarizes the multi group cross section libraries used in assembly calculations and also presents the methods adopted for resonance and assembly calculation. (author)

Iterative resonance self-shielding methods using resonance integral table in heterogeneous transport lattice calculations

International Nuclear Information System (INIS)

Hong, Ser Gi; Kim, Kang-Seog

2011-01-01

This paper describes the iteration methods using resonance integral tables to estimate the effective resonance cross sections in heterogeneous transport lattice calculations. Basically, these methods have been devised to reduce an effort to convert resonance integral table into subgroup data to be used in the physical subgroup method. Since these methods do not use subgroup data but only use resonance integral tables directly, these methods do not include an error in converting resonance integral into subgroup data. The effective resonance cross sections are estimated iteratively for each resonance nuclide through the heterogeneous fixed source calculations for the whole problem domain to obtain the background cross sections. These methods have been implemented in the transport lattice code KARMA which uses the method of characteristics (MOC) to solve the transport equation. The computational results show that these iteration methods are quite promising in the practical transport lattice calculations.

Calculations of Neutron Flux Distributions by Means of Integral Transport Methods

Energy Technology Data Exchange (ETDEWEB)

Carlvik, I

1967-05-15

Flux distributions have been calculated mainly in one energy group, for a number of systems representing geometries interesting for reactor calculations. Integral transport methods of two kinds were utilised, collision probabilities (CP) and the discrete method (DIT). The geometries considered comprise the three one-dimensional geometries, planes, sphericals and annular, and further a square cell with a circular fuel rod and a rod cluster cell with a circular outer boundary. For the annular cells both methods (CP and DIT) were used and the results were compared. The purpose of the work is twofold, firstly to demonstrate the versatility and efficacy of integral transport methods and secondly to serve as a guide for anybody who wants to use the methods.

Design of a transport calculation system for logging sondes simulation

International Nuclear Information System (INIS)

Marquez Damian, Jose Ignacio

2005-01-01

Analysis of available resources in earth crust is performed by different techniques, one of them is neutron logging. Design of sondes that are used to make such logging is supported by laboratory experiments as well as by numerical calculations.This work presents several calculation schemes, designed to simplify the task of whom has to planify such experiments or optimize parameters of this kind of sondes.These schemes use transport calculation codes, especially DaRT, TORT and MCNP, and cross section processing modules from SCALE system.Additionally a system for DaRT and TORT data postprocessing using OpenDX is presented.It allows scalar flux spatial distribution analysis, as wells as cross section condensation and reaction rates calculation

Research and design calculation of multipurpose critical assembly using moderated light water and low enriched fuel from 1.6 to 5.0% U-235

International Nuclear Information System (INIS)

Nguyen Kien Cuong; Vo Doan Hai Dang; Luong Ba Vien; Le Vinh Vinh; Huynh Ton Nghiem; Nguyen Minh Tuan; Nguyen Manh Hung; Pham Quang Huy; Tran Quoc Duong; Tran Tri Vien

2015-01-01

Basing on the idea in ??using fuel of nuclear power plants such as PWR (AP-1000) and VVER-1000 with light water as moderation, design calculation of critical assembly was performed to confirm the possibility of using these fuels. Designed critical assembly has simple structure consisting of low enriched fuel from 1.6% to 5% U-235; water has functions as cooling, biological protection and control. Critical assembly is operated at nominal power 100 W with fuel pitch about 2.0 cm. Applications of the critical assembly are quite abundant in basic research, education and training with low investment cost compare with research reactor and easy in operation. So critical assembly can be used for university or training centre for nuclear engineering training. Main objectives of the project are: design calculation in neutronics, thermal hydraulics and safety analysis for critical configuration benchmarks using low enriched fuel; design in mechanical and auxiliary systems for critical assembly; determine technical specifications and estimate construction, installation cost of critical assembly. The process of design, fabrication, installation and construction of critical assembly will be considered with different implementation phases and localization capabilities in installation of critical assembly is highly feasibility. Cost estimation of construction and installation of critical assembly was implemented and showed that investment cost for critical assembly is much lower than research reactor and most of components, systems of critical assembly can be localized with current technique quality of the country. (author)

AUTOMATION OF CALCULATION ALGORITHMS FOR EFFICIENCY ESTIMATION OF TRANSPORT INFRASTRUCTURE DEVELOPMENT

Directory of Open Access Journals (Sweden)

Sergey Kharitonov

2015-06-01

Full Text Available Optimum transport infrastructure usage is an important aspect of the development of the national economy of the Russian Federation. Thus, development of instruments for assessing the efficiency of infrastructure is impossible without constant monitoring of a number of significant indicators. This work is devoted to the selection of indicators and the method of their calculation in relation to the transport subsystem as airport infrastructure. The work also reflects aspects of the evaluation of the possibilities of algorithmic computational mechanisms to improve the tools of public administration transport subsystems.

Preliminary integrated calculation of radionuclide cation and anion transport at Yucca Mountain using a geochemical model

International Nuclear Information System (INIS)

Birdsell, K.H.; Campbell, K.; Eggert, K.G.; Travis, B.J.

1989-01-01

This paper presents preliminary transport calculations for radionuclide movement at Yucca Mountain using preliminary data for mineral distributions, retardation parameter distributions, and hypothetical recharge scenarios. These calculations are not performance assessments, but are used to study the effectiveness of the geochemical barriers at the site at mechanistic level. The preliminary calculations presented have many shortcomings and should be viewed only as a demonstration of the modeling methodology. The simulations were run with TRACRN, a finite-difference porous flow and radionuclide transport code developed for the Yucca Mountain Project. Approximately 30,000 finite-difference nodes are used to represent the unsaturated and saturated zones underlying the repository in three dimensions. Sorption ratios for the radionuclides modeled are assumed to be functions of mineralogic assemblages of the underlying rock. These transport calculations present a representative radionuclide cation, 135 Cs and anion, 99 Tc. The effects on transport of many of the processes thought to be active at Yucca Mountain may be examined using this approach. The model provides a method for examining the integration of flow scenarios, transport, and retardation processes as currently understood for the site. It will also form the basis for estimates of the sensitivity of transport calculations to retardation processes. 11 refs., 17 figs., 1 tab

Mechanical and charge transport properties of alkanethiol self-assembled monolayers on Au (111) surface: The Role of Molecular Tilt

Energy Technology Data Exchange (ETDEWEB)

Mulleregan, Alice; Qi, Yabing; Ratera, Imma; Park, Jeong Y.; Ashby, Paul D.; Quek, Su Ying; Neaton, J. B.; Salmeron, Miquel

2007-11-12

The relationship between charge transport and mechanical properties of alkanethiol self-assembled monolayers (SAM) on Au(111) films has been investigated using an atomic force microscope with a conductive tip. Molecular tilts induced by the pressure applied by the tip cause stepwise increases in film conductivity. A decay constant {beta} = 0.57 {+-} 0.03 {angstrom}{sup -1} was found for the current passing through the film as a function of tip-substrate separation due to this molecular tilt. This is significantly smaller than the value of {approx} 1 {angstrom}{sup -1} found when the separation is changed by changing the length of the alkanethiol molecules. Calculations indicate that for isolated dithiol molecules S-bonded to hollow sites, the junction conductance does not vary significantly as a function of molecular tilt. The impact of S-Au bonding on SAM conductance is discussed.

A simplified treatment of radial enrichment distributions of LWR fuel assemblies in criticality calculations

International Nuclear Information System (INIS)

Hennebach, M.; Schnorrenberg, N.

2008-01-01

Criticality safety assessments are usually performed for fuel assembly models that are as generic as possible to encompass small modifications in geometry that have no impact on criticality. Dealing with different radial enrichment distributions for a fuel assembly type, which is especially important for BWR fuel, poses more of a challenge, since this characteristic is rather obviously influencing the neutronic behaviour of the system. Nevertheless, the large variability of enrichment distributions makes it very desirable and even necessary to treat them in a generalized way, both to keep the criticality safety assessment from becoming too unwieldy and to avoid having to extend it every time a new variation comes up. To be viable, such a generic treatment has to be demonstrably covering, i.e. lead to a higher effective neutron multiplication factor k eff than any of the radial enrichment distributions it represents. Averaging the enrichment evenly over the fuel rods of the assembly is a general and simple approach, and under reactor conditions, it is also a covering assumption: the graded distribution is introduced to achieve a linear power distribution, therefore reducing the enrichment of the better moderated rods at the edge of the assembly. With an even distribution of the average enrichment over all rods, these wellmoderated rods will cause increased fission rates at the assembly edges and a rise in k eff . Since the moderator conditions in a spent nuclear fuel cask differ strongly from a reactor even when considering optimal moderation, the proof that a uniform enrichment distribution is a covering assumption compared with detailed enrichment distributions has to be cask-specific. In this report, a method for making that proof is presented along with results for fuel assemblies from BWR reactors. All results are from three-dimensional Monte Carlo calculations with the SCALE 5.1 code package [1], using a 44-group neutron crosssection library based on ENDF

International report to validate criticality safety calculations for fissile material transport

International Nuclear Information System (INIS)

Whitesides, G.E.

1984-01-01

During the past three years a Working Group established by the Organization for Economic Co-operation and Development's Nuclear Energy Agency (OECD-NEA) in Paris, France, has been studying the validity and applicability of a variety of criticality safety computer programs and their associated nuclear data for the computation of the neutron multiplication factor, k/sub eff/, for various transport packages used in the fuel cycle. The principal objective of this work has been to provide an internationally acceptable basis for the licensing authorities in a country to honor licensing approvals granted by other participating countries. Eleven countries participated in the initial study which consisted of examining criticality safety calculations for packages designed for spent light water reactor fuel transport. This paper presents a summary of this study which has been completed and reported in an OECD-NEA Report No. CSNI-71. The basic goal of this study was to outline a satisfactory validation procedure for this particular application. First, a set of actual critical experiments were chosen which contained the various material and geometric properties present in typical LWR transport containers. Secondly, calculations were made by each of the methods in order to determine how accurately each method reproduced the experimental values. This successful effort in developing a benchmark procedure for validating criticality calculations for spent LWR transport packages along with the successful intercomparison of a number of methods should provide increased confidence by licensing authorities in the use of these methods for this area of application. 4 references, 2 figures

Calculating the Responses of Self-Powered Radiation Detectors.

Science.gov (United States)

Thornton, D. A.

Available from UMI in association with The British Library. The aim of this research is to review and develop the theoretical understanding of the responses of Self -Powered Radiation Detectors (SPDs) in Pressurized Water Reactors (PWRs). Two very different models are considered. A simple analytic model of the responses of SPDs to neutrons and gamma radiation is presented. It is a development of the work of several previous authors and has been incorporated into a computer program (called GENSPD), the predictions of which have been compared with experimental and theoretical results reported in the literature. Generally, the comparisons show reasonable consistency; where there is poor agreement explanations have been sought and presented. Two major limitations of analytic models have been identified; neglect of current generation in insulators and over-simplified electron transport treatments. Both of these are developed in the current work. A second model based on the Explicit Representation of Radiation Sources and Transport (ERRST) is presented and evaluated for several SPDs in a PWR at beginning of life. The model incorporates simulation of the production and subsequent transport of neutrons, gamma rays and electrons, both internal and external to the detector. Neutron fluxes and fuel power ratings have been evaluated with core physics calculations. Neutron interaction rates in assembly and detector materials have been evaluated in lattice calculations employing deterministic transport and diffusion methods. The transport of the reactor gamma radiation has been calculated with Monte Carlo, adjusted diffusion and point-kernel methods. The electron flux associated with the reactor gamma field as well as the internal charge deposition effects of the transport of photons and electrons have been calculated with coupled Monte Carlo calculations of photon and electron transport. The predicted response of a SPD is evaluated as the sum of contributions from individual

About calculation results of heat transfer in the fuel assembly clusters cooled by water with supercritical parameters

International Nuclear Information System (INIS)

Grabezhnaya, V.A.

2008-01-01

Paper reviews the numerical investigation into the heat transfer in the supercritical water cooled fuel assemblies on the basis of the various commercial codes. The turbulence available models specified in the codes describe adequately the experimental data in tubes within the range of flow temperatures away from the pseudocritical point, as well as under high mass velocities. There are k-ε type turbulence models that show qualitatively the local acceleration (slowdown) of the heat transfer in tubes, but they fail to describe the mentioned phenomena quantitatively. To determine the effect of grid spacers on the suppression of the heat transfer local slowdown and on the heat transfer acceleration in fuel assemblies and to ensure more accurate calculation of the fuel element cladding maximum temperature one should perform a number of the experiments making use of the fuel assembly models [ru

Spin-dependent transport properties of oleic acid molecule self-assembled La0.7Sr0.3MnO3 nanoparticles

International Nuclear Information System (INIS)

Xi, L.; Du, J.H.; Ma, J.H.; Wang, Z.; Zuo, Y.L.; Xue, D.S.

2013-01-01

Highlights: ► Spin-dependent transport property of LSMO/oleic acid nanoparticles is investigated. ► Transport properties and MR measured by Cu/nanoparticle assembly/elargol device. ► Non-linear I–V curve indicates a tunneling type transport properties. ► Tunnel barrier height around 1.3 ± 0.15 eV was obtained by fitting I–V curves. ► LFMR of LSMO/oleic acid molecules value reaches −18% with current of 0.1 μA at 10 K. - Abstract: Spin-dependent transport property through molecules is investigated using a monolayer of oleic acid molecule self-assembled half metallic La 0.7 Sr 0.3 MnO 3 (LSMO) nanoparticles, which was synthesized using a coprecipitation method. Fourier transform infrared spectroscopy was used to confirm that one-monolayer oleic acid molecules chemically bond to the LSMO nanoparticles. The transport properties and magnetoresistance (MR) effect of the oleic acid molecule coated LSMO nanoparticles were measured by a direct current four probes method using a Cu/nanoparticle assembly/elargol electrode sandwich device with various temperatures and bias voltages. The non-linear I–V curve indicates a tunneling type transport properties. The tunnel barrier height around 1.3 ± 0.15 eV was obtained by fitting the I–V curve according to the Simmons equation. The magnetoresistance curves can be divided to high-field MR and low-field MR (LFMR) parts. The former is ascribed to the influence of spin disorder or canting within the LSMO nanoparticle surface and the latter one with strong bias dependence is attributed to the spin-dependent tunneling effect through the insulating surface layer of LSMO and oleic acid molecules. The enhanced LFMR effect for oleic acid coated LSMO with respect to the bare LSMO was attributed to the enhanced tunneling transport and weak spin scattering in oleic acid molecule barrier.

Two-dimensional full-core transport theory Benchmarks for the WWER reactors

International Nuclear Information System (INIS)

Petkov, P.T.

2002-01-01

Several two-dimensional full-core real geometry many-group steady-state problems for the WWER-440 and WWER-1000 reactors have been solved by the MARIKO code, based on the method of characteristics. The reference transport theory solutions include assembly-wise and pin-wise power distributions. Homogenized two-group diffusion parameters and discontinuity factors have been calculated by MARIKO for each assembly type both for the whole assembly and for each cell in the smallest sector of symmetry, using the B1 method for calculation of the critical spectrum. Accurate albedo-type boundary conditions have been calculated by MARIKO for the core-reflector and core-absorber boundaries, both for each outer assembly face and for each outer cell face. Comparison with the reference solutions of the two-group nodal diffusion code SPPS-1.6 and the few-group fine-mesh diffusion codes HEX2DA and HEX2DB are presented (Authors)

Peak cladding temperature in a spent fuel storage or transportation cask

International Nuclear Information System (INIS)

Li, J.; Murakami, H.; Liu, Y.; Gomez, P.E.A.; Gudipati, M.; Greiner, M.

2007-01-01

From reactor discharge to eventual disposition, spent nuclear fuel assemblies from a commercial light water reactor are typically exposed to a variety of environments under which the peak cladding temperature (PCT) is an important parameter that can affect the characteristics and behavior of the cladding and, thus, the functions of the spent fuel during storage, transportation, and disposal. Three models have been identified to calculate the peak cladding temperature of spent fuel assemblies in a storage or transportation cask: a coupled effective thermal conductivity and edge conductance model developed by Manteufel and Todreas, an effective thermal conductivity model developed by Bahney and Lotz, and a computational fluid dynamics model. These models were used to estimate the PCT for spent fuel assemblies for light water reactors under helium, nitrogen, and vacuum environments with varying decay heat loads and temperature boundary conditions. The results show that the vacuum environment is more challening than the other gas environments in that the PCT limit is exceeded at a lower boundary temperature for a given decay heat load of the spent fuel assembly. This paper will highlight the PCT calculations, including a comparison of the PCTs obtained by different models.

Demonstration of Coupled Tiamat Single Assembly Calculations

Energy Technology Data Exchange (ETDEWEB)

Novascone, Stephen R. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Hales, Jason D. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Gardner, Russell [Idaho National Lab. (INL), Idaho Falls, ID (United States); Pawlowski, R. P. P. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Pastore, Giovanni [Idaho National Lab. (INL), Idaho Falls, ID (United States); Toth, Alex [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Clarno, Kevin T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Collins, Benjamin S. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Stimpson, Shane G. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Powers, Jeffrey J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2017-06-01

This report corresponds to milestone L3:PHI.PCI.P15.03, which was originally intended to investigate the time discretization approaches with the newly developed fully coupled Tiamat capability, targeting single assembly problems.

Directed Self-Assembly of Nanodispersions

Energy Technology Data Exchange (ETDEWEB)

Furst, Eric M [University of Delaware

2013-11-15

Directed self-assembly promises to be the technologically and economically optimal approach to industrial-scale nanotechnology, and will enable the realization of inexpensive, reproducible and active nanostructured materials with tailored photonic, transport and mechanical properties. These new nanomaterials will play a critical role in meeting the 21st century grand challenges of the US, including energy diversity and sustainability, national security and economic competitiveness. The goal of this work was to develop and fundamentally validate methods of directed selfassembly of nanomaterials and nanodispersion processing. The specific aims were: 1. Nanocolloid self-assembly and interactions in AC electric fields. In an effort to reduce the particle sizes used in AC electric field self-assembly to lengthscales, we propose detailed characterizations of field-driven structures and studies of the fundamental underlying particle interactions. We will utilize microscopy and light scattering to assess order-disorder transitions and self-assembled structures under a variety of field and physicochemical conditions. Optical trapping will be used to measure particle interactions. These experiments will be synergetic with calculations of the particle polarizability, enabling us to both validate interactions and predict the order-disorder transition for nanocolloids. 2. Assembly of anisotropic nanocolloids. Particle shape has profound effects on structure and flow behavior of dispersions, and greatly complicates their processing and self-assembly. The methods developed to study the self-assembled structures and underlying particle interactions for dispersions of isotropic nanocolloids will be extended to systems composed of anisotropic particles. This report reviews several key advances that have been made during this project, including, (1) advances in the measurement of particle polarization mechanisms underlying field-directed self-assembly, and (2) progress in the

Self-assembly of pi-conjugated peptides in aqueous environments leading to energy-transporting bioelectronic nanostructures

Energy Technology Data Exchange (ETDEWEB)

Tavor, John [Johns Hopkins Univ., Baltimore, MD (United States)

2016-12-06

The realization of new supramolecular pi-conjugated organic structures inspired and driven by peptide-based self-assembly will offer a new approach to interface with the biotic environment in a way that will help to meet many DOE-recognized grand challenges. Previously, we developed pi-conjugated peptides that undergo supramolecular self-assembly into one-dimensional (1-D) organic electronic nanomaterials under benign aqueous conditions. The intermolecular interactions among the pi-conjugated organic segments within these nanomaterials lead to defined perturbations of their optoelectronic properties and yield nanoscale conduits that support energy transport within individual nanostructures and throughout bulk macroscopic collections of nanomaterials. Our objectives for future research are to construct and study biomimetic electronic materials for energy-related technology optimized for harsher non-biological environments where peptide-driven self-assembly enhances pi-stacking within nanostructured biomaterials, as detailed in the following specific tasks: (1) synthesis and detailed optoelectronic characterization of new pi-electron units to embed within homogeneous self assembling peptides, (2) molecular and data-driven modeling of the nanomaterial aggregates and their higher-order assemblies, and (3) development of new hierarchical assembly paradigms to organize multiple electronic subunits within the nanomaterials leading to heterogeneous electronic properties (i.e. gradients and localized electric fields). These intertwined research tasks will lead to the continued development and fundamental mechanistic understanding of a powerful bioinspired materials set capable of making connections between nanoscale electronic materials and macroscopic bulk interfaces, be they those of a cell, a protein or a device.

Constant system for by-channel thermal-hydraulic calculation of fuel assembly operational conditions in reactors with natural and mixed convection

International Nuclear Information System (INIS)

Bogatyrev, I.L.; Bogoslovskaya, G.P.; Zhukov, A.V.; Sorokin, A.P.; Titov, P.A.

1992-01-01

System of constants for mass, impulse and energy conservation equations (drag, mixing, heat transfer coefficients, azimuthal unquality of temperature) is reported in region with small Re number for wide range of geometrical assembly parameters. This system can be used in subchannel calculations of assemblies with natural and mixed convection under conditions with loss of flow accident. The formulae are compared with experimental data. 30 refs.; 12 figs.; 1 tab

Performance of a fine-grained parallel model for multi-group nodal-transport calculations in three-dimensional pin-by-pin reactor geometry

International Nuclear Information System (INIS)

Masahiro, Tatsumi; Akio, Yamamoto

2003-01-01

A production code SCOPE2 was developed based on the fine-grained parallel algorithm by the red/black iterative method targeting parallel computing environments such as a PC-cluster. It can perform a depletion calculation in a few hours using a PC-cluster with the model based on a 9-group nodal-SP3 transport method in 3-dimensional pin-by-pin geometry for in-core fuel management of commercial PWRs. The present algorithm guarantees the identical convergence process as that in serial execution, which is very important from the viewpoint of quality management. The fine-mesh geometry is constructed by hierarchical decomposition with introduction of intermediate management layer as a block that is a quarter piece of a fuel assembly in radial direction. A combination of a mesh division scheme forcing even meshes on each edge and a latency-hidden communication algorithm provided simplicity and efficiency to message passing to enhance parallel performance. Inter-processor communication and parallel I/O access were realized using the MPI functions. Parallel performance was measured for depletion calculations by the 9-group nodal-SP3 transport method in 3-dimensional pin-by-pin geometry with 340 x 340 x 26 meshes for full core geometry and 170 x 170 x 26 for quarter core geometry. A PC cluster that consists of 24 Pentium-4 processors connected by the Fast Ethernet was used for the performance measurement. Calculations in full core geometry gave better speedups compared to those in quarter core geometry because of larger granularity. Fine-mesh sweep and feedback calculation parts gave almost perfect scalability since granularity is large enough, while 1-group coarse-mesh diffusion acceleration gave only around 80%. The speedup and parallel efficiency for total computation time were 22.6 and 94%, respectively, for the calculation in full core geometry with 24 processors. (authors)

TRING: a computer program for calculating radionuclide transport in groundwater

International Nuclear Information System (INIS)

Maul, P.R.

1984-12-01

The computer program TRING is described which enables the transport of radionuclides in groundwater to be calculated for use in long term radiological assessments using methods described previously. Examples of the areas of application of the program are activity transport in groundwater associated with accidental spillage or leakage of activity, the shutdown of reactors subject to delayed decommissioning, shallow land burial of intermediate level waste and geologic disposal of high level waste. Some examples of the use of the program are given, together with full details to enable users to run the program. (author)

Robust volume calculations for Constructive Solid Geometry (CSG) components in Monte Carlo transport calculations

Energy Technology Data Exchange (ETDEWEB)

Millman, D. L. [Dept. of Computer Science, Univ. of North Carolina at Chapel Hill (United States); Griesheimer, D. P.; Nease, B. R. [Bechtel Marine Propulsion Corporation, Bertis Atomic Power Laboratory (United States); Snoeyink, J. [Dept. of Computer Science, Univ. of North Carolina at Chapel Hill (United States)

2012-07-01

In this paper we consider a new generalized algorithm for the efficient calculation of component object volumes given their equivalent constructive solid geometry (CSG) definition. The new method relies on domain decomposition to recursively subdivide the original component into smaller pieces with volumes that can be computed analytically or stochastically, if needed. Unlike simpler brute-force approaches, the proposed decomposition scheme is guaranteed to be robust and accurate to within a user-defined tolerance. The new algorithm is also fully general and can handle any valid CSG component definition, without the need for additional input from the user. The new technique has been specifically optimized to calculate volumes of component definitions commonly found in models used for Monte Carlo particle transport simulations for criticality safety and reactor analysis applications. However, the algorithm can be easily extended to any application which uses CSG representations for component objects. The paper provides a complete description of the novel volume calculation algorithm, along with a discussion of the conjectured error bounds on volumes calculated within the method. In addition, numerical results comparing the new algorithm with a standard stochastic volume calculation algorithm are presented for a series of problems spanning a range of representative component sizes and complexities. (authors)

Robust volume calculations for Constructive Solid Geometry (CSG) components in Monte Carlo transport calculations

International Nuclear Information System (INIS)

Millman, D. L.; Griesheimer, D. P.; Nease, B. R.; Snoeyink, J.

2012-01-01

In this paper we consider a new generalized algorithm for the efficient calculation of component object volumes given their equivalent constructive solid geometry (CSG) definition. The new method relies on domain decomposition to recursively subdivide the original component into smaller pieces with volumes that can be computed analytically or stochastically, if needed. Unlike simpler brute-force approaches, the proposed decomposition scheme is guaranteed to be robust and accurate to within a user-defined tolerance. The new algorithm is also fully general and can handle any valid CSG component definition, without the need for additional input from the user. The new technique has been specifically optimized to calculate volumes of component definitions commonly found in models used for Monte Carlo particle transport simulations for criticality safety and reactor analysis applications. However, the algorithm can be easily extended to any application which uses CSG representations for component objects. The paper provides a complete description of the novel volume calculation algorithm, along with a discussion of the conjectured error bounds on volumes calculated within the method. In addition, numerical results comparing the new algorithm with a standard stochastic volume calculation algorithm are presented for a series of problems spanning a range of representative component sizes and complexities. (authors)

Experience from the transportation of irradiated WWER-440 fuel assemblies at the Kozloduy NPP site after a short cooling time

International Nuclear Information System (INIS)

Stoyanova, I.; Kamenov, A.; Byrzev, L.; Christoskov, I.

2003-01-01

Results from the analysis and experimental verification of the radiation and shielding characteristics of non-standard loading patterns of the VSPOT transport cask used for transportation of irradiated fuel assemblies after a short cooling time (120 180 days) on the site of the Kozloduy NPP are presented. An additional safety criterion related to the introduced modifications to the standard procedure of using the transport cask is formulated and discussed (Authors)

Multigroup transport calculations of critical and fuel assemblies with taking into account the scattering anisotropy

International Nuclear Information System (INIS)

Rubin, I.E.; Dneprovskaya, N.M.

2005-01-01

A technique for calculation of reactor lattices by means of the transmission probabilities with taking into account the scattering anisotropy is generalized for the multigroup case. The errors of the calculated multiplication coefficients and energy release distributions do noe exceed practically the errors, of these values, obtained by the Monte Carlo method. The proposed method is most effective when determining the small difference effects [ru

One-group transport theory calculation for three slabs cells

International Nuclear Information System (INIS)

Maia, C.R.M.

1979-01-01

As an idealized model of plate type fuel assemblies for nuclear reactors, three-slab cells are analysed numerically based on the exact solution of the transport equation in the one-group isotropic scattering model. From the equations describing the interface conditions, a set of regular integral equations for the coefficients of the singular eigenfunctions expansions is derived using the half-range orthogonality relations of the eigenfunctions and the recently developed method of regularization. Numerical solutions are obtained by solving this set of equations iteratively. The thermal utilization factor and thermal disadvantage factors as well as flux and current distributions are reported for the first time for various sets of parameters. The accuracy of the P sub(N) approximations is also analysed compared to the exact results. (Author) [pt

Shielding calculations in support of the Spallation Neutron Source (SNS) proton beam transport system

International Nuclear Information System (INIS)

Johnson, Jeffrey O.; Gallmeier, Franz X.; Popova, Irina

2002-01-01

Determining the bulk shielding requirements for accelerator environments is generally an easy task compared to analyzing the radiation transport through the complex shield configurations and penetrations typically associated with the detailed Title II design efforts of a facility. Shielding calculations for penetrations in the SNS accelerator environment are presented based on hybrid Monte Carlo and discrete ordinates particle transport methods. This methodology relies on coupling tools that map boundary surface leakage information from the Monte Carlo calculations to boundary sources for one-, two-, and three-dimensional discrete ordinates calculations. The paper will briefly introduce the coupling tools for coupling MCNPX to the one-, two-, and three-dimensional discrete ordinates codes in the DOORS code suite. The paper will briefly present typical applications of these tools in the design of complex shield configurations and penetrations in the SNS proton beam transport system

Evaluation of PWR and BWR assembly benchmark calculations. Status report of EPRI computational benchmark results, performed in the framework of the Netherlands` PINK programme (Joint project of ECN, IRI, KEMA and GKN)

Energy Technology Data Exchange (ETDEWEB)

Gruppelaar, H. [Netherlands Energy Research Foundation (ECN), Petten (Netherlands); Klippel, H.T. [Netherlands Energy Research Foundation (ECN), Petten (Netherlands); Kloosterman, J.L. [Netherlands Energy Research Foundation (ECN), Petten (Netherlands); Hoogenboom, J.E. [Technische Univ. Delft (Netherlands). Interfacultair Reactor Instituut; Leege, P.F.A. de [Technische Univ. Delft (Netherlands). Interfacultair Reactor Instituut; Verhagen, F.C.M. [Keuring van Electrotechnische Materialen NV, Arnhem (Netherlands); Bruggink, J.C. [Gemeenschappelijke Kernenergiecentrale Nederland N.V., Dodewaard (Netherlands)

1993-11-01

Benchmark results of the Dutch PINK working group on calculational benchmarks on single pin cell and multipin assemblies as defined by EPRI are presented and evaluated. First a short update of methods used by the various institutes involved is given as well as an update of the status with respect to previous performed pin-cell calculations. Problems detected in previous pin-cell calculations are inspected more closely. Detailed discussion of results of multipin assembly calculations is given. The assembly consists of 9 pins in a multicell square lattice in which the central pin is filled differently, i.e. a Gd pin for the BWR assembly and a control rod/guide tube for the PWR assembly. The results for pin cells showed a rather good overall agreement between the four participants although BWR pins with high void fraction turned out to be difficult to calculate. With respect to burnup calculations good overall agreement for the reactivity swing was obtained, provided that a fine time grid is used. (orig.)

Evaluation of PWR and BWR assembly benchmark calculations. Status report of EPRI computational benchmark results, performed in the framework of the Netherlands' PINK programme (Joint project of ECN, IRI, KEMA and GKN)

International Nuclear Information System (INIS)

Gruppelaar, H.; Klippel, H.T.; Kloosterman, J.L.; Hoogenboom, J.E.; Bruggink, J.C.

1993-11-01

Benchmark results of the Dutch PINK working group on calculational benchmarks on single pin cell and multipin assemblies as defined by EPRI are presented and evaluated. First a short update of methods used by the various institutes involved is given as well as an update of the status with respect to previous performed pin-cell calculations. Problems detected in previous pin-cell calculations are inspected more closely. Detailed discussion of results of multipin assembly calculations is given. The assembly consists of 9 pins in a multicell square lattice in which the central pin is filled differently, i.e. a Gd pin for the BWR assembly and a control rod/guide tube for the PWR assembly. The results for pin cells showed a rather good overall agreement between the four participants although BWR pins with high void fraction turned out to be difficult to calculate. With respect to burnup calculations good overall agreement for the reactivity swing was obtained, provided that a fine time grid is used. (orig.)

Carrier Transport Enhancement in Conjugated Polymers through Interfacial Self-Assembly of Solution-State Aggregates

KAUST Repository

Zhao, Kui

2016-07-13

We demonstrate that local and long range orders of poly(3-hexylthiophene) (P3HT) semicrystalline films can be synergistically improved by combining chemical functionalization of the dielectric surface with solution-state disentanglement and pre-aggregation of P3HT in a theta solvent, leading to a very significant enhancement of the field effect carrier mobility. The pre-aggregation and surface functionalization effects combine to enhance the carrier mobility nearly 100-fold as compared with standard film preparation by spin-coating, and nearly 10-fold increase over the benefits of pre-aggregation alone. In situ quartz crystal microbalance with dissipation (QCM-D) experiments reveal enhanced deposition of pre-aggregates on surfaces modified with an alkyl-terminated self-assembled monolayer (SAM) in comparison to un-aggregated polymer chains. Additional investigations reveal the combined pre-aggregation and surface functionalization significantly enhances local order of the conjugated polymer through planarization and extension of the conjugated backbone of the polymer which clearly translate to significant improvements of carrier transport at the semiconductor-dielectric interface in organic thin film transistors. This study points to opportunities in combining complementary routes, such as well-known pre-aggregation with substrate chemical functionalization, to enhance the polymer self-assembly and improve its interfacial order with benefits for transport properties.

Improved method for calculating neoclassical transport coefficients in the banana regime

Energy Technology Data Exchange (ETDEWEB)

Taguchi, M., E-mail: taguchi.masayoshi@nihon-u.ac.jp [College of Industrial Technology, Nihon University, Narashino 275-8576 (Japan)

2014-05-15

The conventional neoclassical moment method in the banana regime is improved by increasing the accuracy of approximation to the linearized Fokker-Planck collision operator. This improved method is formulated for a multiple ion plasma in general tokamak equilibria. The explicit computation in a model magnetic field shows that the neoclassical transport coefficients can be accurately calculated in the full range of aspect ratio by the improved method. The some neoclassical transport coefficients for the intermediate aspect ratio are found to appreciably deviate from those obtained by the conventional moment method. The differences between the transport coefficients with these two methods are up to about 20%.

Thermal-hydraulic calculation and analysis on helium cooled ceramic breeder pebble bed assembly for in-pile irradiation and in-situ tritium extraction

International Nuclear Information System (INIS)

Guo Chunqiu; Xie Jiachun; Liu Xingmin

2013-01-01

In-pile irradiation and in-situ tritium extraction experiment is one of associated domestic research projects in ITER special program. According to the technical requirements of in-pile irradiation experiment of helium cooled ceramic breeder (ceramic) pebble bed assembly in a research reactor, the feasibility of the design for the in-pile irradiation and in-situ tritium extraction experiment of ceramic pebble bed assembly was evaluated. By conducting thermal-hydraulic design calculation with different in-pile irradiation channels, locations and structure parameters for ceramic pebble bed assembly, a reasonable design scheme of ceramic pebble bed assembly satisfying the design requirements for in-pile irradiation was obtained. (authors)

Evaluation of radiation shielding performance in sea transport of radioactive material by using simple calculation method

International Nuclear Information System (INIS)

Odano, N.; Ohnishi, S.; Sawamura, H.; Tanaka, Y.; Nishimura, K.

2004-01-01

A modified code system based on the point kernel method was developed to use in evaluation of shielding performance for maritime transport of radioactive material. For evaluation of shielding performance accurately in the case of accident, it is required to preciously model the structure of transport casks and shipping vessel, and source term. To achieve accurate modelling of the geometry and source term condition, we aimed to develop the code system by using equivalent information regarding structure and source term used in the Monte Carlo calculation code, MCNP. Therefore, adding an option to use point kernel method to the existing Monte Carlo code, MCNP4C, the code system was developed. To verify the developed code system, dose rate distribution in an exclusive shipping vessel to transport the low level radioactive wastes were calculated by the developed code and the calculated results were compared with measurements and Monte Carlo calculations. It was confirmed that the developed simple calculation method can obtain calculation results very quickly with enough accuracy comparing with the Monte Carlo calculation code MCNP4C

Considerations of beta and electron transport in internal dose calculations. Progress report

Energy Technology Data Exchange (ETDEWEB)

Bolch, W.E.

1994-11-01

The goal of this particular task is to consider, for the first time, the explicit transport of beta particles and photon-generated electrons in the series of six phantoms developed by Cristy and Eckerman (1987) at the Oak Ridge National Laboratory. In their report, ORNL/TM-8381, specific absorbed fractions of energy are reported for phantoms representing the newborn (3.4 kg), the one-year-old (9.8 kg), the five-year-old (19 kg), the ten-year-old (32 kg), the fifteen-year-old/adult female (55-58 kg), and the adult male (70 kg). Radiation transport calculations were performed with the Monte Carlo code ALGAMP which allows photon transport only. In subsequent calculations of radionuclide S values as is done in the MIRDOSE2 computer program, electron absorbed fractions are thus considered to be either unity or zero depending upon whether the source region does or does not equal the target region, respectively.

Beam transport calculations for BARC-TIFR 14UD pelletron

International Nuclear Information System (INIS)

Prasad, K.G.

1993-01-01

The 14UD pelletron tandem accelerator installed at Tata Institute of Fundamental Research (TIFR) as a joint BARC-TIFR project, is supplied by National Electrostatic Corporation (NEC), U.S.A. To optimise the parameters of various elements along the beam path, it is essential to work out the beam optics of the entire system. There are various computer codes in use for such calculations. All these codes, except the detailed ray tracing programs, use matrix formulation. Thus each ion optical element is characterised in terms of a transport matrix, whose elements are assumed to be independent of particle trajectory. We have performed only the first order calculations, meaning thereby that no aberrations are included. Further, all calculations are carried out assuming ideal conditions like axial beam injection, perfectly aligned beam line elements, etc. The main code that has been employed in our calculations is based on the one at the Australian National University, Canberra, suitably modified for use with CYBER 170/730 computer at TIFR. However, codes at NEC and Stony Brook were also used for the checking the results. The results of calculations are given and discussed. (author). 2 figs

Channel selective tunnelling through a nanographene assembly

International Nuclear Information System (INIS)

Wong, H S; Durkan, C; Feng, X; Müllen, K; Chandrasekhar, N

2012-01-01

We report selective tunnelling through a nanographene intermolecular tunnel junction achieved via scanning tunnelling microscope tip functionalization with hexa-peri-hexabenzocoronene (HBC) molecules. This leads to an offset in the alignment between the energy levels of the tip and the molecular assembly, resulting in the imaging of a variety of distinct charge density patterns in the HBC assembly, not attainable using a bare metallic tip. Different tunnelling channels can be selected by the application of an electric field in the tunnelling junction, which changes the condition of the HBC on the tip. Density functional theory-based calculations relate the imaged HBC patterns to the calculated molecular orbitals at certain energy levels. These patterns bear a close resemblance to the π-orbital states of the HBC molecule calculated at the relevant energy levels, mainly below the Fermi energy of HBC. This correlation demonstrates the ability of an HBC functionalized tip as regards accessing an energy range that is restricted to the usual operating bias range around the Fermi energy with a normal metallic tip at room temperature. Apart from relating to molecular orbitals, some patterns could also be described in association with the Clar aromatic sextet formula. Our observations may help pave the way towards the possibility of controlling charge transport between organic interfaces. (paper)

Theoretical analysis of time-dependent neutron spectra in bulk assemblies

International Nuclear Information System (INIS)

Akimoto, Tadashi; Ogawa, Yuichi; Togawa, Orihiko.

1988-01-01

Time-dependent neutron spectra in an iron assembly and in a graphite assembly are obtained with the one-dimensional S N calculation, in order an attempt to investigate the availability of these spectra to the benchmark test by the LINAC-TOF method for evaluation of nuclear data and numerical methods. The group constants are taken from the JAERI FAST SET Version 1, 2 and the ABBN SET. It was demonstrated by a sensitivity test that the time-dependent neutron spectra are sensitive to changes in the inelastic scattering cross section data in the iron assembly and to changes in the elastic scattering cross section data in the graphite assembly. Moreover, it is shown that the time-dependent spectra in the graphite assembly are sensitive to the group structure. Because some information about the neutron transport phenomena which has not been obtained in the stationary spectra is observed in the time-dependent spectra, the availability of the benchmark test based on the time-dependent spectra is indicated from the theoretical analysis. (author)

Program for calculating multi-component high-intense ion beam transport

International Nuclear Information System (INIS)

Kazarinov, N.Yu.; Prejzendorf, V.A.

1985-01-01

The CANAL program for calculating transport of high-intense beams containing ions with different charges in a channel consisting of dipole magnets and quadrupole lenses is described. The equations determined by the method of distribution function momenta and describing coordinate variations of the local mass centres and r.m.s. transverse sizes of beams with different charges form the basis of the calculation. The program is adapted for the CDC-6500 and SM-4 computers. The program functioning is organized in the interactive mode permitting to vary the parameters of any channel element and quickly choose the optimum version in the course of calculation. The calculation time for the CDC-6500 computer for the 30-40 m channel at the integration step of 1 cm is about 1 min. The program is used for calculating the channel for the uranium ion beam injection from the collective accelerator into the heavy-ion synchrotron

Burnup calculations for KIPT accelerator driven subcritical facility using Monte Carlo computer codes-MCB and MCNPX

International Nuclear Information System (INIS)

Gohar, Y.; Zhong, Z.; Talamo, A.

2009-01-01

Argonne National Laboratory (ANL) of USA and Kharkov Institute of Physics and Technology (KIPT) of Ukraine have been collaborating on the conceptual design development of an electron accelerator driven subcritical (ADS) facility, using the KIPT electron accelerator. The neutron source of the subcritical assembly is generated from the interaction of 100 KW electron beam with a natural uranium target. The electron beam has a uniform spatial distribution and electron energy in the range of 100 to 200 MeV. The main functions of the subcritical assembly are the production of medical isotopes and the support of the Ukraine nuclear power industry. Neutron physics experiments and material structure analyses are planned using this facility. With the 100 KW electron beam power, the total thermal power of the facility is ∼375 kW including the fission power of ∼260 kW. The burnup of the fissile materials and the buildup of fission products reduce continuously the reactivity during the operation, which reduces the neutron flux level and consequently the facility performance. To preserve the neutron flux level during the operation, fuel assemblies should be added after long operating periods to compensate for the lost reactivity. This process requires accurate prediction of the fuel burnup, the decay behavior of the fission produces, and the introduced reactivity from adding fresh fuel assemblies. The recent developments of the Monte Carlo computer codes, the high speed capability of the computer processors, and the parallel computation techniques made it possible to perform three-dimensional detailed burnup simulations. A full detailed three-dimensional geometrical model is used for the burnup simulations with continuous energy nuclear data libraries for the transport calculations and 63-multigroup or one group cross sections libraries for the depletion calculations. Monte Carlo Computer code MCNPX and MCB are utilized for this study. MCNPX transports the electrons and the

Fission blanket benchmark experiment on spherical assembly of uranium and PE with PE reflector

Energy Technology Data Exchange (ETDEWEB)

Zhu, Tonghua; Lu, Xinxin; Wang, Mei; Han, Zijie, E-mail: neutron_integral@aliyun.com; Jiang, Li; Wen, Zhongwei; Liu, Rong

2016-04-15

Highlights: • The fission rate distribution on two depleted uranium assemblies was measured with plate fission chambers. • We do calculations using MCNP code and ENDF/B-V.0 library. • The overestimation of calculations to the measured fission rates was found. • The observed discrepancy are discussed. - Abstract: New concept of fusion-fission hybrid for energy generation has been proposed. To validate the nuclear performance of fission blanket of hybrid, as part of series of validation experiment, two types of fission blanket assemblies were setup in this work and measurements were made of the reaction rate distribution for uranium fission in the spherical assembly of depleted uranium and polyethylene by Plate Fission Chamber (PFC). There are two PFCs in experiment, one is depleted uranium chamber and the other is enriched uranium chamber. The Monte-Carlo transport code MCNP5 and continuous energy cross sections library ENDF/BV.0 were used for the analysis of fission rate distribution in the two types of assemblies. The calculated results were compared with the experimental ones. The overestimation of fission rate for depleted uranium and enriched uranium were found in the inner boundary of the two assemblies. However, the C/E ratio tends to decrease for the distance from the core slightly and the results for enriched uranium are better than that for depleted uranium.

Electrical transport properties in Co nanocluster-assembled granular film

Science.gov (United States)

Zhang, Qin-Fu; Wang, Lai-Sen; Wang, Xiong-Zhi; Zheng, Hong-Fei; Liu, Xiang; Xie, Jia; Qiu, Yu-Long; Chen, Yuanzhi; Peng, Dong-Liang

2017-03-01

A Co nanocluster-assembled granular film with three-dimensional cross-connection paralleled conductive paths was fabricated by using the plasma-gas-condensation method in a vacuum environment. The temperature-dependent longitudinal resistivity and anomalous Hall effect of this new type granular film were systematically studied. The longitudinal resistivity of the Co nanocluster-assembled granular film first decreased and then increased with increasing measuring temperature, revealing a minimum value at certain temperature, T min . In a low temperature region ( T governed the electrical transport process, and the temperature coefficient of resistance (TCR) showed an insulator-type behavior. The thermal fluctuation-induced tunneling conduction progressively increased with increasing temperature, which led to a decrease in the longitudinal resistivity. In a high temperature region, the TCR showed a metallic-type behavior, which was primarily attributed to the temperature-dependent scattering. Different from the longitudinal resistivity behavior, the saturated anomalous Hall resistivity increased monotonically with increasing measuring temperature. The value of the anomalous Hall coefficient ( R S ) reached 2.3 × 10-9 (Ω cm)/G at 300 K, which was about three orders of magnitude larger than previously reported in blocky single-crystal Co [E. N. Kondorskii, Sov. Phys. JETP 38, 977 (1974)]. Interestingly, the scaling relation ( ρx y A ∝ ρx x γ ) between saturated anomalous Hall resistivity ( ρx y A ) and longitudinal resistivity ( ρ x x ) was divided into two regions by T min . However, after excluding the contribution of tunneling, the scaling relation followed the same rule. The corresponding physical mechanism was also proposed to explain these phenomena.

Risks of transport of radioactive materials on the road; some exploring calculations performed with the INTERTRAN-model

International Nuclear Information System (INIS)

1987-04-01

Under the auspices of the IAEA a computercode, named INTERTRAN, has been developed in order to calculate the risks of the transport of radioactive materials. This code has to be tested nearer. For the Dutch situation a number of calculations has been performed of more or less realistic cases in which four transport streams have been investigated. Two transport routes are chosen. The risks thus obtained are compared quantitatively with the risks of LPG-transports. 4 refs.; 9 figs

Ab Initio Calculations of Transport Properties of Vanadium Oxides

Science.gov (United States)

Lamsal, Chiranjivi; Ravindra, N. M.

2018-04-01

The temperature-dependent transport properties of vanadium oxides have been studied near the Fermi energy using the Kohn-Sham band structure approach combined with Boltzmann transport equations. V2O5 exhibits significant thermoelectric properties, which can be attributed to its layered structure and stability. Highly anisotropic electrical conduction in V2O5 is clearly manifested in the calculations. Due to specific details of the band structure and anisotropic electron-phonon interactions, maxima and crossovers are also seen in the temperature-dependent Seebeck coefficient of V2O5. During the phase transition of VO2, the Seebeck coefficient changes by 18.9 µV/K, which is close to (within 10% of) the observed discontinuity of 17.3 µV/K.

Burnup code for fuel assembly by Monte Carlo code. MKENO-BURN

International Nuclear Information System (INIS)

Naito, Yoshitaka; Suyama, Kenya; Masukawa, Fumihiro; Matsumoto, Kiyoshi; Kurosawa, Masayoshi; Kaneko, Toshiyuki.

1996-12-01

The evaluation of neutron spectrum is so important for burnup calculation of the heterogeneous geometry like recent BWR fuel assembly. MKENO-BURN is a multi dimensional burnup code that based on the three dimensional monte carlo neutron transport code 'MULTI-KENO' and the routine for the burnup calculation of the one dimensional burnup code 'UNITBURN'. MKENO-BURN analyzes the burnup problem of arbitrary regions after evaluating the neutron spectrum and making one group cross section in three dimensional geometry with MULTI-KENO. It enables us to do three dimensional burnup calculation. This report consists of general description of MKENO-BURN and the input data. (author)

Burnup code for fuel assembly by Monte Carlo code. MKENO-BURN

Energy Technology Data Exchange (ETDEWEB)

Naito, Yoshitaka; Suyama, Kenya; Masukawa, Fumihiro; Matsumoto, Kiyoshi; Kurosawa, Masayoshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Kaneko, Toshiyuki

1996-12-01

The evaluation of neutron spectrum is so important for burnup calculation of the heterogeneous geometry like recent BWR fuel assembly. MKENO-BURN is a multi dimensional burnup code that based on the three dimensional monte carlo neutron transport code `MULTI-KENO` and the routine for the burnup calculation of the one dimensional burnup code `UNITBURN`. MKENO-BURN analyzes the burnup problem of arbitrary regions after evaluating the neutron spectrum and making one group cross section in three dimensional geometry with MULTI-KENO. It enables us to do three dimensional burnup calculation. This report consists of general description of MKENO-BURN and the input data. (author)

Transport calculation of neutron flux distribution in reflector of PW reactor

International Nuclear Information System (INIS)

Remec, I.

1982-01-01

Two-dimensional transport calculation of the neutron flux and spectrum in the equatorial plain of PW reactor, using computer program DOT 3, is presented. Results show significant differences between neutron fields in which test samples and reactor vessel are exposed. (author)

Point defects and atomic transport in crystals

International Nuclear Information System (INIS)

Lidiard, A.B.

1981-02-01

There are two principle aspects to the theory of atomic transport in crystals as caused by the action of point defects, namely (1) the calculation of relevant properties of the point defects (energies and other thermodynamic characteristics of the different possible defects, activation energies and other mobility parameters) and (2) the statistical mechanics of assemblies of defects, both equilibrium and non-equilibrium assemblies. In the five lectures given here both these aspects are touched on. The first two lectures are concerned with the calculation of relevant point defect properties, particularly in ionic crystals. The first lecture is more general, the second is concerned particularly with some recent calculations of the free volumes of formation of defects in various ionic solids; these solve a rather long-standing problem in this area. The remaining three lectures are concerned with the kinetic theory of defects mainly in relaxation, drift and diffusion situations

Parallel MCNP Monte Carlo transport calculations with MPI

International Nuclear Information System (INIS)

Wagner, J.C.; Haghighat, A.

1996-01-01

The steady increase in computational performance has made Monte Carlo calculations for large/complex systems possible. However, in order to make these calculations practical, order of magnitude increases in performance are necessary. The Monte Carlo method is inherently parallel (particles are simulated independently) and thus has the potential for near-linear speedup with respect to the number of processors. Further, the ever-increasing accessibility of parallel computers, such as workstation clusters, facilitates the practical use of parallel Monte Carlo. Recognizing the nature of the Monte Carlo method and the trends in available computing, the code developers at Los Alamos National Laboratory implemented the message-passing general-purpose Monte Carlo radiation transport code MCNP (version 4A). The PVM package was chosen by the MCNP code developers because it supports a variety of communication networks, several UNIX platforms, and heterogeneous computer systems. This PVM version of MCNP has been shown to produce speedups that approach the number of processors and thus, is a very useful tool for transport analysis. Due to software incompatibilities on the local IBM SP2, PVM has not been available, and thus it is not possible to take advantage of this useful tool. Hence, it became necessary to implement an alternative message-passing library package into MCNP. Because the message-passing interface (MPI) is supported on the local system, takes advantage of the high-speed communication switches in the SP2, and is considered to be the emerging standard, it was selected

Calculated and experimental research of WWER-1000 assembly vibration and fretting damage

International Nuclear Information System (INIS)

Drozdov, Y.; Afanasyev, A.; Makarov, V.; Tutnov, A.; Tutnov, A.; Alekseev, E.

2008-01-01

The report covers the methods and results of the latest analytical and experimental studies of fretting corrosion and natural vibrations of a WWER-1000 reactor fuel assemblies (FA). The process of fretting-corrosion was investigated using a multi-specimen facility that simulated fragments of fuel rod-to-spacer grid and lower support grid mating units. A computational model was developed for vibrations in the mechanical system of a fuel rod fragment and a spacer grid fragment. A calculational and experimental modal analysis of a FA was performed. Natural frequencies, modes and decrements of FA vibrations were determined and a satisfactory coincidence of analytical and experimental results was obtained. The assessment of fretting-corrosion process dynamics was made and its dependences on operational factors were obtained. (authors)

Simplified calculation method for radiation dose under normal condition of transport

International Nuclear Information System (INIS)

Watabe, N.; Ozaki, S.; Sato, K.; Sugahara, A.

1993-01-01

In order to estimate radiation dose during transportation of radioactive materials, the following computer codes are available: RADTRAN, INTERTRAN, J-TRAN. Because these codes consist of functions for estimating doses not only under normal conditions but also in the case of accidents, when nuclei may leak and spread into the environment by air diffusion, the user needs to have special knowledge and experience. In this presentation, we describe how, with a view to preparing a method by which a person in charge of transportation can calculate doses in normal conditions, the main parameters upon which the value of doses depends were extracted and the dose for a unit of transportation was estimated. (J.P.N.)

Experience with the loading and transport of fuel assembly transport casks, including CASTOR casks, and the radiation exposure of personnel

International Nuclear Information System (INIS)

Bentele, W.; Kinzelmann, T.

1999-01-01

In 1997 and 1998, six spent fuel assembly transports started from the nuclear power plant Gemeinschaftskernkraftwerk Neckar (GKN), using CASTOR-V19 casks. Professor Kuni of Marburg University challenged the statement made by the German Federal Office for Radiation Protection (Bundesamt fuer Strahlenschutz (BfS)) based on accepted scientific knowledge, according to which so-called CASTOR transports present no risk, either to the population or to the escorting police units. This paper shows that the collective dose during the loading of the CASTOR casks amounted to 4.5 mSv (gamma and neutrons) per cask at the most, and that the maximum individual dose amounted to 0.26 mSv. In addition to these doses, the collective dose during handling and transport must be considered: this amounted to 0.35 mSv (gamma and neutrons). The dose to the police escort was -2 (limit for surface contamination), presented degrees of contamination >4 Bq cm -2 upon reaching the Valognes/Cogema terminal. However, transport casks coming from French plants also revealed degrees of contamination >4 Bq cm -2 , as well as 'hot spots'. No such contamination was found on NTL 11 casks transported from the GKN to Sellafield. Neither was any increased contamination found upon the arrival of CASTOR-V19 casks transported from GKN to Gorleben or Ahaus. The partially sensationalist media reports were inversely proportional to the actual radiological relevance of the matter. The German Commission on Radiation Protection (SSK) confirmed that the radiological effect of such contaminated spent fuel transports is negligible. (author)

Finite volume thermal-hydraulics and neutronics coupled calculations - 15300

International Nuclear Information System (INIS)

Araujo Silva, V.; Campagnole dos Santos, A.A.; Mesquit, A.Z.; Bernal, A.; Miro, R.; Verdu, G.; Pereira, C.

2015-01-01

The computational power available nowadays allows the coupling of neutronics and thermal-hydraulics codes for reactor studies. The present methodology foresees at least one constraint to the separated codes in order to perform coupled calculations: both codes must use the same geometry, however, meshes can be different for each code as long as the internal surfaces stays the same. Using the finite volume technique, a 3D diffusion nodal code was implemented to deal with neutron transport. This code can handle non-structured meshes which allows for complicated geometries calculations and therefore more flexibility. A computational fluid dynamics (CFD) code was used in order to obtain the same level of details for the thermal hydraulics calculations. The chosen code is OpenFOAM, an open-source CFD tool. Changes in OpenFOAM allow simple coupled calculations of a PWR fuel rod with neutron transport code. OpenFOAM sends coolant density information and fuel temperature to the neutron transport code that sends back power information. A mapping function is used to average values when one node in one side corresponds to many nodes in the other side. Data is exchanged between codes by library calls. As the results of a fuel rod calculations progress, more complicated and processing demanding geometries will be simulated, aiming to the simulation of a real scale PWR fuel assembly

Thermal transport across metal silicide-silicon interfaces: First-principles calculations and Green's function transport simulations

Science.gov (United States)

Sadasivam, Sridhar; Ye, Ning; Feser, Joseph P.; Charles, James; Miao, Kai; Kubis, Tillmann; Fisher, Timothy S.

2017-02-01

Heat transfer across metal-semiconductor interfaces involves multiple fundamental transport mechanisms such as elastic and inelastic phonon scattering, and electron-phonon coupling within the metal and across the interface. The relative contributions of these different transport mechanisms to the interface conductance remains unclear in the current literature. In this work, we use a combination of first-principles calculations under the density functional theory framework and heat transport simulations using the atomistic Green's function (AGF) method to quantitatively predict the contribution of the different scattering mechanisms to the thermal interface conductance of epitaxial CoSi2-Si interfaces. An important development in the present work is the direct computation of interfacial bonding from density functional perturbation theory (DFPT) and hence the avoidance of commonly used "mixing rules" to obtain the cross-interface force constants from bulk material force constants. Another important algorithmic development is the integration of the recursive Green's function (RGF) method with Büttiker probe scattering that enables computationally efficient simulations of inelastic phonon scattering and its contribution to the thermal interface conductance. First-principles calculations of electron-phonon coupling reveal that cross-interface energy transfer between metal electrons and atomic vibrations in the semiconductor is mediated by delocalized acoustic phonon modes that extend on both sides of the interface, and phonon modes that are localized inside the semiconductor region of the interface exhibit negligible coupling with electrons in the metal. We also provide a direct comparison between simulation predictions and experimental measurements of thermal interface conductance of epitaxial CoSi2-Si interfaces using the time-domain thermoreflectance technique. Importantly, the experimental results, performed across a wide temperature range, only agree well with

Some benchmark calculations for VVER-1000 assemblies by WIMS-7B code

International Nuclear Information System (INIS)

Sultanov, N.V.

2001-01-01

Our aim in this report is to compare of calculation results, obtained with the use of different libraries, which are in the variant of the WIMS7B code. We had the three libraries: the 1986 library is based on the UKNDL files, the two 1996 libraries are based on the JEF-2.2 files, the one having the 69 group approximation, the other having the 172 group approximation. We wanted also to have some acquaintance with the new option of WIMS-7B - CACTUS. The variant of WIMS-7B was placed at our disposal by the code authors for a temporal use for 9 months. It was natural to make at comparisons with analogous values of TVS-M, MCU, Apollo-2, Casmo-4, Conkemo, MCNP, HELIOS codes, where the other different libraries were used. In accordance with our aims the calculations of unprofiled and profiled assemblies of the VVER-1000 reactor have been carried out by the option CACTUS. This option provides calculations by the characteristics method. The calculation results have been compared with the K ∞ values obtained by other codes in work. The conclusion from this analysis is such: the methodical parts of errors of these codes have nearly the same values. Spacing for K eff values can be explained of the library microsections differences mainly. Nevertheless, the more detailed analysis of the results obtained is required. In conclusion the calculation of a depletion of VVER-1000 cell has been carried out. The comparison of the dependency of the multiply factor from the depletion obtained by WIMS-7B with different libraries and by the TVS-M, MCU, HELIOS and WIMS-ABBN codes in work has been performed. (orig.)

Lagrangian Transport Calculations Using UARS Data. Part I: Passive Tracers

Science.gov (United States)

Manney, G. L.; Lahoz, W. A.; Harwood, R. S.; Zurek, R. W.; Kumer, J. B.; Mergenthaler, J. L.; Roche, A. E.; O'Neill, A; Swinbank, R.; Waters, J. W.

1994-01-01

The transport of passive tracers observed by UARS has been simulated using computed trajectories of thousands of air parcels initialized on a three-dimensional stratospheric grid. These trajectories are calculated in isentropic coordinates using horizontal winds provided by the United Kingdom Meteorological Office data assimilation system and vertical (cross-isentropic) velocities computed using a fast radiation code.

Parallel processing of two-dimensional Sn transport calculations

International Nuclear Information System (INIS)

Uematsu, M.

1997-01-01

A parallel processing method for the two-dimensional S n transport code DOT3.5 has been developed to achieve a drastic reduction in computation time. In the proposed method, parallelization is achieved with angular domain decomposition and/or space domain decomposition. The calculational speed of parallel processing by angular domain decomposition is largely influenced by frequent communications between processing elements. To assess parallelization efficiency, sample problems with up to 32 x 32 spatial meshes were solved with a Sun workstation using the PVM message-passing library. As a result, parallel calculation using 16 processing elements, for example, was found to be nine times as fast as that with one processing element. As for parallel processing by geometry segmentation, the influence of processing element communications on computation time is small; however, discontinuity at the segment boundary degrades convergence speed. To accelerate the convergence, an alternate sweep of angular flux in conjunction with space domain decomposition and a two-step rescaling method consisting of segmentwise rescaling and ordinary pointwise rescaling have been developed. By applying the developed method, the number of iterations needed to obtain a converged flux solution was reduced by a factor of 2. As a result, parallel calculation using 16 processing elements was found to be 5.98 times as fast as the original DOT3.5 calculation

FAMREC, PWR Lateral Mechanical Fuel Rod Assembly Response

International Nuclear Information System (INIS)

Guenzler, R.C.

1995-01-01

1 - Description of program or function: The Fuel Assembly Mechanical Response Code (FAMREC) calculates the lateral mechanical response of a row of fuel assemblies while allowing for two types of nonlinearities. The first type is a geometric nonlinearity in the form of gaps between individual assemblies and between peripheral assemblies and a boundary wall. Impacting is monitored across the gaps. The second nonlinearity is the permanent deformation of the fuel assembly spacer grid to compressive loading. 2 - Method of solution: The response is calculated in the modal plane. The coupled differential equations are solved in closed form using Laplace transformations. The discrete displacements and velocities are then calculated and the gaps in the system monitored at each axial elevation for impacting. These impact forces are then applied statistically at a given time-step, and equilibrium is found using a Gaussian elimination technique. Three impact force calculation methods are available: 1- a linear impact force and crushing load audit calculation, 2- a more detailed linear impact force and crushing load calculation, and 3- a non-linear grid calculation which allows for plastic deformation of the fuel assembly spacer grids. 3 - Restrictions on the complexity of the problem: Maxima of: 3601 time-steps and forces; 80 modes; 30 applied forces; 15 fuel assemblies; and 5 impact grids per assembly

Criticality calculations of a generic fuel container for fuel assemblies PWR, by means of the code MCNP

International Nuclear Information System (INIS)

Vargas E, S.; Esquivel E, J.; Ramirez S, J. R.

2013-10-01

The purpose of the concept of burned consideration (Burn-up credit) is determining the capacity of the calculation codes, as well as of the nuclear data associates to predict the isotopic composition and the corresponding neutrons effective multiplication factor in a generic container of spent fuel during some time of relevant storage. The present work has as objective determining this capacity of the calculation code MCNP in the prediction of the neutrons effective multiplication factor for a fuel assemblies arrangement type PWR inside a container of generic storage. The calculations are divided in two parts, the first, in the decay calculations with specified nuclide concentrations by the reference for a pressure water reactor (PWR) with enriched fuel to 4.5% and a discharge burned of 50 GW d/Mtu. The second, in criticality calculations with isotopic compositions dependent of the time for actinides and important fission products, taking 30 time steps, for two actinide groups and fission products. (Author)

Development of packagings for 'MONJU' blanket fuel assemblies

International Nuclear Information System (INIS)

Shibata, Kan; Ouchi, Yuichiro; Matsuzaki, Masaaki; Okuda, Yoshihisa

1995-01-01

Blanket assemblies for prototype Fast Breeder Reactor 'MONJU' are made at commercial fuel fabrication plants capable of handling deplete Uranium in Japan. For the purpose of transport the assemblies are inserted into a packaging that is set horizontally at the fabrication plants because of compatibility with equipment installed at the plants. On the other hand, the assemblies must be taken out from the packaging set vertically at 'MONJU' due to compatibility. For this reason development of a new packaging, which makes it possible to take assemblies in and out both horizontally and vertically, is needed to carry out transport of assemblies for reload. The development and fabrication of the packagings, taking about two years, were completed in March 1995. The packagings were used in transport of assemblies in June 1995 for the first change. This report introduces the outline of the packaging and confirmation tests done in the process of development. (author)

GAPER-1D, 1-D Multigroup 1. Order Perturbation Transport Theory for Reactivity Coefficient

International Nuclear Information System (INIS)

Koch, P.K.

1976-01-01

1 - Description of problem or function: Reactivity coefficients are computed using first-order transport perturbation theory for one- dimensional multi-region reactor assemblies. The number of spatial mesh-points and energy groups is arbitrary. An elementary synthesis scheme is employed for treatment of two- and three-dimensional problems. The contributions to the change in inverse multiplication factor, delta(1/k), from perturbations in the individual capture, net fission, total scattering, (n,2n), inelastic scattering, and leakage cross sections are computed. A multi-dimensional prompt neutron lifetime calculation is also available. 2 - Method of solution: Broad group cross sections for the core and perturbing or sample materials are required as input. Scalar neutron fluxes and currents, as computed by SN transport calculations, are then utilized to solve the first-order transport perturbation theory equations. A synthesis scheme is used, along with independent SN calculations in two or three dimensions, to treat a multi- dimensional assembly. Spherical harmonics expansions of the angular fluxes and scattering source terms are used with leakage and anisotropic scattering treated in a P1 approximation. The angular integrations in the perturbation theory equations are performed analytically. Various reactivity coefficients and material worths are then easily computed at specified positions in the assembly. 3 - Restrictions on the complexity of the problem: The formulation of the synthesis scheme used for two- and three-dimensional problems assumes that the fluxes and currents were computed by the DTF4 code (NESC Abstract 209). Therefore, fluxes and currents from two- or three-dimensional transport or diffusion theory codes cannot be used

Calculation of Single Cell and Fuel Assembly IRIS Benchmarks Using WIMSD5B and GNOMER Codes

International Nuclear Information System (INIS)

Pevec, D.; Grgic, D.; Jecmenica, R.

2002-01-01

fuel assemblies, and assembly-wise 2-D burnup distribution at the end of cycle. Comparison of results obtained and reference results for Benchmark 44 showed adequacy of our calculational tools for IRIS core design calculations. (author)

Theoretical background and user's manual for the computer code on groundwater flow and radionuclide transport calculation in porous rock

International Nuclear Information System (INIS)

Shirakawa, Toshihiko; Hatanaka, Koichiro

2001-11-01

In order to document a basic manual about input data, output data, execution of computer code on groundwater flow and radionuclide transport calculation in heterogeneous porous rock, we investigated the theoretical background about geostatistical computer codes and the user's manual for the computer code on groundwater flow and radionuclide transport which calculates water flow in three dimension, the path of moving radionuclide, and one dimensional radionuclide migration. In this report, based on above investigation we describe the geostatistical background about simulating heterogeneous permeability field. And we describe construction of files, input and output data, a example of calculating of the programs which simulates heterogeneous permeability field, and calculates groundwater flow and radionuclide transport. Therefore, we can document a manual by investigating the theoretical background about geostatistical computer codes and the user's manual for the computer code on groundwater flow and radionuclide transport calculation. And we can model heterogeneous porous rock and analyze groundwater flow and radionuclide transport by utilizing the information from this report. (author)

Calculation of Transport Coefficients in Dense Plasma Mixtures

Science.gov (United States)

Haxhimali, T.; Cabot, W. H.; Caspersen, K. J.; Greenough, J.; Miller, P. L.; Rudd, R. E.; Schwegler, E. R.

2011-10-01

We use classical molecular dynamics (MD) to estimate species diffusivity and viscosity in mixed dense plasmas. The Yukawa potential is used to describe the screened Coulomb interaction between the ions. This potential has been used widely, providing the basis for models of dense stellar materials, inertial confined plasmas, and colloidal particles in electrolytes. We calculate transport coefficients in equilibrium simulations using the Green- Kubo relation over a range of thermodynamic conditions including the viscosity and the self - diffusivity for each component of the mixture. The interdiffusivity (or mutual diffusivity) can then be related to the self-diffusivities by using a generalization of the Darken equation. We have also employed non-equilibrium MD to estimate interdiffusivity during the broadening of the interface between two regions each with a high concentration of either species. Here we present results for an asymmetric mixture between Ar and H. These can easily be extended to other plasma mixtures. A main motivation for this study is to develop accurate transport models that can be incorporated into the hydrodynamic codes to study hydrodynamic instabilities. We use classical molecular dynamics (MD) to estimate species diffusivity and viscosity in mixed dense plasmas. The Yukawa potential is used to describe the screened Coulomb interaction between the ions. This potential has been used widely, providing the basis for models of dense stellar materials, inertial confined plasmas, and colloidal particles in electrolytes. We calculate transport coefficients in equilibrium simulations using the Green- Kubo relation over a range of thermodynamic conditions including the viscosity and the self - diffusivity for each component of the mixture. The interdiffusivity (or mutual diffusivity) can then be related to the self-diffusivities by using a generalization of the Darken equation. We have also employed non-equilibrium MD to estimate interdiffusivity during

Neutron and gamma ray transport calculations in shielding system

Energy Technology Data Exchange (ETDEWEB)

Masukawa, Fumihiro; Sakamoto, Hiroki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

1998-03-01

In the shields for radiation in nuclear facilities, the penetrating holes of various kinds and irregular shapes are made for the reasons of operation, control and others. These penetrating holes and gaps are filled with air or the substances with relatively small shielding performance, and radiation flows out through them, which is called streaming. As the calculation techniques for the shielding design or analysis related to the streaming problem, there are the calculations by simplified evaluation, transport calculation and Monte Carlo method. In this report, the example of calculation by Monte Carlo method which is represented by MCNP code is discussed. A number of variance reduction techniques which seem effective for the analysis of streaming problem were tried. As to the investigation of the applicability of MCNP code to streaming analysis, the object of analysis which are the concrete walls without hole and with horizontal hole, oblique hole and bent oblique hole, the analysis procedure, the composition of concrete, and the conversion coefficient of dose equivalent, and the results of analysis are reported. As for variance reduction technique, cell importance was adopted. (K.I.)

Neutron and photon transport calculations in fusion system. 2

Energy Technology Data Exchange (ETDEWEB)

Sato, Satoshi [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment

1998-03-01

On the application of MCNP to the neutron and {gamma}-ray transport calculations for fusion reactor system, the wide range design calculation has been carried out in the engineering design activities for the international thermonuclear fusion experimental reactor (ITER) being developed jointly by Japan, USA, EU and Russia. As the objects of shielding calculation for fusion reactors, there are the assessment of dose equivalent rate for living body shielding and the assessment of the nuclear response for the soundness of in-core structures. In the case that the detailed analysis of complicated three-dimensional shapes is required, the assessment using MCNP has been carried out. Also when the nuclear response of peripheral equipment due to the gap streaming between blanket modules is evaluated with good accuracy, the calculation with MCNP has been carried out. The analyses of the shieldings for blanket modules and NBI port are explained, and the examples of the results of analyses are shown. In the blanket modules, there are penetrating holes and continuous gap. In the case of the NBI port, shielding plug cannot be installed. These facts necessitate the MCNP analysis with high accuracy. (K.I.)

Exploring the Tilt-Angle Dependence of electron tunneling across Molecular junction of Self-Assembled Alkanethiols

DEFF Research Database (Denmark)

Frederiksen, Thomas; Munuera, C.; Ocal, C.

2009-01-01

Electronic transport mechanisms in molecular junctions are investigated by a combination of first-principles calculations and current−voltage measurements of several well-characterized structures. We study self-assembled layers of alkanethiols grown on Au(111) and form tunnel junctions...... for the longer molecular chains. Our calculations confirm the observed trends and explain them as a result of two mechanisms, namely, a previously proposed intermolecular tunneling enhancement as well as a hitherto overlooked tilt-dependent molecular gate effect....

Stability analysis of the Backward Euler time discretization for the pin-resolved transport transient reactor calculation

International Nuclear Information System (INIS)

Zhu, Ang; Xu, Yunlin; Downar, Thomas

2016-01-01

Three-dimensional, full core transport modeling with pin-resolved detail for reactor dynamic simulation is important for some multi-physics reactor applications. However, it can be computationally intensive due to the difficulty in maintaining accuracy while minimizing the number of time steps. A recently proposed Transient Multi-Level (TML) methodology overcomes this difficulty by use multi-level transient solvers to capture the physical phenomenal in different time domains and thus maximize the numerical accuracy and computational efficiency. One major problem with the TML method is the negative flux/precursor number density generated using large time steps for the MOC solver, which is due to the Backward Euler discretization scheme. In this paper, the stability issue of Backward Euler discretization is first investigated using the Point Kinetics Equations (PKEs), and the predicted maximum allowed time step for SPERT test 60 case is shown to be less than 10 ms. To overcome this difficulty, linear and exponential transformations are investigated using the PKEs. The linear transformation is shown to increase the maximum time step by a factor of 2, and the exponential transformation is shown to increase the maximum time step by a factor of 5, as well as provide unconditionally stability above a specified threshold. The two sets of transformations are then applied to TML scheme in the MPACT code, and the numerical results presented show good agreement for standard, linear transformed, and exponential transformed maximum time step between the PKEs model and the MPACT whole core transport solution for three different cases, including a pin cell case, a 3D SPERT assembly case and a row of assemblies (“striped assembly case”) from the SPERT model. Finally, the successful whole transient execution of the stripe assembly case shows the ability of the exponential transformation method to use 10 ms and 20 ms time steps, which all failed using the standard method.

A coupled diffusion-transport computational method and its application for the determination of space dependent angular flux distributions at a cold neutron source

International Nuclear Information System (INIS)

Turgut, M.H.

1985-01-01

A fast calculation program ''BRIDGE'' was developed for the calculation of a Cold Neutron Source (CNS) at a radial beam tube of the FRG-I reactor, which couples a total assembly diffusion calculation to a transport calculation for a certain subregion. For the coupling flux and current boundary values at the common surfaces are taken from the diffusion calculation and are used as driving conditions in the transport calculation. 'Equivalence Theorie' is used for the transport feedback effect on the diffusion calculation to improve the consistency of the boundary values. The optimization of a CNS for maximizing the subthermal flux in the wavelength range 4 - 6 A is discussed. (orig.) [de

D0 Silicon Upgrade: West End Assembly Hall Platform Design Calculations

International Nuclear Information System (INIS)

Rucinski, Russ

1996-01-01

This engineering note documents design calculations done for the bayonet feed can platform installed at the far west end of the assembly hall. The platform is mounted off of a cast concrete wall directly south of where the shielding block wall is stacked. A summary of the loading, reaction forces and stresses is shown on the page 3. As can be seen, the calculated stresses are very small, maximum value = 2540 psi. The material used is structural steel tubing, ASTM A500 Gr. B, with a minimum yield strength of 46 ksi and minimum ultimate tensile strength of 58 ksi. The reaction forces for the upper two members will be carried together by a 1/2-inch mounting plate. The mounting plate is attached to the wall by four 1/2-inch Hilti wedge anchors. The allowables for each wedge anchor are 2400 lbs. tensile, 1960 lbs. shear. The major reaction load for the top members is a combined 3627 lbs. tensile load which can easily be handled by the four bolt pattern. Some small moment reactions not listed on the summary page add negligible (400 lbs.) force couples to the axial loading. The bottom members are also attached to a mounting plate that is bolted to the wall. See page 15 for Hilti wedge anchor data.

A sub-structure method for multidimensional integral transport calculations

International Nuclear Information System (INIS)

Kavenoky, A.; Stankovski, Z.

1983-03-01

A new method has been developed for fine structure burn-up calculations of very heterogeneous large size media. It is a generalization of the well-known surface-source method, allowing coupling actual two-dimensional heterogeneous assemblies, called sub-structures. The method has been applied to a rectangular medium, divided into sub-structures, containing rectangular and/or cylindrical fuel, moderator and structure elements. The sub-structures are divided into homogeneous zones. A zone-wise flux expansion is used to formulate a direct collision probability problem within it (linear or flat flux expansion in the rectangular zones, flat flux in the others). The coupling of the sub-structures is performed by making extra assumptions on the currents entering and leaving the interfaces. The accuracies and computing times achieved are illustrated by numerical results on two benchmark problems

Dissecting the sequential assembly and localization of intraflagellar transport particle complex B in Chlamydomonas.

Directory of Open Access Journals (Sweden)

Elizabeth A Richey

Full Text Available Intraflagellar transport (IFT, the key mechanism for ciliogenesis, involves large protein particles moving bi-directionally along the entire ciliary length. IFT particles contain two large protein complexes, A and B, which are constructed with proteins in a core and several peripheral proteins. Prior studies have shown that in Chlamydomonas reinhardtii, IFT46, IFT52, and IFT88 directly interact with each other and are in a subcomplex of the IFT B core. However, ift46, bld1, and ift88 mutants differ in phenotype as ift46 mutants are able to form short flagella, while the other two lack flagella completely. In this study, we investigated the functional differences of these individual IFT proteins contributing to complex B assembly, stability, and basal body localization. We found that complex B is completely disrupted in bld1 mutant, indicating an essential role of IFT52 for complex B core assembly. Ift46 mutant cells are capable of assembling a relatively intact complex B, but such complex is highly unstable and prone to degradation. In contrast, in ift88 mutant cells the complex B core still assembles and remains stable, but the peripheral proteins no longer attach to the B core. Moreover, in ift88 mutant cells, while complex A and the anterograde IFT motor FLA10 are localized normally to the transition fibers, complex B proteins instead are accumulated at the proximal ends of the basal bodies. In addition, in bld2 mutant, the IFT complex B proteins still localize to the proximal ends of defective centrioles which completely lack transition fibers. Taken together, these results revealed a step-wise assembly process for complex B, and showed that the complex first localizes to the proximal end of the centrioles and then translocates onto the transition fibers via an IFT88-dependent mechanism.

Calculation of Quasi-Particle Energies of Aromatic Self-Assembled Monolayers on Au(111).

Science.gov (United States)

Li, Yan; Lu, Deyu; Galli, Giulia

2009-04-14

We present many-body perturbation theory calculations of the electronic properties of phenylene diisocyanide self-assembled monolayers (SAMs) on a gold surface. Using structural models obtained within density functional theory (DFT), we have investigated how the SAM molecular energies are modified by self-energy corrections and how they are affected by the presence of the surface. We have employed a combination of GW (G = Green's function; W = screened Coulomb interaction) calculations of the SAM quasi-particle energies and a semiclassical image potential model to account for surface polarization effects. We find that it is essential to include both quasi-particle corrections and surface screening in order to provide a reasonable estimate of the energy level alignment at a SAM-metal interface. In particular, our results show that within the GW approximation the energy distance between phenylene diisocyanide SAM energy levels and the gold surface Fermi level is much larger than that found within DFT, e.g., more than double in the case of low packing densities of the SAM.

Calibration of spent fuel measurement assembly

International Nuclear Information System (INIS)

Koleska, Michal; Viererbl, Ladislav; Marek, Milan

2014-01-01

The LVR-15 research reactor (Czech Republic) had been converted from the highly enriched IRT-2M to the low enriched IRT-4M fuel. For the possibility of the independent pre-transport evaluation of IRT-2M burnup, a spectrometric system was developed. This spectrometric system consists of the fuel holder, the collimator and the portable Canberra Big MAC HPGe (High Purity Germanium) detector. In order to have well reproducible and reliable experimental data for modeling of the measurement system, calibration with the 110m Ag isotope with known activity was performed. This isotope was chosen for having energies similar to isotopes measured in fuel assemblies. The 110m Ag isotope was prepared by irradiating of the silver foil in LVR-15 research reactor; its activity was evaluated in the LVR-15's spectrometric laboratory. From the measured data, an efficiency curve of the spectrometric system has been determined. The experimental data were compared to the calculation results with the MCNPX model of the spectrometric system. - Highlights: • Calibration of research reactor spent fuel measurement assembly. • On-site prepared 110m Ag isotope used for the measurement. • Calculated self-shielding factor for the IRT-2M fuel. • Applicable to other research reactor fuel geometries

Reactivity measurements on an experimental assembly of 4.31 wt % 235U enriched UO2 fuel rods arranged in a shipping cask geometry

International Nuclear Information System (INIS)

Bierman, S.R.

1989-10-01

A research program was initiated for the US Department of Energy (DOE) Sandia National Laboratory Transportation Systems Development Department in 1982 to provide benchmark type experimental criticality data in support of the design and safe operations of nuclear fuel transportation systems. The overall objective of the program is to identify and provide the experimental data needed to form a consistent, firm, and complete data base for verifying calculational models used in the criticality analyses of nuclear transport and related systems. A report, PNL-6205, issued in June 1988 (Bierman 1988) covered measurement results obtained from a series of experimental assemblies (TIC-1, 2, 3 and 4) involving neutron flux traps. The results obtained on a fifth experimental assembly (TIC-5), modeled after a calculational problem of the Organization for Economic Cooperation and Development (OECD) Nuclear Energy Agency (NEA) Committee on the Safety of Nuclear Installations (CSNI) Working Group, are covered in this report. 10 refs., 10 figs., 7 tabs

Uncertainties in calculations of nuclear design code system for the high temperature engineering test reactor (HTTR)

International Nuclear Information System (INIS)

Shindo, R.; Yamashita, K.; Murata, I.

1991-01-01

The nuclear design code system for the HTTR consists of one dimensional cell burnup computer code, developed in JAERI and the TWOTRAN-2 transport code. In order to satisfy related design criteria, uncertainty of the calculation was investigated by comparing the calculated and experimental results. The experiments were performed with a graphite moderated critical assembly. It was confirmed that discrepancies between calculations and experiments were small enough to be allowed in the nuclear design of HTTR. 8 refs, 6 figs

An investigation of fission models for high-energy radiation transport calculations

International Nuclear Information System (INIS)

Armstrong, T.W.; Cloth, P.; Filges, D.; Neef, R.D.

1983-07-01

An investigation of high-energy fission models for use in the HETC code has been made. The validation work has been directed checking the accuracy of the high-energy radiation transport computer code HETC to investigate the appropriate model for routine calculations, particularly for spallation neutron source applications. Model calculations are given in terms of neutron production, fission fragment energy release, and residual nuclei production for high-energy protons incident on thin uranium targets. The effect of the fission models on neutron production from thick uranium targets is also shown. (orig.)

THE CALCULATION OF THE ENERGY RECOVERY ELECTRIFIED URBAN TRANSPORT DURING THE INSTALLATION DRIVE FOR TRACTION SUBSTATION

Directory of Open Access Journals (Sweden)

A. A. Sulim

2014-01-01

Full Text Available At present a great attention is paid to increasing of energy efficiency at operated electrified urban transport. Perspective direction for increasing energy efficiency at that type of transport is the application of regenerative braking. For additional increasing of energy efficiency there were suggested the use of capacitive drive on tires of traction substation. One of the main task is the analysis of energy recovery application  with drive and without it.These analysis demonstrated that the calculation algorithms don’t allow in the full volume to carry out calculations of amount and cost of energy recovery without drive and with it. That is why we see the current interest to this topic. The purpose of work is to create methods of algorithms calculation for definite amount and cost of consumed, redundant and recovery energy of electrified urban transport due to definite regime of motion on wayside. There is algorithm developed, which allow to calculate amount and cost of consumed, redundant and recovery energy of electrified urban transport on wayside during the installation capacitive drive at traction substation. On the basis of developed algorithm for the definite regime of wagon motion of subway there were fulfilled the example of energy recovery amount and its cost calculation, among them with limited energy intensity drive, when there are 4 trains on wayside simultaneously.

Investigation of the structural anisotropy in a self-assembling glycinate layer on Cu(100) by scanning tunneling microscopy and density functional theory calculations

Energy Technology Data Exchange (ETDEWEB)

Kuzmin, Mikhail [Surface Science Laboratory, Optoelectronics Research Centre, Tampere University of Technology, P.O. Box 692, FI-33101 Tampere (Finland); Ioffe Physical Technical Institute, Russian Academy of Sciences, 26 Polytekhnicheskaya, St Petersburg 194021 (Russian Federation); Lahtonen, Kimmo; Vuori, Leena [Surface Science Laboratory, Optoelectronics Research Centre, Tampere University of Technology, P.O. Box 692, FI-33101 Tampere (Finland); Sánchez-de-Armas, Rocío [Materials Theory Division, Department of Physics and Astronomy, Uppsala University, P.O. Box 516, S75120 Uppsala (Sweden); Hirsimäki, Mika, E-mail: mikahirsi@gmail.com [Surface Science Laboratory, Optoelectronics Research Centre, Tampere University of Technology, P.O. Box 692, FI-33101 Tampere (Finland); Valden, Mika [Surface Science Laboratory, Optoelectronics Research Centre, Tampere University of Technology, P.O. Box 692, FI-33101 Tampere (Finland)

2017-07-01

Highlights: • Deprotonation reaction of glycine and self-assembly of glycinate is observed on Cu. • Bias-dependent scanning tunneling microscopy indicates two glycinate geometries. • Density functional theory calculations confirm the two non-identical configurations. • Non-identical adsorption explains the anisotropy in adlayer’s electronic structure. - Abstract: Self-assembling organic molecule-metal interfaces exhibiting free-electron like (FEL) states offers an attractive bottom-up approach to fabricating materials for molecular electronics. Accomplishing this, however, requires detailed understanding of the fundamental driving mechanisms behind the self-assembly process. For instance, it is still unresolved as to why the adsorption of glycine ([NH{sub 2}(CH{sub 2})COOH]) on isotropic Cu(100) single crystal surface leads, via deprotonation and self-assembly, to a glycinate ([NH{sub 2}(CH{sub 2})COO–]) layer that exhibits anisotropic FEL behavior. Here, we report on bias-dependent scanning tunneling microscopy (STM) experiments and density functional theory (DFT) calculations for glycine adsorption on Cu(100) single crystal surface. We find that after physical vapor deposition (PVD) of glycine on Cu(100), glycinate self-assembles into an overlayer exhibiting c(2 × 4) and p(2 × 4) symmetries with non-identical adsorption sites. Our findings underscore the intricacy of electrical conductivity in nanomolecular organic overlayers and the critical role the structural anisotropy at molecule-metal interface plays in the fabrication of materials for molecular electronics.

KALIMER-600-clad Core Fuel Assembly Calculation using MATRA-LMR (V2.0) Code

International Nuclear Information System (INIS)

Kim, Young Gyun; Kim, Young Il

2006-12-01

Since the sodium boiling point is very high, maximum cladding and pin temperatures are used for design limit condition in sodium cooled liquid metal reactor. It is necessary to predict accurately the temperature distribution in the core and in the subassemblies to increase the sodium coolant efficiency. Based on the MATRA code, which is developed for PWR analysis, MATRA-LMR has been developed for SFR. The major modifications are: the sodium properties table is implemented as subprogram in the code, Heat transfer coefficients are changed for SFR, te pressure drop correlations are changed for more accurate calculations, which are Novendstern, Chiu-Rohsenow-Todreas, and Cheng-Todreas correlations. This This report describes briefly code structure and equations of MATRA-LMR (Version 2.0), explains input data preparation and shows some calculation results for the KALIMER-600-clad core fuel assembly for which has been performed the conceptual design of the core in the year 2006

Neutron transport calculation for Activation Evaluation for Decommissioning of PET cyclotron Facility

Science.gov (United States)

Nobuhara, Fumiyoshi; Kuroyanagi, Makoto; Masumoto, Kazuyoshi; Nakamura, Hajime; Toyoda, Akihiro; Takahashi, Katsuhiko

2017-09-01

In order to evaluate the state of activation in a cyclotron facility used for the radioisotope production of PET diagnostics, we measured the neutron flux by using gold foils and TLDs. Then, the spatial distribution of neutrons and induced activity inside the cyclotron vault were simulated with the Monte Calro calculation code for neutron transport and DCHAIN-SP for activation calculation. The calculated results are in good agreement with measured values within factor 3. Therefore, the adaption of the advanced evaluation procedure for activation level is proved to be important for the planning of decommissioning of these facilities.

Development of a database system for the calculation of indicators of environmental pressure caused by transport

DEFF Research Database (Denmark)

Giannouli, Myrsini; Samaras, Zissis; Keller, Mario

2006-01-01

The scope of this paper is to summarise a methodology developed for TRENDS (TRansport and ENvironment Database System-TRENDS). The main objective of TRENDS was the calculation of environmental pressure indicators caused by transport. The environmental pressures considered are associated with air...... emissions from the four main transport modes, i.e. road, rail, ships and air. In order to determine these indicators a system for calculating a range of environmental pressures due to transport was developed within a PC-based MS Access environment. Emphasis is given oil the latest features incorporated...... the production of collective results for all transport modes as well as a comparative assessment of air emissions produced by the various modes. Traffic activity and emission data obtained according to a basic (reference) scenario are displayed for the time period 1970-2020. In addition, a detailed assessment...

Central Reactivity Measurements on Assemblies 1 and 3 of the Fast Reactor FR0

International Nuclear Information System (INIS)

Londen, S.O.

1966-01-01

The reactivity effects of small samples of various materials have been measured, by the period method at the core centre of Assemblies 1 and 3 of the fast zero power reactor FR0. For some materials the reactivity change as a function of sample size has also been determined experimentally. The core of Assembly 1 consisted only of uranium enriched to 20 % whereas the core of Assembly 3 was diluted with 30 % graphite. The results have been compared with calculated values obtained with a second-order transport-theoretical perturbation model and using differently shielded cross sections depending upon sample size. Qualitative agreement has generally been found, although discrepancies still exist. The spectrum perturbation caused by the experimental arrangement has been analyzed and found to be rather important

Central Reactivity Measurements on Assemblies 1 and 3 of the Fast Reactor FR0

Energy Technology Data Exchange (ETDEWEB)

Londen, S O

1966-01-15

The reactivity effects of small samples of various materials have been measured, by the period method at the core centre of Assemblies 1 and 3 of the fast zero power reactor FR0. For some materials the reactivity change as a function of sample size has also been determined experimentally. The core of Assembly 1 consisted only of uranium enriched to 20 % whereas the core of Assembly 3 was diluted with 30 % graphite. The results have been compared with calculated values obtained with a second-order transport-theoretical perturbation model and using differently shielded cross sections depending upon sample size. Qualitative agreement has generally been found, although discrepancies still exist. The spectrum perturbation caused by the experimental arrangement has been analyzed and found to be rather important.

Continuous Energy, Multi-Dimensional Transport Calculations for Problem Dependent Resonance Self-Shielding

International Nuclear Information System (INIS)

Downar, T.

2009-01-01

The overall objective of the work here has been to eliminate the approximations used in current resonance treatments by developing continuous energy multi-dimensional transport calculations for problem dependent self-shielding calculations. The work here builds on the existing resonance treatment capabilities in the ORNL SCALE code system. The overall objective of the work here has been to eliminate the approximations used in current resonance treatments by developing continuous energy multidimensional transport calculations for problem dependent self-shielding calculations. The work here builds on the existing resonance treatment capabilities in the ORNL SCALE code system. Specifically, the methods here utilize the existing continuous energy SCALE5 module, CENTRM, and the multi-dimensional discrete ordinates solver, NEWT to develop a new code, CENTRM( ) NEWT. The work here addresses specific theoretical limitations in existing CENTRM resonance treatment, as well as investigates advanced numerical and parallel computing algorithms for CENTRM and NEWT in order to reduce the computational burden. The result of the work here will be a new computer code capable of performing problem dependent self-shielding analysis for both existing and proposed GENIV fuel designs. The objective of the work was to have an immediate impact on the safety analysis of existing reactors through improvements in the calculation of fuel temperature effects, as well as on the analysis of more sophisticated GENIV/NGNP systems through improvements in the depletion/transmutation of actinides for Advanced Fuel Cycle Initiatives.

The new deterministic 3-D radiation transport code Multitrans: C5G7 MOX fuel assembly benchmark

International Nuclear Information System (INIS)

Kotiluoto, P.

2003-01-01

The novel deterministic three-dimensional radiation transport code MultiTrans is based on combination of the advanced tree multigrid technique and the simplified P3 (SP3) radiation transport approximation. In the tree multigrid technique, an automatic mesh refinement is performed on material surfaces. The tree multigrid is generated directly from stereo-lithography (STL) files exported by computer-aided design (CAD) systems, thus allowing an easy interface for construction and upgrading of the geometry. The deterministic MultiTrans code allows fast solution of complicated three-dimensional transport problems in detail, offering a new tool for nuclear applications in reactor physics. In order to determine the feasibility of a new code, computational benchmarks need to be carried out. In this work, MultiTrans code is tested for a seven-group three-dimensional MOX fuel assembly transport benchmark without spatial homogenization (NEA C5G7 MOX). (author)

Secondary gamma-ray data for shielding calculation

International Nuclear Information System (INIS)

Miyasaka, Sunichi

1979-01-01

In deep penetration transport calculations, the integral design parameters is determined mainly by secondary particles which are produced by interactions of the primary radiation with materials. The shield thickness and the biological dose rate at a given point of a bulk shield are determined from the contribution from secondary gamma rays. The heat generation and the radiation damage in the structural and shield materials depend strongly on the secondary gamma rays. In this paper, the status of the secondary gamma ray data and its further problems are described from the viewpoint of shield design. The secondary gamma-ray data in ENDF/B-IV and POPOP4 are also discussed based on the test calculations made for several shield assemblies. (author)

Numerical verification/validation of the theory of coupled reactors for deuterium critical assembly, using MCNP5 and Serpent codes

International Nuclear Information System (INIS)

Hussein, M.S; Lewis, B.J.; Bonin, H.W.

2013-01-01

The theory of multipoint coupled reactors developed by multi-group transport is verified by using the probabilistic transport code MCNP5 and the continuous-energy Monte Carlo reactor physics burnup calculation Serpent code. The verification was performed by calculating the multiplication factors (or criticality factors) and coupling coefficients for a two-region test reactor known as the Deuterium Critical Assembly, DCA. The multiplication factors k eff calculated numerically and independently from simulations of the DCA by MCNP5 and Serpent codes are compared with the multiplication factors k eff calculated based on the coupled reactor theory. Excellent agreement was obtained between the multiplication factors k eff calculated with the Serpent code, with MCNP5, and from the coupled reactor theory. This analysis demonstrates that the Serpent code is valid for the multipoint coupled reactor calculations. (author)

Numerical verification/validation of the theory of coupled reactors for deuterium critical assembly, using MCNP5 and Serpent codes

Energy Technology Data Exchange (ETDEWEB)

Hussein, M.S, E-mail: mohamed.hussein@rmc.ca [Royal Military College of Canada, Dept. of Chemistry and Chemical Engineering, Kingston, Ontario (Canada); Lewis, B.J., E-mail: Brent.Lewis@uoit.ca [Univ. of Ontario Inst. of Technology, Faculty of Energy Systems and Nuclear Science, Oshawa, Ontario (Canada); Bonin, H.W., E-mail: bonin-h@rmc.ca [Royal Military College of Canada, Dept. of Chemistry and Chemical Engineering, Kingston, Ontario (Canada)

2013-07-01

The theory of multipoint coupled reactors developed by multi-group transport is verified by using the probabilistic transport code MCNP5 and the continuous-energy Monte Carlo reactor physics burnup calculation Serpent code. The verification was performed by calculating the multiplication factors (or criticality factors) and coupling coefficients for a two-region test reactor known as the Deuterium Critical Assembly, DCA. The multiplication factors k{sub eff} calculated numerically and independently from simulations of the DCA by MCNP5 and Serpent codes are compared with the multiplication factors k{sub eff} calculated based on the coupled reactor theory. Excellent agreement was obtained between the multiplication factors k{sub eff} calculated with the Serpent code, with MCNP5, and from the coupled reactor theory. This analysis demonstrates that the Serpent code is valid for the multipoint coupled reactor calculations. (author)

Repair for scattering expansion truncation errors in transport calculations

International Nuclear Information System (INIS)

Emmett, M.B.; Childs, R.L.; Rhoades, W.A.

1980-01-01

Legendre expansion of angular scattering distributions is usually limited to P 3 in practical transport calculations. This truncation often results in non-trivial errors, especially alternating negative and positive lateral scattering peaks. The effect is especially prominent in forward-peaked situations such as the within-group component of the Compton Scattering of gammas. Increasing the expansion to P 7 often makes the peaks larger and narrower. Ward demonstrated an accurate repair, but his method requires special cross section sets and codes. The DOT IV code provides fully-compatible, but heuristic, repair of the erroneous scattering. An analytical Klein-Nishina estimator, newly available in the MORSE code, allows a test of this method. In the MORSE calculation, particle scattering histories are calculated in the usual way, with scoring by an estimator routine at each collision site. Results for both the conventional P 3 estimator and the analytical estimator were obtained. In the DOT calculation, the source moments are expanded into the directional representation at each iteration. Optionally a sorting procedure removes all negatives, and removes enough small positive values to restore particle conservation. The effect of this is to replace the alternating positive and negative values with positive values of plausible magnitude. The accuracy of those values is examined herein

Rapid method of calculating the fluence and spectrum of neutrons from a critical assembly and the resulting dose

International Nuclear Information System (INIS)

Bessis, J.

1977-01-01

The proposed calculation method is unsophisticated but rapid. The first part (computer code CRITIC), which is based on the Fermi age equation, evaluates the number of neutrons per fission emitted from a moderated critical assembly and their energy spectrum. The second part (computer code NARCISSE), which uses the current albedo for concrete, evaluates the product of neutron reflection on the walls and calculates the fluence resulting at any point in the room and its energy distribution by 21 groups. The results obtained are shown to compare satisfactorily with those obtained through the use of a Monte Carlo program

Theoretical prediction of the electronic transport properties of the Al-Cu alloys based on the first-principle calculation and Boltzmann transport equation

Science.gov (United States)

Choi, Garam; Lee, Won Bo

Metal alloys, especially Al-based, are commonly-used materials for various industrial applications. In this paper, the Al-Cu alloys with varying the Al-Cu ratio were investigated based on the first-principle calculation using density functional theory. And the electronic transport properties of the Al-Cu alloys were carried out using Boltzmann transport theory. From the results, the transport properties decrease with Cu-containing ratio at the temperature from moderate to high, but with non-linearity. It is inferred by various scattering effects from the calculation results with relaxation time approximation. For the Al-Cu alloy system, where it is hard to find the reliable experimental data for various alloys, it supports understanding and expectation for the thermal electrical properties from the theoretical prediction. Theoretical and computational soft matters laboratory.

Probabilistic Risk Assessment of Cask Drop Accident during On-site Spent Nuclear Fuel Transportation

Energy Technology Data Exchange (ETDEWEB)

Ham, Jae Hyun; Christian, Robby; Momani, Belal Al; Kang, Hyun Gook [KAERI, Daejeon (Korea, Republic of)

2016-05-15

There are two ways to transfer the SNF from a site to other site, one is land transportation and the other is maritime transportation. Maritime transportation might be used because this way uses more safe route which is far from populated area. The whole transportation process can be divided in two parts: transferring the SNF between SNP and wharf in-Nuclear Power Plant (NPP) site by truck, and transferring the SNF from the wharf to the other wharf by ship. In this research, on-site SNF transportation between SNP and wharf was considered. Two kinds of single accident can occur during this type of SNF transportation, impact and fire, caused by internal events and external events. In this research, PRA of cask drop accident during onsite SNF transportation was done, risk to a person (mSv/person) from a case with specific conditions was calculated. In every 11 FEM simulation drop cases, FDR is 1 even the fuel assemblies are located inside of the cask. It is a quite larger value for all cases than the results with similar drop condition from the reports which covers the PRA on cask storage system. Because different from previous reports, subsequent impact was considered. Like in figure 8, accelerations which are used to calculate the FDR has extremely higher values in subsequent impact than the first impact for all SNF assemblies.

Method for calculating anisotropic neutron transport using scattering kernel without polynomial expansion

International Nuclear Information System (INIS)

Takahashi, Akito; Yamamoto, Junji; Ebisuya, Mituo; Sumita, Kenji

1979-01-01

A new method for calculating the anisotropic neutron transport is proposed for the angular spectral analysis of D-T fusion reactor neutronics. The method is based on the transport equation with new type of anisotropic scattering kernels formulated by a single function I sub(i) (μ', μ) instead of polynomial expansion, for instance, Legendre polynomials. In the calculation of angular flux spectra by using scattering kernels with the Legendre polynomial expansion, we often observe the oscillation with negative flux. But in principle this oscillation disappears by this new method. In this work, we discussed anisotropic scattering kernels of the elastic scattering and the inelastic scatterings which excite discrete energy levels. The other scatterings were included in isotropic scattering kernels. An approximation method, with use of the first collision source written by the I sub(i) (μ', μ) function, was introduced to attenuate the ''oscillations'' when we are obliged to use the scattering kernels with the Legendre polynomial expansion. Calculated results with this approximation showed remarkable improvement for the analysis of the angular flux spectra in a slab system of lithium metal with the D-T neutron source. (author)

Transportation of 33 irradiated MTR fuel assemblies from FRM/Garching to Savannah River Site, USA, using a GNS transport cask and using a new loading device

International Nuclear Information System (INIS)

Dreesen, K.; Goetze, H.G.; Holst, L.; Gerstenberg, H.; Schreckenbach, K.

2000-01-01

According to the Department of Energy program of the return spent fuel from the foreign research reactors operators, 33 irradiated MTR box shaped fuel assemblies from the Technical University Munich were shipped to SRS/USA. The fuel assemblies were irradiated for typically 800 full days and, after a sufficient cooling time, loaded into a GNS 16 cask. The GNS 16 cask is a new transport cask for box shaped MTR fuel assemblies and TRIGA fuel assemblies and was used for the first time at the FRM Garching. The capacity of the cask is 33 box shaped MTR fuel assemblies. During the loading of the fuel assemblies, a newly developed loading device was used. The main components of the loading device are the transfer flask, the shielded loading lock, adapter plate and a mobile water tank. The loading device works mechanically with manpower. For the handling of the transfer flask, a crane with a capacity of 5 metric tons is necessary. During installation of the lid the mobile water pool is filled with demineralized water and the shielded loading passage is taken away. After that the lid is put on the cask. After drainage, the mobile water pool is disassembled, and the cask is dewatered. Finally leak tests of all seals are made. The achieved leakage rate was -5 Pa x I/s. The work in FRM was done between 03.02.99 and 12.02.99 including a dry run and leak test. (author)

Hybrid PN-SN Calculations with SAAF for the Multiscale Transport Capability in Rattlesnake

Energy Technology Data Exchange (ETDEWEB)

Wang, Yaqi; Schunert, Sebastian; DeHart, Mark; Martineau, Richard

2016-05-01

Two interface conditions, the Lagrange multiplier method and the upwinding method, for hybrid \\pn-\\sn calculations is proposed for the self-adjoint angular flux (SAAF) formulation of the transport equation using the continuous finite element method (FEM) for spatial discretization. These interface conditions are implemented in Rattlesnake, the radiation transport application built on MOOSE, for the on-going multiscale transport simulation effort at INL. For smoothing the solution at the interface for the Lagrange multiplier method, a method based on \\sn Lagrange interpolation on the sphere is proposed. Numerical results indicate that the interface conditions give the expected convergence.

The RanGTP pathway: from nucleo-cytoplasmic transport to spindle assembly and beyond

Directory of Open Access Journals (Sweden)

Tommaso eCavazza

2016-01-01

Full Text Available The small GTPase Ran regulates the interaction of transport receptors with a number of cellular cargo proteins. The high affinity binding of the GTP-bound form of Ran to import receptors promotes cargo release, whereas its binding to export receptors stabilizes their interaction with the cargo. This basic mechanism linked to the asymmetric distribution of the two nucleotide-bound forms of Ran between the nucleus and the cytoplasm generates a switch like mechanism controlling nucleo-cytoplasmic transport. Since 1999, we have known that after nuclear envelope breakdown (NEBD Ran and the above transport receptors also provide a local control over the activity of factors driving spindle assembly and regulating other aspects of cell division. The identification and functional characterization of RanGTP mitotic targets is providing novel insights into mechanisms essential for cell division. Here we review our current knowledge on the RanGTP system and its regulation and we focus on the recent advances made through the characterization of its mitotic targets. We then briefly review the novel functions of the pathway that were recently described. Altogether, the RanGTP system has moonlighting functions exerting a spatial control over protein interactions that drive specific functions depending on the cellular context.

Dynamic behaviour of diagnostic assemblies

International Nuclear Information System (INIS)

Pecinka, L.

1980-01-01

The methodology is shown of calculating the frequency spectrum of a diagnostic assembly. The oscillations of the assembly as a whole, of a fuel rod bundle, the assembly jacket and of the individual rods in the bundle were considered. The manufacture is suggested of a model assembly which would be used for testing forced vibrations using an experimental water loop. (M.S.)

The effect of gamma-ray transport on afterheat calculations for accident analysis

International Nuclear Information System (INIS)

Reyes, S.; Latkowski, J.F.; Sanz, J.

2000-01-01

Radioactive afterheat is an important source term for the release of radionuclides in fusion systems under accident conditions. Heat transfer calculations are used to determine time-temperature histories in regions of interest, but the true source term needs to be the effective afterheat, which considers the transport of penetrating gamma rays. Without consideration of photon transport, accident temperatures may be overestimated in others. The importance of this effect is demonstrated for a simple, one-dimensional problem. The significance of this effect depends strongly on the accident scenario being analyzed

Error reduction techniques for Monte Carlo neutron transport calculations

International Nuclear Information System (INIS)

Ju, J.H.W.

1981-01-01

Monte Carlo methods have been widely applied to problems in nuclear physics, mathematical reliability, communication theory, and other areas. The work in this thesis is developed mainly with neutron transport applications in mind. For nuclear reactor and many other applications, random walk processes have been used to estimate multi-dimensional integrals and obtain information about the solution of integral equations. When the analysis is statistically based such calculations are often costly, and the development of efficient estimation techniques plays a critical role in these applications. All of the error reduction techniques developed in this work are applied to model problems. It is found that the nearly optimal parameters selected by the analytic method for use with GWAN estimator are nearly identical to parameters selected by the multistage method. Modified path length estimation (based on the path length importance measure) leads to excellent error reduction in all model problems examined. Finally, it should be pointed out that techniques used for neutron transport problems may be transferred easily to other application areas which are based on random walk processes. The transport problems studied in this dissertation provide exceptionally severe tests of the error reduction potential of any sampling procedure. It is therefore expected that the methods of this dissertation will prove useful in many other application areas

Calculation of the coherent transport properties of a symmetric spin nanocontact

International Nuclear Information System (INIS)

Bourahla, B.; Khater, A.; Tigrine, R.

2009-01-01

A theoretical study is presented for the coherent transport properties of a magnetic nanocontact. In particular, we study a symmetric nanocontact between two identical waveguides composed of semi-infinite spin ordered ferromagnetic chains. The coherent transmission and reflection scattering cross sections via the nanocontact, for spin waves incident from the bulk waveguide, are calculated with the use of the matching method. The inter-atomic magnetic exchange on the nanocontact is allowed to vary to investigate the consequences of magnetic softening and hardening for the calculated spectra. Transmission spectra underline the filtering properties of the nanocontact. The localized spin density of states in the nanocontact domain is also calculated, and analyzed. The results yield an understanding of the relationship between coherent conductance and the structural configuration of the nanocontact.

U.S. Commercial Spent Nuclear Fuel Assembly Characteristics - 1968-2013

Energy Technology Data Exchange (ETDEWEB)

Hu, Jianwei [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Peterson, Joshua L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Gauld, Ian C. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Bowman, Stephen M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2016-09-01

Activities related to management of spent nuclear fuel (SNF) are increasing in the US and many other countries. Over 240,000 SNF assemblies have been discharged from US commercial reactors since the late 1960s. The enrichment and burnup of SNF have changed significantly over the past 40 years, and fuel assembly designs have also evolved. Understanding the general characteristics of SNF helps regulators and other stakeholders form overall strategies towards the final disposal of US SNF. This report documents a survey of all US commercial SNF assemblies in the GC-859 database and provides reference SNF source terms (e.g., nuclide inventories, decay heat, and neutron/photon emission) at various cooling times up to 200 years after fuel discharge. This study reviews the distribution and evolution of fuel parameters of all SNF assemblies discharged over the past 40 years. Assemblies were categorized into three groups based on discharge year, and the median burnups and enrichments of each group were used to establish representative cases. An extended burnup case was created for boiling water reactor (BWR) fuels, and another was created for the pressurized water reactor (PWR) fuels. Two additional cases were developed to represent the eight mixed oxide (MOX) fuel assemblies in the database. Burnup calculations were performed for each representative case. Realistic parameters for fuel design and operations were used to model the SNF and to provide reference fuel characteristics representative of the current inventory. Burnup calculations were performed using the ORIGEN code, which is part of the SCALE nuclear modeling and simulation code system. Results include total activity, decay heat, photon emission, neutron flux, gamma heat, and plutonium content, as well as concentrations for 115 significant nuclides. These quantities are important in the design, regulation, and operations of SNF storage, transportation, and disposal systems.

Management number identification method for nuclear fuel assembly

International Nuclear Information System (INIS)

Furuya, Nobuo; Mori, Kazuma.

1995-01-01

In the present invention, a management number indicated to appropriate portions of a fuel assembly can be read with no error for the management of nuclear fuel materials in the nuclear fuel assembly (counting management) and physical protection: PP. Namely, bar codes as a management number are printed by electrolytic polishing to one or more portions of a side surface of an upper nozzle of the assembly, an upper surface of a clamp and a side surface of a lower nozzle. The bar codes are read by a reader at one or more portions in a transporting path for transporting the fuel assembly and at a fuel detection device disposed in a fuel storage pool. The read signals are inputted to a computer. With such procedures, the nuclear fuel assembly can be identified with no error by reading the bar codes and without applying no danger to a human body. Since the reader is disposed in the course of the transportation and test for the assembly, and the read signals are inputted to the computer, the management for the counting number and PP is facilitated. (I.S.)

A method for local transport analysis in tokamaks with error calculation

International Nuclear Information System (INIS)

Hogeweij, G.M.D.; Hordosy, G.; Lopes Cardozo, N.J.

1989-01-01

Global transport studies have revealed that heat transport in a tokamak is anomalous, but cannot provide information about the nature of the anomaly. Therefore, local transport analysis is essential for the study of anomalous transport. However, the determination of local transport coefficients is not a trivial affair. Generally speaking one can either directly measure the heat diffusivity, χ, by means of heat pulse propagation analysis, or deduce the profile of χ from measurements of the profiles of the temperature, T, and the power deposition. Here we are concerned only with the latter method, the local power balance analysis. For the sake of clarity heat diffusion only is considered: ρ=-gradT/q (1) where ρ=κ -1 =(nχ) -1 is the heat resistivity and q is the heat flux per unit area. It is assumed that the profiles T(r) and q(r) are given with some experimental error. In practice T(r) is measured directly, e.g. from ECE spectroscopy, while q(r) is deduced from the power deposition and loss profiles. The latter cannot be measured directly and is partly determined on the basis of models. This complication will not be considered here. Since in eq. (1) the gradient of T appears, noise on T can severely affect the solution ρ. This means that in general some form of smoothing must be applied. A criterion is needed to select the optimal smoothing. Too much smoothing will wipe out the details, whereas with too little smoothing the noise will distort the reconstructed profile of ρ. Here a new method to solve eq. (1) is presented which expresses ρ(r) as a cosine-series. The coefficients of this series are given as linear combinations of the Fourier coefficients of the measured T- and q-profiles. This formulation allows 1) the stable and accurate calculation of the ρ-profile, and 2) the analytical calculation of the error in this profile. (author) 5 refs., 3 figs

Anisotropic scattering effect in calculations of nuclear reactor cells by the surface preseudosource method

International Nuclear Information System (INIS)

Laletin, N.I.; Sultanov, N.V.; Boyarinov, V.F.

1992-01-01

Estimation is fulfilled of an influence of scattering anisotropy on K ef the TRX and BAPL assemblies by the WIMS-D4 program in the transport (TA) and linear-anisotropic (LAA) approximations. It is shown that account for the scattering anisotropy in the LAA in comparison with TA decreases K ef by 0.8% for TRX assemblies and by 0.5-0.6% for BAPL ones. For more detailed account for the scattering anisotropy in calculations of cylindrical and cluster cells in the one-velocity approximation is developed a technique for account for the anisotropy in the methods of surface pseudosources

3D Monte-Carlo transport calculations of whole slab reactor cores: validation of deterministic neutronic calculation routes

International Nuclear Information System (INIS)

Palau, J.M.

2005-01-01

This paper presents how Monte-Carlo calculations (French TRIPOLI4 poly-kinetic code with an appropriate pre-processing and post-processing software called OVNI) are used in the case of 3-dimensional heterogeneous benchmarks (slab reactor cores) to reduce model biases and enable a thorough and detailed analysis of the performances of deterministic methods and their associated data libraries with respect to key neutron parameters (reactivity, local power). Outstanding examples of application of these tools are presented regarding the new numerical methods implemented in the French lattice code APOLLO2 (advanced self-shielding models, new IDT characteristics method implemented within the discrete-ordinates flux solver model) and the JEFF3.1 nuclear data library (checked against JEF2.2 previous file). In particular we have pointed out, by performing multigroup/point-wise TRIPOLI4 (assembly and core) calculations, the efficiency (in terms of accuracy and computation time) of the new IDT method developed in APOLLO2. In addition, by performing 3-dimensional TRIPOLI4 calculations of the whole slab core (few millions of elementary volumes), the high quality of the new JEFF3.1 nuclear data files and revised evaluations (U 235 , U 238 , Hf) for reactivity prediction of slab cores critical experiments has been stressed. As a feedback of the whole validation process, improvements in terms of nuclear data (mainly Hf capture cross-sections) and numerical methods (advanced quadrature formulas accounting validation results, validation of new self-shielding models, parallelization) are suggested to improve even more the APOLLO2-CRONOS2 standard calculation route. (author)

3D Monte-Carlo transport calculations of whole slab reactor cores: validation of deterministic neutronic calculation routes

Energy Technology Data Exchange (ETDEWEB)

Palau, J M [CEA Cadarache, Service de Physique des Reacteurs et du Cycle, Lab. de Projets Nucleaires, 13 - Saint-Paul-lez-Durance (France)

2005-07-01

This paper presents how Monte-Carlo calculations (French TRIPOLI4 poly-kinetic code with an appropriate pre-processing and post-processing software called OVNI) are used in the case of 3-dimensional heterogeneous benchmarks (slab reactor cores) to reduce model biases and enable a thorough and detailed analysis of the performances of deterministic methods and their associated data libraries with respect to key neutron parameters (reactivity, local power). Outstanding examples of application of these tools are presented regarding the new numerical methods implemented in the French lattice code APOLLO2 (advanced self-shielding models, new IDT characteristics method implemented within the discrete-ordinates flux solver model) and the JEFF3.1 nuclear data library (checked against JEF2.2 previous file). In particular we have pointed out, by performing multigroup/point-wise TRIPOLI4 (assembly and core) calculations, the efficiency (in terms of accuracy and computation time) of the new IDT method developed in APOLLO2. In addition, by performing 3-dimensional TRIPOLI4 calculations of the whole slab core (few millions of elementary volumes), the high quality of the new JEFF3.1 nuclear data files and revised evaluations (U{sup 235}, U{sup 238}, Hf) for reactivity prediction of slab cores critical experiments has been stressed. As a feedback of the whole validation process, improvements in terms of nuclear data (mainly Hf capture cross-sections) and numerical methods (advanced quadrature formulas accounting validation results, validation of new self-shielding models, parallelization) are suggested to improve even more the APOLLO2-CRONOS2 standard calculation route. (author)

MCNP and OMEGA criticality calculations

International Nuclear Information System (INIS)

Seifert, E.

1998-04-01

The reliability of OMEGA criticality calculations is shown by a comparison with calculations by the validated and widely used Monte Carlo code MCNP. The criticality of 16 assemblies with uranium as fissionable is calculated with the codes MCNP (Version 4A, ENDF/B-V cross sections), MCNP (Version 4B, ENDF/B-VI cross sections), and OMEGA. Identical calculation models are used for the three codes. The results are compared mutually and with the experimental criticality of the assemblies. (orig.)

The performance of ENDF/B-V.2 nuclear data for fast reactor calculations

International Nuclear Information System (INIS)

Atkinson, C.A.; Collins, P.J.

1987-01-01

Calculations with ENDF/B-V.2 data have been made for twenty-five fast-spectrum integral assemblies covering a wide range of sizes and compositions. Analysis was done by transport codes with refined cross section processing methods and detailed reactor modelling. The predictions of fission rate distributions and control rod worths were emphasized for the more prototypic benchmark cores. The results show considerable improvements in agreement with experiment compared with analysis using ENDF/B-IV data, but it is apparent that significant errors remain for fast reactor design calculations

Radiation doses from radiation sources of neutrons and photons by different computer calculation

International Nuclear Information System (INIS)

Siciliano, F.; Lippolis, G.; Bruno, S.G.

1995-12-01

In the present paper the calculation technique aspects of dose rate from neutron and photon radiation sources are covered with reference both to the basic theoretical modeling of the MERCURE-4, XSDRNPM-S and MCNP-3A codes and from practical point of view performing safety analyses of irradiation risk of two transportation casks. The input data set of these calculations -regarding the CEN 10/200 HLW container and dry PWR spent fuel assemblies shipping cask- is frequently commented as for as connecting points of input data and understanding theoric background are concerned

Spectral measurements in critical assemblies: MCNP specifications and calculated results

Energy Technology Data Exchange (ETDEWEB)

Stephanie C. Frankle; Judith F. Briesmeister

1999-12-01

Recently, a suite of 86 criticality benchmarks for the Monte Carlo N-Particle (MCNP) transport code was developed, and the results of testing the ENDF/B-V and ENDF/B-VI data (through Release 2) were published. In addition to the standard k{sub eff} measurements, other experimental measurements were performed on a number of these benchmark assemblies. In particular, the Cross Section Evaluation Working Group (CSEWG) specifications contain experimental data for neutron leakage and central-flux measurements, central-fission ratio measurements, and activation ratio measurements. Additionally, there exists another set of fission reaction-rate measurements performed at the National Institute of Standards and Technology (NIST) utilizing a {sup 252}Cf source. This report will describe the leakage and central-flux measurements and show a comparison of experimental data to MCNP simulations performed using the ENDF/B-V and B-VI (Release 2) data libraries. Central-fission and activation reaction-rate measurements will be described, and the comparison of experimental data to MCNP simulations using available data libraries for each reaction of interest will be presented. Finally, the NIST fission reaction-rate measurements will be described. A comparison of MCNP results published previously with the current MCNP simulations will be presented for the NIST measurements, and a comparison of the current MCNP simulations to the experimental measurements will be presented.

Spectral measurements in critical assemblies: MCNP specifications and calculated results

International Nuclear Information System (INIS)

Frankle, Stephanie C.; Briesmeister, Judith F.

1999-01-01

Recently, a suite of 86 criticality benchmarks for the Monte Carlo N-Particle (MCNP) transport code was developed, and the results of testing the ENDF/B-V and ENDF/B-VI data (through Release 2) were published. In addition to the standard k eff measurements, other experimental measurements were performed on a number of these benchmark assemblies. In particular, the Cross Section Evaluation Working Group (CSEWG) specifications contain experimental data for neutron leakage and central-flux measurements, central-fission ratio measurements, and activation ratio measurements. Additionally, there exists another set of fission reaction-rate measurements performed at the National Institute of Standards and Technology (NIST) utilizing a 252 Cf source. This report will describe the leakage and central-flux measurements and show a comparison of experimental data to MCNP simulations performed using the ENDF/B-V and B-VI (Release 2) data libraries. Central-fission and activation reaction-rate measurements will be described, and the comparison of experimental data to MCNP simulations using available data libraries for each reaction of interest will be presented. Finally, the NIST fission reaction-rate measurements will be described. A comparison of MCNP results published previously with the current MCNP simulations will be presented for the NIST measurements, and a comparison of the current MCNP simulations to the experimental measurements will be presented

Structural instability of atmospheric flows under perturbations of the mass balance and effect in transport calculations

International Nuclear Information System (INIS)

Núñez, M A; Mendoza, R

2015-01-01

Several methods to estimate the velocity field of atmospheric flows, have been proposed to the date for applications such as emergency response systems, transport calculations and for budget studies of all kinds. These applications require a wind field that satisfies the conservation of mass but, in general, estimated wind fields do not satisfy exactly the continuity equation. An approach to reduce the effect of using a divergent wind field as input in the transport-diffusion equations, was proposed in the literature. In this work, a linear local analysis of a wind field, is used to show analytically that the perturbation of a large-scale nondivergent flow can yield a divergent flow with a substantially different structure. The effects of these structural changes in transport calculations are illustrated by means of analytic solutions of the transport equation

New nonlinear methods for linear transport calculations

International Nuclear Information System (INIS)

Adams, M.L.

1993-01-01

We present a new family of methods for the numerical solution of the linear transport equation. With these methods an iteration consists of an 'S N sweep' followed by an 'S 2 -like' calculation. We show, by analysis as well as numerical results, that iterative convergence is always rapid. We show that this rapid convergence does not depend on a consistent discretization of the S 2 -like equations - they can be discretized independently from the S N equations. We show further that independent discretizations can offer significant advantages over consistent ones. In particular, we find that in a wide range of problems, an accurate discretization of the S 2 -like equation can be combined with a crude discretization of the S N equations to produce an accurate S N answer. We demonstrate this by analysis as well as numerical results. (orig.)

Charge splitters and charge transport junctions based on guanine quadruplexes

Science.gov (United States)

Sha, Ruojie; Xiang, Limin; Liu, Chaoren; Balaeff, Alexander; Zhang, Yuqi; Zhang, Peng; Li, Yueqi; Beratan, David N.; Tao, Nongjian; Seeman, Nadrian C.

2018-04-01

Self-assembling circuit elements, such as current splitters or combiners at the molecular scale, require the design of building blocks with three or more terminals. A promising material for such building blocks is DNA, wherein multiple strands can self-assemble into multi-ended junctions, and nucleobase stacks can transport charge over long distances. However, nucleobase stacking is often disrupted at junction points, hindering electric charge transport between the two terminals of the junction. Here, we show that a guanine-quadruplex (G4) motif can be used as a connector element for a multi-ended DNA junction. By attaching specific terminal groups to the motif, we demonstrate that charges can enter the structure from one terminal at one end of a three-way G4 motif, and can exit from one of two terminals at the other end with minimal carrier transport attenuation. Moreover, we study four-way G4 junction structures by performing theoretical calculations to assist in the design and optimization of these connectors.

Spent Fuel Pool Dose Rate Calculations Using Point Kernel and Hybrid Deterministic-Stochastic Shielding Methods

International Nuclear Information System (INIS)

Matijevic, M.; Grgic, D.; Jecmenica, R.

2016-01-01

This paper presents comparison of the Krsko Power Plant simplified Spent Fuel Pool (SFP) dose rates using different computational shielding methodologies. The analysis was performed to estimate limiting gamma dose rates on wall mounted level instrumentation in case of significant loss of cooling water. The SFP was represented with simple homogenized cylinders (point kernel and Monte Carlo (MC)) or cuboids (MC) using uranium, iron, water, and dry-air as bulk region materials. The pool is divided on the old and new section where the old one has three additional subsections representing fuel assemblies (FAs) with different burnup/cooling time (60 days, 1 year and 5 years). The new section represents the FAs with the cooling time of 10 years. The time dependent fuel assembly isotopic composition was calculated using ORIGEN2 code applied to the depletion of one of the fuel assemblies present in the pool (AC-29). The source used in Microshield calculation is based on imported isotopic activities. The time dependent photon spectra with total source intensity from Microshield multigroup point kernel calculations was then prepared for two hybrid deterministic-stochastic sequences. One is based on SCALE/MAVRIC (Monaco and Denovo) methodology and another uses Monte Carlo code MCNP6.1.1b and ADVANTG3.0.1. code. Even though this model is a fairly simple one, the layers of shielding materials are thick enough to pose a significant shielding problem for MC method without the use of effective variance reduction (VR) technique. For that purpose the ADVANTG code was used to generate VR parameters (SB cards in SDEF and WWINP file) for MCNP fixed-source calculation using continuous energy transport. ADVATNG employs a deterministic forward-adjoint transport solver Denovo which implements CADIS/FW-CADIS methodology. Denovo implements a structured, Cartesian-grid SN solver based on the Koch-Baker-Alcouffe parallel transport sweep algorithm across x-y domain blocks. This was first

Analysis of measurements for a uranium-free core experiment at the BFS-2 critical assembly

Energy Technology Data Exchange (ETDEWEB)

Hunter, Stuart [Japan Nuclear Cycle Development Inst., Oarai, Ibaraki (Japan). Oarai Engineering Center

1999-04-01

This document describes a series of calculations that were carried out to model various measurements from the BFS-58-1-I1 experiment. BFS-58-1-I1 was a mock-up of a uranium-free, Pu burning core at BFS-2, a Russian critical assembly operated by IPPE. The experiment measured values of Keff, Na void reactivity worth, material sample reactivity worths and reaction rate ratios. The experiments were modelled using a number of different methods. Basic nuclear data was taken from JENDL-3.2, in either 70 or 18 groups. Cross-section data for the various material regions of the assembly were calculated by either SLAROM or CASUP; the heterogeneous structure of the core regions was modelled in these calculations, with 3 different options considered for representing the (essentially 2D) geometry of the assembly components in a 1D cell model. Whole reactor calculations of flux and Keff were done using both a diffusion model (CITATION) and a transport model (TWOTRAN2), both using an RZ geometry. Reactivity worths were calculated both directly from differences in Keff values and by using the exact perturbation calculations of PERKY and SN-PERT (for CITATION and TWOTRAN2, respectively). Initial calculations included a number of inaccuracies in the assembly representation, a result of communication difficulties between JNC and IPPE; these errors were removed for the final calculations that are presented. Calculations for the experiments have also been carried out in Russia (IPPE) and France (CEA) as part of an international comparison exercise, some of those results are also presented here. The calculated value of Keff was 1.1%{delta}k/k higher than the measured value, Na void worth C/E values were {approx}1.06; these results were considered to be reasonable. (Discrepancies in certain Na void values were probably due to experimental causes , though the effect should be investigated in any future experiments.) Several sample worth values were small compared with calculational

Transport calculations for a 14.8 MeV neutron beam in a water phantom

International Nuclear Information System (INIS)

Goetsch, S.J.

1981-01-01

A coupled neutron/photon Monte Carlo radiation transport code (MORSE-CG) has been used to calculate neutron and photon doses in a water phantom irradiated by 14.8 MeV neutrons from the Gas Target Neutron Source. The source-collimator-phantom geometry was carefully simulated. Results of calculations utilizing two different statistical estimators (next-collision and track-length) are presented

Calculation of health risks from spent-nuclear-fuel transportation accidents

International Nuclear Information System (INIS)

Chen, S.Y.; Yuan, Y.C.

1987-01-01

Models developed to analyze potential radiological health risks from various accident scenarios during transportation of spent nuclear fuels are described. The models are designed both for detailed route-specific risk analyses and for use in conducting overall risk analyses for route selection and related decision-making activities. The radiological risks calculated include individual dose commitments, collective dose commitments, and long-term (100-year) environmental dose commitments to a population following release of radioactivity. To facilitate route-specific analysis, a state-level database was developed and incorporated into the model. Route-specific analysis is demonstrated by the calculation of radiological risks resulting from various accident scenarios, as postulated by the recent US Nuclear Regulatory Commission Modal Study, for four representative states selected from various regions of the United States. 10 refs., 3 figs., 3 tabs

Development and validation of a criticality calculation scheme based on French deterministic transport codes

International Nuclear Information System (INIS)

Santamarina, A.

1991-01-01

A criticality-safety calculational scheme using the automated deterministic code system, APOLLO-BISTRO, has been developed. The cell/assembly code APOLLO is used mainly in LWR and HCR design calculations, and its validation spans a wide range of moderation ratios, including voided configurations. Its recent 99-group library and self-shielded cross-sections has been extensively qualified through critical experiments and PWR spent fuel analysis. The PIC self-shielding formalism enables a rigorous treatment of the fuel double heterogeneity in dissolver medium calculations. BISTRO is an optimized multidimensional SN code, part of the modular CCRR package used mainly in FBR calculations. The APOLLO-BISTRO scheme was applied to the 18 experimental benchmarks selected by the OECD/NEACRP Criticality Calculation Working Group. The Calculation-Experiment discrepancy was within ± 1% in ΔK/K and always looked consistent with the experimental uncertainty margin. In the critical experiments corresponding to a dissolver type benchmark, our tools computed a satisfactory Keff. In the VALDUC fuel storage experiments, with hafnium plates, the computed Keff ranged between 0.994 and 1.003 for the various watergaps spacing the fuel clusters from the absorber plates. The APOLLO-KENOEUR statistic calculational scheme, based on the same self-shielded multigroup library, supplied consistent results within 0.3% in ΔK/K. (Author)

Development of whole core thermal-hydraulic analysis program ACT. 4. Simplified fuel assembly model and parallelization by MPI

International Nuclear Information System (INIS)

Ohshima, Hiroyuki

2001-10-01

A whole core thermal-hydraulic analysis program ACT is being developed for the purpose of evaluating detailed in-core thermal hydraulic phenomena of fast reactors including the effect of the flow between wrapper-tube walls (inter-wrapper flow) under various reactor operation conditions. As appropriate boundary conditions in addition to a detailed modeling of the core are essential for accurate simulations of in-core thermal hydraulics, ACT consists of not only fuel assembly and inter-wrapper flow analysis modules but also a heat transport system analysis module that gives response of the plant dynamics to the core model. This report describes incorporation of a simplified model to the fuel assembly analysis module and program parallelization by a message passing method toward large-scale simulations. ACT has a fuel assembly analysis module which can simulate a whole fuel pin bundle in each fuel assembly of the core and, however, it may take much CPU time for a large-scale core simulation. Therefore, a simplified fuel assembly model that is thermal-hydraulically equivalent to the detailed one has been incorporated in order to save the simulation time and resources. This simplified model is applied to several parts of fuel assemblies in a core where the detailed simulation results are not required. With regard to the program parallelization, the calculation load and the data flow of ACT were analyzed and the optimum parallelization has been done including the improvement of the numerical simulation algorithm of ACT. Message Passing Interface (MPI) is applied to data communication between processes and synchronization in parallel calculations. Parallelized ACT was verified through a comparison simulation with the original one. In addition to the above works, input manuals of the core analysis module and the heat transport system analysis module have been prepared. (author)

Transport and hydrodynamic calculations of direct photons at FAIR

International Nuclear Information System (INIS)

Baeuchle, Bjorn; Bleicher, Marcus

2011-01-01

The microscopic transport model UrQMD and a micro + macro hybrid model are used to calculate direct photon spectra from U+U-collisions at E lab =35 A GeV as will be measured by the CBM Collaboration at FAIR. In the hybrid model, the intermediate high-density part of the nuclear interaction is described with ideal 3+1-dimensional hydrodynamics. Different equations of state of the matter created in the heavy-ion collisions are investigated and the resulting spectra of direct photons are predicted. The emission patterns of direct photons in space and time are discussed.

Evaluation of scattering laws and cross sections for calculation of production and transport of cold and ultracold neutrons

International Nuclear Information System (INIS)

Bernnat, W.; Keinert, J.; Mattes, M.

2004-01-01

For the calculation of neutron spectra in cold and super thermal sources scattering laws for a variety of liquid and solid cyrogenic materials were evaluated and prepared for use in deterministic and Monte Carlo transport calculations. For moderator materials like liquid and solid H 2 O, liquid He, liquid D 2 O, liquid and solid H 2 and D 2 , solid CH 4 and structure materials such as Al, Bi, Pb, ZrHx, and graphite scattering law data and cross sections are available. The evaluated data were validated by comparison with measured cross sections and comparison of measured and calculated neutron spectra as far as available. Further applications are the calculation of production and transport and storing of ultra cold neutrons (UCN) in different UCN sources. The data structures of the evaluated data are prepared for the common S N -transport codes and the Monte Carlo Code MCNP. (orig.)

Towards quantitative accuracy in first-principles transport calculations: The GW method applied to alkane/gold junctions

DEFF Research Database (Denmark)

Strange, Mikkel; Thygesen, Kristian Sommer

2011-01-01

-electron interactions are described by th=e many-body GW approximation. The conductance follows an exponential length dependence: Gn = Gc exp(-βn). The main difference from standard density functional theory (DFT) calculations is a significant reduction of the contact conductance, Gc, due to an improved alignment......The calculation of the electronic conductance of nanoscale junctions from first principles is a long-standing problem in the field of charge transport. Here we demonstrate excellent agreement with experiments for the transport properties of the gold/alkanediamine benchmark system when electron...

Nonlinear Projective-Iteration Methods for Solving Transport Problems on Regular and Unstructured Grids

International Nuclear Information System (INIS)

Dmitriy Y. Anistratov; Adrian Constantinescu; Loren Roberts; William Wieselquist

2007-01-01

This is a project in the field of fundamental research on numerical methods for solving the particle transport equation. Numerous practical problems require to use unstructured meshes, for example, detailed nuclear reactor assembly-level calculations, large-scale reactor core calculations, radiative hydrodynamics problems, where the mesh is determined by hydrodynamic processes, and well-logging problems in which the media structure has very complicated geometry. Currently this is an area of very active research in numerical transport theory. main issues in developing numerical methods for solving the transport equation are the accuracy of the numerical solution and effectiveness of iteration procedure. The problem in case of unstructured grids is that it is very difficult to derive an iteration algorithm that will be unconditionally stable

Comparison of Monte Carlo method and deterministic method for neutron transport calculation

International Nuclear Information System (INIS)

Mori, Takamasa; Nakagawa, Masayuki

1987-01-01

The report outlines major features of the Monte Carlo method by citing various applications of the method and techniques used for Monte Carlo codes. Major areas of its application include analysis of measurements on fast critical assemblies, nuclear fusion reactor neutronics analysis, criticality safety analysis, evaluation by VIM code, and calculation for shielding. Major techniques used for Monte Carlo codes include the random walk method, geometric expression method (combinatorial geometry, 1, 2, 4-th degree surface and lattice geometry), nuclear data expression, evaluation method (track length, collision, analog (absorption), surface crossing, point), and dispersion reduction (Russian roulette, splitting, exponential transform, importance sampling, corrected sampling). Major features of the Monte Carlo method are as follows: 1) neutron source distribution and systems of complex geometry can be simulated accurately, 2) physical quantities such as neutron flux in a place, on a surface or at a point can be evaluated, and 3) calculation requires less time. (Nogami, K.)

Gyrokinetic Calculations of Microinstabilities and Transport During RF H-Modes on Alcator C-Mod

International Nuclear Information System (INIS)

Redi, M.H.; Fiore, C.; Bonoli, P.; Bourdelle, C.; Budny, R.; Dorland, W.D.; Ernst, D.; Hammett, G.; Mikkelsen, D.; Rice, J.; Wukitch, S.

2002-01-01

Physics understanding for the experimental improvement of particle and energy confinement is being advanced through massively parallel calculations of microturbulence for simulated plasma conditions. The ultimate goal, an experimentally validated, global, non-local, fully nonlinear calculation of plasma microturbulence is still not within reach, but extraordinary progress has been achieved in understanding microturbulence, driving forces and the plasma response in recent years. In this paper we discuss gyrokinetic simulations of plasma turbulence being carried out to examine a reproducible, H-mode, RF heated experiment on the Alcator CMOD tokamak3, which exhibits an internal transport barrier (ITB). This off axis RF case represents the early phase of a very interesting dual frequency RF experiment, which shows density control with central RF heating later in the discharge. The ITB exhibits steep, spontaneous density peaking: a reduction in particle transport occurring without a central particle source. Since the central temperature is maintained while the central density is increasing, this also suggests a thermal transport barrier exists. TRANSP analysis shows that ceff drops inside the ITB. Sawtooth heat pulse analysis also shows a localized thermal transport barrier. For this ICRF EDA H-mode, the minority resonance is at r/a * 0.5 on the high field side. There is a normal shear profile, with q monotonic

A spectrum correction method for fuel assembly rehomogenization

International Nuclear Information System (INIS)

Lee, Kyung Taek; Cho, Nam Zin

2004-01-01

To overcome the limitation of existing homogenization methods based on the single assembly calculation with zero current boundary condition, we propose a new rehomogenization method, named spectrum correction method (SCM), consisting of the multigroup energy spectrum approximation by spectrum correction and the condensed two-group heterogeneous single assembly calculations with non-zero current boundary condition. In SCM, the spectrum shifting phenomena caused by current across assembly interfaces are considered by the spectrum correction at group condensation stage at first. Then, heterogeneous single assembly calculations with two-group cross sections condensed by using corrected multigroup energy spectrum are performed to obtain rehomogenized nodal diffusion parameters, i.e., assembly-wise homogenized cross sections and discontinuity factors. To evaluate the performance of SCM, it was applied to the analytic function expansion nodal (AFEN) method and several test problems were solved. The results show that SCM can reduce the errors significantly both in multiplication factors and assembly averaged power distributions

Evaluation of scattering laws and cross sections for calculation of production and transport of cold and ultracold neutrons

Energy Technology Data Exchange (ETDEWEB)

Bernnat, W.; Keinert, J.; Mattes, M. [Inst. for Nuclear Energy and Energy Systems, Univ. of Stuttgart, Stuttgart (Germany)

2004-03-01

For the calculation of neutron spectra in cold and super thermal sources scattering laws for a variety of liquid and solid cyrogenic materials were evaluated and prepared for use in deterministic and Monte Carlo transport calculations. For moderator materials like liquid and solid H{sub 2}O, liquid He, liquid D{sub 2}O, liquid and solid H{sub 2} and D{sub 2}, solid CH{sub 4} and structure materials such as Al, Bi, Pb, ZrHx, and graphite scattering law data and cross sections are available. The evaluated data were validated by comparison with measured cross sections and comparison of measured and calculated neutron spectra as far as available. Further applications are the calculation of production and transport and storing of ultra cold neutrons (UCN) in different UCN sources. The data structures of the evaluated data are prepared for the common S{sub N}-transport codes and the Monte Carlo Code MCNP. (orig.)

The MARVEL assembly for neutron multiplication

Energy Technology Data Exchange (ETDEWEB)

David L. Chichester; Mathew T. Kinlaw

2013-10-01

A new multiplying test assembly is under development at Idaho National Laboratory to support research, validation, evaluation, and learning. The item is comprised of three stacked, highly-enriched uranium (HEU) cylinders, each 11.4 cm in diameter and having a combined height of up to 11.7 cm. The combined mass of all three cylinders is 20.3 kg of HEU. Calculations for the bare configuration of the assembly indicate a multiplication level of >3.5 (keff=0.72). Reflected configurations of the assembly, using either polyethylene or tungsten, are possible and have the capability of raising the assembly's multiplication level to greater than 10. This paper describes simulations performed to assess the assembly's multiplication level under different conditions and describes the resources available at INL to support the use of these materials. We also describe some preliminary calculations and test activities using the assembly to study neutron multiplication.

The MARVEL assembly for neutron multiplication.

Science.gov (United States)

Chichester, David L; Kinlaw, Mathew T

2013-10-01

A new multiplying test assembly is under development at Idaho National Laboratory to support research, validation, evaluation, and learning. The item is comprised of three stacked, highly-enriched uranium (HEU) cylinders, each 11.4 cm in diameter and having a combined height of up to 11.7 cm. The combined mass of all three cylinders is 20.3 kg of HEU. Calculations for the bare configuration of the assembly indicate a multiplication level of >3.5 (k(eff)=0.72). Reflected configurations of the assembly, using either polyethylene or tungsten, are possible and have the capability of raising the assembly's multiplication level to greater than 10. This paper describes simulations performed to assess the assembly's multiplication level under different conditions and describes the resources available at INL to support the use of these materials. We also describe some preliminary calculations and test activities using the assembly to study neutron multiplication. Copyright © 2013 Elsevier Ltd. All rights reserved.

Assaying Used Nuclear Fuel Assemblies Using Lead Slowing-Down Spectroscopy and Singular Value Decomposition

International Nuclear Information System (INIS)

Kulisek, Jonathan A.; Anderson, Kevin K.; Casella, Andrew M.; Gesh, Christopher J.; Warren, Glen A.

2013-01-01

This study investigates the use of a Lead Slowing-Down Spectrometer (LSDS) for the direct and independent measurement of fissile isotopes in light-water nuclear reactor fuel assemblies. The current study applies MCNPX, a Monte Carlo radiation transport code, to simulate the measurement of the assay of the used nuclear fuel assemblies in the LSDS. An empirical model has been developed based on the calibration of the LSDS to responses generated from the simulated assay of six well-characterized fuel assemblies. The effects of self-shielding are taken into account by using empirical basis vectors calculated from the singular value decomposition (SVD) of a matrix containing the self-shielding functions from the assay of assemblies in the calibration set. The performance of the empirical algorithm was tested on version 1 of the Next-Generation Safeguards Initiative (NGSI) used fuel library consisting of 64 assemblies, as well as a set of 27 diversion assemblies, both of which were developed by Los Alamos National Laboratory. The potential for direct and independent assay of the sum of the masses of Pu-239 and Pu-241 to within 2%, on average, has been demonstrated

Electromagnetic energy transport in nanoparticle chains via dark plasmon modes.

Science.gov (United States)

Solis, David; Willingham, Britain; Nauert, Scott L; Slaughter, Liane S; Olson, Jana; Swanglap, Pattanawit; Paul, Aniruddha; Chang, Wei-Shun; Link, Stephan

2012-03-14

Using light to exchange information offers large bandwidths and high speeds, but the miniaturization of optical components is limited by diffraction. Converting light into electron waves in metals allows one to overcome this problem. However, metals are lossy at optical frequencies and large-area fabrication of nanometer-sized structures by conventional top-down methods can be cost-prohibitive. We show electromagnetic energy transport with gold nanoparticles that were assembled into close-packed linear chains. The small interparticle distances enabled strong electromagnetic coupling causing the formation of low-loss subradiant plasmons, which facilitated energy propagation over many micrometers. Electrodynamic calculations confirmed the dark nature of the propagating mode and showed that disorder in the nanoparticle arrangement enhances energy transport, demonstrating the viability of using bottom-up nanoparticle assemblies for ultracompact opto-electronic devices. © 2012 American Chemical Society

CLEAR (Calculates Logical Evacuation And Response): A generic transportation network model for the calculation of evacuation time estimates

International Nuclear Information System (INIS)

Moeller, M.P.; Desrosiers, A.E.; Urbanik, T. II

1982-03-01

This paper describes the methodology and application of the computer model CLEAR (Calculates Logical Evacuation And Response) which estimates the time required for a specific population density and distribution to evacuate an area using a specific transportation network. The CLEAR model simulates vehicle departure and movement on a transportation network according to the conditions and consequences of traffic flow. These include handling vehicles at intersecting road segments, calculating the velocity of travel on a road segment as a function of its vehicle density, and accounting for the delay of vehicles in traffic queues. The program also models the distribution of times required by individuals to prepare for an evacuation. In order to test its accuracy, the CLEAR model was used to estimate evacuation times for the emergency planning zone surrounding the Beaver Valley Nuclear Power Plant. The Beaver Valley site was selected because evacuation time estimates had previously been prepared by the licensee, Duquesne Light, as well as by the Federal Emergency Management Agency and the Pennsylvania Emergency Management Agency. A lack of documentation prevented a detailed comparison of the estimates based on the CLEAR model and those obtained by Duquesne Light. However, the CLEAR model results compared favorably with the estimates prepared by the other two agencies. (author)

CLEAR (Calculates Logical Evacuation And Response): A Generic Transportation Network Model for the Calculation of Evacuation Time Estimates

Energy Technology Data Exchange (ETDEWEB)

Moeller, M. P.; Urbanik, II, T.; Desrosiers, A. E.

1982-03-01

This paper describes the methodology and application of the computer model CLEAR (Calculates Logical Evacuation And Response) which estimates the time required for a specific population density and distribution to evacuate an area using a specific transportation network. The CLEAR model simulates vehicle departure and movement on a transportation network according to the conditions and consequences of traffic flow. These include handling vehicles at intersecting road segments, calculating the velocity of travel on a road segment as a function of its vehicle density, and accounting for the delay of vehicles in traffic queues. The program also models the distribution of times required by individuals to prepare for an evacuation. In order to test its accuracy, the CLEAR model was used to estimate evacuatlon tlmes for the emergency planning zone surrounding the Beaver Valley Nuclear Power Plant. The Beaver Valley site was selected because evacuation time estimates had previously been prepared by the licensee, Duquesne Light, as well as by the Federal Emergency Management Agency and the Pennsylvania Emergency Management Agency. A lack of documentation prevented a detailed comparison of the estimates based on the CLEAR model and those obtained by Duquesne Light. However, the CLEAR model results compared favorably with the estimates prepared by the other two agencies.

MOCA, Criticality of VVER Reactor Hexagonal Fuel Assemblies

International Nuclear Information System (INIS)

KYNCL, Jan

1994-01-01

1 - Description of program or function: Criticality problem in neutron transport for hexagonal fuel assembly in VVER nuclear reactor. The assembly is assumed to be either arranged in an infinite hexagonal array or placed in vacuum. The problem is solved in three- dimensional geometry, using standard energy group formalism and assuming that effective scattering cross sections are presented as Legendre polynomial expansions. The code evaluates ten different physical quantities, e.g. multiplication factor, neutron flux per energy group and spatial zone, integrated over angle and power in any zone of the assembly. 2 - Method of solution: Monte Carlo method of successive generations is applied. Computation proceeds according to an analog random process. The code is organized into three blocks: In the first block, the input data are converted to quantities for use in the Monte Carlo calculation. An initial neutron distribution is calculated, which corresponds to a fission spectrum uniform in spatial and angular variables. The main calculations are carried out in the second block (subroutine PROC2). This block is subdivided into geometrical and physical parts. Neutron tracks in individual zones and groups as well as probabilities for the formation of secondary neutrons are calculated. In the third block (subroutine PROC3), the results are evaluated statistically. Effective multiplication coefficients, the neutron flux per group and zone, and respective errors are computed. These quantities serve as a basis for the evaluation of other quantities. The results are either printed or stored for future evaluations. 3 - Restrictions on the complexity of the problem: In the PC version of the program, the maximum number of neutrons is 1000, the maximum number of energy groups is 4, and the maximum number of material compositions is 15. Angular expansion of scattering cross sections is allowed up to P10. These restrictions can easily be removed by increasing input parameters and

Computer program for calculation of complex chemical equilibrium compositions and applications. Supplement 1: Transport properties

Science.gov (United States)

Gordon, S.; Mcbride, B.; Zeleznik, F. J.

1984-01-01

An addition to the computer program of NASA SP-273 is given that permits transport property calculations for the gaseous phase. Approximate mixture formulas are used to obtain viscosity and frozen thermal conductivity. Reaction thermal conductivity is obtained by the same method as in NASA TN D-7056. Transport properties for 154 gaseous species were selected for use with the program.

Regulatory practices of radiation safety of SNF transportation in Russia

International Nuclear Information System (INIS)

Kuryndina, Lidia; Kuryndin, Anton; Stroganov, Anatoly

2008-01-01

This paper overviews current regulatory practices for the assurance of nuclear and radiation safety during railway transportation of SNF on the territory of Russian Federation from NPPs to longterm-storage of reprocessing sites. The legal and regulatory requirements (mostly compliant with IAEA ST-1), licensing procedure for NM transportation are discussed. The current procedure does not require a regulatory approval for each particular shipment if the SNF fully comply with the Rosatom's branch standard and is transported in approved casks. It has been demonstrated that SNF packages compliant with the branch standard, which is knowingly provide sufficient safety margin, will conform to the federal level regulations. The regulatory approval is required if a particular shipment does not comply with the branch standard. In this case, the shipment can be approved only after regulatory review of Applicant's documents to demonstrate that the shipment still conformant to the higher level (federal) regulations. The regulatory review frequently needs a full calculation test of the radiation safety assurance. This test can take a lot of time. That's why the special calculation tools were created in SEC NRS. These tools aimed for precision calculation of the radiation safety parameters by SNF transportation use preliminary calculated Green's functions. Such approach allows quickly simulate any source distribution and optimize spent fuel assemblies placement in cask due to the transport equation property of linearity relatively the source. The short description of calculation tools are presented. Also, the paper discusses foreseen implications related to transportation of mixed-oxide SNF. (author)

Efficient Ab-Initio Electron Transport Calculations for Heterostructures by the Nonequilibrium Green’s Function Method

Directory of Open Access Journals (Sweden)

Hirokazu Takaki

2014-01-01

Full Text Available We present an efficient computation technique for ab-initio electron transport calculations based on density functional theory and the nonequilibrium Green’s function formalism for application to heterostructures with two-dimensional (2D interfaces. The computational load for constructing the Green’s functions, which depends not only on the energy but also on the 2D Bloch wave vector along the interfaces and is thus catastrophically heavy, is circumvented by parallel computational techniques with the message passing interface, which divides the calculations of the Green’s functions with respect to energy and wave vectors. To demonstrate the computational efficiency of the present code, we perform ab-initio electron transport calculations of Al(100-Si(100-Al(100 heterostructures, one of the most typical metal-semiconductor-metal systems, and show their transmission spectra, density of states (DOSs, and dependence on the thickness of the Si layers.

An assessment of the transportation costs of shipping non-fuel assembly hardware to FWMS facilities

International Nuclear Information System (INIS)

Shappert, L.B.; Joy, D.S.; Johnson, P.E.; Danese, F.L.; Best, R.E.

1991-01-01

This study examines the cost of using Department of Energy (DOE) Office of Civilian Radioactive Waste Management (OCRWM) Initiative I casks for transporting 62,700 MTU of spent fuel plus associated non-fuel assembly hardware (NFAH) between reactor sites and either a monitored retrievable storage (MRS) or a repository facility. The study further considers the benefits of increasing the cell size of the Initiative I BWR cask baskets to accommodate the fuel plus channels (which also would decrease the capacity of the cask to carry BWR fuel without channels) and the use of a commercial, non-spent-fuel cask to carry compacted NFAH that could not be shipped integrally. Costs that are developed involve transportation charges, capital costs for casks, and canning costs of NFAH that have been separated from the fuel. The results indicate that significant cost savings are possible if NFAH is accepted into the Federal Waste Management System (FWMS) that is either integral with the spent fuel, or consolidated if it has been separated. Shipment of unconsolidated NFAH is very expensive. Transportation costs for shipping to a western repository are approximately 50 to 75% higher than the costs for shipping to an eastern MRS

GUIDE TO CALCULATING TRANSPORT EFFICIENCY OF AEROSOLS IN OCCUPATIONAL AIR SAMPLING SYSTEMS

Energy Technology Data Exchange (ETDEWEB)

Hogue, M.; Hadlock, D.; Thompson, M.; Farfan, E.

2013-11-12

This report will present hand calculations for transport efficiency based on aspiration efficiency and particle deposition losses. Because the hand calculations become long and tedious, especially for lognormal distributions of aerosols, an R script (R 2011) will be provided for each element examined. Calculations are provided for the most common elements in a remote air sampling system, including a thin-walled probe in ambient air, straight tubing, bends and a sample housing. One popular alternative approach would be to put such calculations in a spreadsheet, a thorough version of which is shared by Paul Baron via the Aerocalc spreadsheet (Baron 2012). To provide greater transparency and to avoid common spreadsheet vulnerabilities to errors (Burns 2012), this report uses R. The particle size is based on the concept of activity median aerodynamic diameter (AMAD). The AMAD is a particle size in an aerosol where fifty percent of the activity in the aerosol is associated with particles of aerodynamic diameter greater than the AMAD. This concept allows for the simplification of transport efficiency calculations where all particles are treated as spheres with the density of water (1g cm-3). In reality, particle densities depend on the actual material involved. Particle geometries can be very complicated. Dynamic shape factors are provided by Hinds (Hinds 1999). Some example factors are: 1.00 for a sphere, 1.08 for a cube, 1.68 for a long cylinder (10 times as long as it is wide), 1.05 to 1.11 for bituminous coal, 1.57 for sand and 1.88 for talc. Revision 1 is made to correct an error in the original version of this report. The particle distributions are based on activity weighting of particles rather than based on the number of particles of each size. Therefore, the mass correction made in the original version is removed from the text and the calculations. Results affected by the change are updated.

MORSE-C, Neutron Transport, Gamma Transport for Criticality Calculation by Monte-Carlo Method

International Nuclear Information System (INIS)

2002-01-01

1 - Description of program or function: MORSE-C is a Monte-Carlo code to solve the multiple energy group form of the Boltzmann transport equation in order to obtain the eigenvalue (multiplication) when fissionable materials are present. Cross sections for up to 100 energy groups may be employed. The angular scattering is treated by the usual Legendre expansion as used in the discrete ordinates codes. Up-scattering may be specified. The geometry is defined by relationships to general 1. or 2. degree surfaces. Array units may be specified. Output includes, besides the usual values of input quantities, plots of the geometry, calculated volumes and masses, and graphs of results to assist the user in determining the correctness of the problem's solution

A set of integrated environmental transport and diffusion models for calculating hazardous releases

International Nuclear Information System (INIS)

Pepper, D.W.

1996-01-01

A set of numerical transport and dispersion models is incorporated within a graphical interface shell to predict hazardous material released into the environment. The visual shell (EnviroView) consists of an object-oriented knowledge base, which is used for inventory control, site mapping and orientation, and monitoring of materials. Graphical displays of detailed sites, building locations, floor plans, and three-dimensional views within a room are available to the user using a point and click interface. In the event of a release to the environment, the user can choose from a selection of analytical, finite element, finite volume, and boundary element methods, which calculate atmospheric transport, groundwater transport, and dispersion within a building interior. The program runs on 486 personal computers under WINDOWS

MOx Depletion Calculation Benchmark

International Nuclear Information System (INIS)

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

2016-01-01

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

Numerical shoves and countershoves in electron transport calculations

International Nuclear Information System (INIS)

Filippone, W.L.

1986-01-01

The justification for applying the relatively complex (compared to S/sub n/) streaming ray (SR) algorithm to electron transport problems is its potential for doing rapid and accurate calculations. Because of the Lagrangian treatment of the cell-uncollided electrons, the only significant sources of error are the numerical treatment of the scattering kernel and the spatial differencing scheme used for the cell-collided electrons. Considerable progress has been made in reducing the former source of error. If one is willing to pay the price, the latter source of error can be reduced to any desired level by refining the mesh size or by using high-order differencing schemes. Here the method of numerical shoves and countershoves is introduced, which reduces spatial differencing errors using relatively little additional computational effort

Assembly procedure for Shot Loading Platform

International Nuclear Information System (INIS)

Routh, R.D.

1995-01-01

This supporting document describes the assembly procedure for the Shot Loading Platform. The Shot Loading Platform is used by multiple equipment removal projects to load shielding shot in the annular spaces of the equipment storage containers. The platform height is adjustable to accommodate different sizes of storage containers and transport assemblies

Evaluation and comparison of SN and Monte-Carlo charged particle transport calculations

International Nuclear Information System (INIS)

Hadad, K.

2000-01-01

A study was done to evaluate a 3-D S N charged particle transport code called SMARTEPANTS 1 and another 3-D Monte Carlo code called Integrated Tiger Series, ITS 2 . The evaluation study of SMARTEPANTS code was based on angular discretization and reflected boundary sensitivity whilst the evaluation of ITS was based on CPU time and variance reduction. The comparison of the two code was based on energy and charge deposition calculation in block of Gallium Arsenide with embedded gold cylinders. The result of evaluation tests shows that an S 8 calculation maintains both accuracy and speed and calculations with reflected boundaries geometry produces full symmetrical results. As expected for ITS evaluation, the CPU time and variance reduction are opposite to a point beyond which the history augmentation while increasing the CPU time do not result in variance reduction. The comparison test problem showed excellent agreement in total energy deposition calculations

Methodology for coupling computational fluid dynamics and integral transport neutronics

International Nuclear Information System (INIS)

Thomas, J. W.; Zhong, Z.; Sofu, T.; Downar, T. J.

2004-01-01

The CFD code STAR-CD was coupled to the integral transport code DeCART in order to provide high-fidelity, full physics reactor simulations. An interface program was developed to perform the tasks of mapping the STAR-CD mesh to the DeCART mesh, managing all communication between STAR-CD and DeCART, and monitoring the convergence of the coupled calculations. The interface software was validated by comparing coupled calculation results with those obtained using an independently developed interface program. An investigation into the convergence characteristics of coupled calculations was performed using several test models on a multiprocessor LINUX cluster. The results indicate that the optimal convergence of the coupled field calculation depends on several factors, to include the tolerance of the STAR-CD solution and the number of DeCART transport sweeps performed before exchanging data between codes. Results for a 3D, multi-assembly PWR problem on 12 PEs of the LINUX cluster indicate the best performance is achieved when the STAR-CD tolerance and number of DeCART transport sweeps are chosen such that the two fields converge at approximately the same rate. (authors)

Generalized Bloch Theorem for Complex Periodic Potentials - A Powerful Application to Quantum Transport Calculations

International Nuclear Information System (INIS)

Zhang, Xiaoguang; Varga, Kalman; Pantelides, Sokrates T

2007-01-01

Band-theoretic methods with periodically repeated supercells have been a powerful approach for ground-state electronic structure calculations, but have not so far been adapted for quantum transport problems with open boundary conditions. Here we introduce a generalized Bloch theorem for complex periodic potentials and use a transfer-matrix formulation to cast the transmission probability in a scattering problem with open boundary conditions in terms of the complex wave vectors of a periodic system with absorbing layers, allowing a band technique for quantum transport calculations. The accuracy and utility of the method is demonstrated by the model problems of the transmission of an electron over a square barrier and the scattering of a phonon in an inhomogeneous nanowire. Application to the resistance of a twin boundary in nanocrystalline copper yields excellent agreement with recent experimental data

Thermal analysis on NAC-STC spent fuel transport cask under different transport conditions

Energy Technology Data Exchange (ETDEWEB)

Xu, Yumei [Institute of Process Equipment, Zhejiang University, Hangzhou (China); Yang, Jian, E-mail: zdhjkz@zju.edu.cn [Institute of Process Equipment, Zhejiang University, Hangzhou (China); Xu, Chao; Wang, Weiping [Institute of Process Equipment, Zhejiang University, Hangzhou (China); Ma, Zhijun [Department of Material Engineering, South China University of Technology, Guangzhou (China)

2013-12-15

Highlights: • Spent fuel cask was investigated as a whole instead of fuel assembly alone. • The cask was successfully modeled and meshed after several simplifications. • Equivalence method was used to calculate the properties of parts. • Both the integral thermal field and peak values are captured to verify safety. • The temperature variations of key parts were also plotted. - Abstract: Transport casks used for conveying spent nuclear fuel are inseparably related to the safety of the whole reprocessing system for spent nuclear fuel. Thus they must be designed according to rigorous safety standards including thermal analysis. In this paper, for NAC-STC cask, a finite element model is established based on some proper simplifications on configurations and the heat transfer mechanisms. Considering the complex components and gaps, the equivalence method is presented to define their material properties. Then an equivalent convection coefficient is introduced to define boundary conditions. Finally, the temperature field is captured and analyzed under both normal and accident transport conditions by using ANSYS software. The validity of numerical calculation is given by comparing its results with theoretical calculation. Obtaining the integral distribution laws of temperature and peak temperature values of all vital components, the security of the cask can be evaluated and verified.

Numerical verification of the theory of coupled reactors for a deuterium critical assembly using MCNP5

International Nuclear Information System (INIS)

Hussein, M.S.; Bonin, H.W.; Lewis, B.J.

2013-01-01

The theory of multipoint coupled reactors developed by multi-group transport is verified by using the probabilistic transport code MCNP5. The verification was performed by calculating the multiplication factors (or criticality factors) and coupling coefficients for a two-region test reactor known as Deuterium Critical Assembly, (DCA). The variations of the criticality factors and the coupling coefficients were investigated by changing of the water levels in the inner and outer cores. The numerical results of the model developed with MCNP5 code were validated and verified against published results and the mathematical model based on coupled reactor theory. (author)

Numerical verification of the theory of coupled reactors for a deuterium critical assembly using MCNP5

Energy Technology Data Exchange (ETDEWEB)

Hussein, M.S.; Bonin, H.W.; Lewis, B.J., E-mail: mohamed.hussein@rmc.ca, E-mail: bonin-h@rmc.ca, E-mail: lewis-b@rmc.ca [Royal Military College of Canada, Dept. of Chemistry and Chemical Engineering, Kingston, Ontario (Canada)

2013-07-01

The theory of multipoint coupled reactors developed by multi-group transport is verified by using the probabilistic transport code MCNP5. The verification was performed by calculating the multiplication factors (or criticality factors) and coupling coefficients for a two-region test reactor known as Deuterium Critical Assembly, (DCA). The variations of the criticality factors and the coupling coefficients were investigated by changing of the water levels in the inner and outer cores. The numerical results of the model developed with MCNP5 code were validated and verified against published results and the mathematical model based on coupled reactor theory. (author)

Cross sections for electron and photon processes required by electron-transport calculations

International Nuclear Information System (INIS)

Peek, J.M.

1979-11-01

Electron-transport calculations rely on a large collection of electron-atom and photon-atom cross-section data to represent the response characteristics of the target medium. These basic atomic-physics quantities, and certain qualities derived from them that are now commonly in use, are critically reviewed. Publications appearing after 1978 are not given consideration. Processes involving electron or photon energies less than 1 keV are ignored, while an attempt is made to exhaustively cover the remaining independent parameters and target possibilities. Cases for which data improvements can be made from existing information are identified. Ranges of parameters for which state-of-the-art data are not available are sought out, and recommendations for explicit measurements and/or calculations with presently available tools are presented. An attempt is made to identify the maturity of the atomic-physics data and to predict the possibilities for rapid changes in the quality of the data. Finally, weaknesses in the state-of-the-art atomic-physics data and in the conceptual usage of these data in the context of electron-transport theory are discussed. Brief attempts are made to weight the various aspects of these questions and to suggest possible remedies

The shielding calculation for the CN guide shielding assembly in HANARO

International Nuclear Information System (INIS)

Kim, H. S.; Lee, B. C.; Lee, K. H.; Kim, H.

2006-01-01

The cold neutron research facility in HANARO is under construction. The area including neutron guides and rotary shutter in the reactor hall should be shielded by the guide shielding assembly which is constructed of heavy concrete blocks and structure. The guide shielding assembly is divided into 2 parts, A and B. Part A is about 6.4 meters apart from the reactor biological shield and it is constructed of heavy concrete blocks whose density is above 4.0g/cm 3 . And part B is a fixed heavy concrete structure whose density is above 3.5g/cm 3 . The rotary shutter is also made with heavy concrete whose density is above 4.0g/cm 3 and includes 5 neutron guides inside. It can block the neutron beam by rotating when CNS is not operating. The dose criterion outside the guide shielding assembly is established as 12.5 μSv/hr which is also applied to reactor shielding in HANARO

49 CFR 570.10 - Wheel assemblies.

Science.gov (United States)

2010-10-01

... bead through one full wheel revolution and note runout in excess of one-eighth of an inch. (c) Mounting... 49 Transportation 6 2010-10-01 2010-10-01 false Wheel assemblies. 570.10 Section 570.10... Pounds or Less § 570.10 Wheel assemblies. (a) Wheel integrity. A tire rim, wheel disc, or spider shall...

Benchmark assemblies of the Los Alamos critical assemblies facility

International Nuclear Information System (INIS)

Dowdy, E.J.

1986-01-01

Several critical assemblies of precisely known materials composition and easily calculated and reproducible geometries have been constructed at the Los Alamos National Laboratory. Some of these machines, notably Jezebel, Flattop, Big Ten, and Godiva, have been used as benchmark assemblies for the comparison of the results of experimental measurements and computation of certain nuclear reaction parameters. These experiments are used to validate both the input nuclear data and the computational methods. The machines and the applications of these machines for integral nuclear data checks are described. (author)

Benchmark assemblies of the Los Alamos Critical Assemblies Facility

International Nuclear Information System (INIS)

Dowdy, E.J.

1985-01-01

Several critical assemblies of precisely known materials composition and easily calculated and reproducible geometries have been constructed at the Los Alamos National Laboratory. Some of these machines, notably Jezebel, Flattop, Big Ten, and Godiva, have been used as benchmark assemblies for the comparison of the results of experimental measurements and computation of certain nuclear reaction parameters. These experiments are used to validate both the input nuclear data and the computational methods. The machines and the applications of these machines for integral nuclear data checks are described

Benchmark assemblies of the Los Alamos critical assemblies facility

International Nuclear Information System (INIS)

Dowdy, E.J.

1985-01-01

Several critical assemblies of precisely known materials composition and easily calculated and reproducible geometries have been constructed at the Los Alamos National Laboratory. Some of these machines, notably Jezebel, Flattop, Big Ten, and Godiva, have been used as benchmark assemblies for the comparison of the results of experimental measurements and computation of certain nuclear reaction parameters. These experiments are used to validate both the input nuclear data and the computational methods. The machines and the applications of these machines for integral nuclear data checks are described

Preliminary neutronics calculation of fusion-fission hybrid reactor breeding spent fuel assembly

International Nuclear Information System (INIS)

Ma Xubo; Chen Yixue; Gao Bin

2013-01-01

The possibility of using the fusion-fission hybrid reactor breeding spent fuel in PWR was preliminarily studied in this paper. According to the fusion-fission hybrid reactor breeding spent fuel characteristics, PWR assembly including fusion-fission hybrid reactor breeding spent fuel was designed. The parameters such as fuel temperature coefficient, moderator temperature coefficient and their variation were investigated. Results show that the neutron properties of uranium-based assembly and hybrid reactor breeding spent fuel assembly are similar. The design of this paper has a smaller uniformity coefficient of power at the same fissile isotope mass percentage. The results will provide technical support for the future fusion-fission hybrid reactor and PWR combined with cycle system. (authors)

A Preliminary Design Study of Ultra-Long-Life SFR Cores having Heterogeneous Fuel Assemblies

Energy Technology Data Exchange (ETDEWEB)

Jung, GeonHee; You, WuSeung; Hong, Ser Gi [Kyung Hee University, Yongin (Korea, Republic of)

2016-10-15

The PWR and CANDU reactors have provided electricity for several decades in our country but they have produced lots of spent fuels and so the safe and efficient disposal of these spent fuels is one of the main issues in nuclear industry. This type ultra-long-life cores are quite efficient in terms of the amount of spent fuel generation per electricity production and they can be used as an interim storage for PWR or CANDU spent fuel over several tens of years if they use the PWR or CANDU spent fuel as the initial fuel. Typically, the previous works have considered radially homogeneous fuel assemblies in which only blanket or driver fuel rods are employed and they considered axially or radially heterogeneous core configurations with the radially homogeneous fuel assemblies. These core configurations result in the propagation of the power distribution which can lead to the significant temperature changes for each fuel assembly over the time. In this work, the radially heterogeneous fuel assemblies are employed in new ultra-long-life SFR (Sodium-cooled Fast Reactor) cores to minimize the propagation of power distribution by allowing the power propagation in the fuel assemblies. In this work, new small ultra-long life SFR cores were designed with heterogeneous fuel assemblies having both blanket and driver fuel rods to minimize the propagation of power distribution over the core by allowing power propagation from driver rods to blanket rods in fuel assemblies. In particular, high fidelity depletion calculation coupled with heterogeneous Monte Carlo neutron transport calculation was performed to assess the neutronic feasibility of the ultralong life cores. The results of the analysis showed that the candidate core has the cycle length of 77 EFPYs, a small burnup reactivity swing of 1590 pcm and acceptably small SVRs both at BOC and EOC.

A drying system for spent fuel assemblies

International Nuclear Information System (INIS)

Suikki, M.; Warinowski, M.; Nieminen, J.

2007-06-01

The report presents a proposed drying apparatus for spent fuel assemblies. The apparatus is used for removing the moisture left in fuel assemblies during intermediate storage and transport. The apparatus shall be installed in connection with the fuel handling cell of an encapsulation plant. The report presents basic requirements for and implementation of the drying system, calculation of the drying process, operation, service and maintenance of the equipment, as well as a cost estimate. Some aspects of the apparatus design are quite specified, but the actual detailed planning and final selection of components have not been included. The report also describes actions for possible malfunction and fault conditions. An objective of the drying system for fuel assemblies is to remove moisture from the assemblies prior to placing the same in a disposal canister for spent nuclear fuel. Drying is performed as a vacuum drying process for vaporizing and draining the moisture present on the surface of the assemblies. The apparatus comprises two pieces of drying equipment. One of the chambers is equipped to take up Lo1-2 fuel assemblies and the other OL1-2 fuel assemblies. The chambers have an internal space sufficient to accommodate also OL3 fuel assemblies, but this requires replacing the internal chamber structure for laying down the assemblies to be dried. The drying chambers can be closed with hatches facing the fuel handling cell. Water vapour pumped out of the chamber is collected in a controlled manner, first by condensing with a heat exchanger and further by freezing in a cold trap. For reasons of safety, the exhaust air of vacuum pumps is further delivered into the ventilation outlet duct of a controlled area. The adequate drying result is ascertained by a low final pressure of about 100 Pa, as well as by a sufficient holding time. The chamber is built for making its cleaning as easy as possible in the event of a fuel rod breaking during a drying, loading or unloading

ZZ AIRFEWG, Gamma, Neutron Transport Calculation in Air Using FEWG1 Cross-Section

International Nuclear Information System (INIS)

1985-01-01

1 - Description of program or function: Format: ANISN; Number of groups: 37 neutron / 21 gamma-ray; Nuclides: air (79% N and 21% O); Origin: DLC-0031/FEWG1 cross sections (ENDF/B-IV). Weighting spectrum: 1/E. The AIRFEWG library has been generated by an ANISN multigroup calculation of gamma-ray, neutron, and secondary gamma-ray transport in infinite homogeneous air using DLC-0031/FEWG1 cross sections. 2 - Method of solution: The results were generated with a P3, ANISN run with a source in a single energy group. Thus, 58 such runs were required. For sources in the 37 neutron groups, both neutron and secondary gamma-ray fluence results were calculated. For gamma-ray sources only gamma-ray fluences were calculated

NEPTUNE: a modular scheme for the calculation of light water reactors

International Nuclear Information System (INIS)

Kavenoky, A.

1975-01-01

The NEPTUNE modular scheme has been developed to provide the physicist and the design engineer with a single system of codes for the calculation of light water reactors. The APOLLO code is included in NEPTUNE for the multigroup transport treatment of cells, groups of cells and complete fuel assemblies; few groups cross section libraries are automatically transmitted to the reactor multidimensional diffusion modules. In the reactor phase, 1D and 2D diffusion calculations can be performed by use of the finite difference method; 2D and 3D calculations are done respectively by the BILAN and TRIDENT modules using the finite element method. For the depletion calculation coarse and refined computations are offered. NEPTUNE is characterized by two special features for the data processing: the OTOMAT system which provides a virtual memory simulation and the intervention Monitor which allow to disconnect the computation modules and the control modules [fr

The nodal discrete-ordinate transport calculation of anisotropy scattering problem in three-dimensional cartesian geometry

International Nuclear Information System (INIS)

Wu Hongchun; Xie Zhongsheng; Zhu Xuehua

1994-01-01

The nodal discrete-ordinate transport calculating model of anisotropy scattering problem in three-dimensional cartesian geometry is given. The computing code NOTRAN/3D has been encoded and the satisfied conclusion is gained

Nuclear safety analysis for transport cask TK-6 (for WWER-440) and cover for fresh assemblies (for WWER-1000) in implementation of new fuel types at Ukrainian NPP

International Nuclear Information System (INIS)

Bilodid, Y.; Kovbasenko, Iu; Dudka, Olena

2006-01-01

According to the fresh fuel management procedure, fuel assemblies - after nuclear fuel delivery to the NPP fresh fuel unit - are vertically loaded into a cover intended for the delivery of fuel assemblies into the containment of the NPP reactor compartment. The cover is placed into an universal jack in the cooling and refueling pond, and then the fresh fuel assemblies are loaded into the reactor core. Based on the nuclear safety analysis carried out by the Russian Research Center 'Kurchatov Institute' for contemporary WWER-1000 fuel, it has become necessary to limit the number of fuel assemblies loaded into a cover below its designed capacity (12 FA instead of 18 FA as originally designed). Such a decision leads to worse economic performances in fuel transportation. The paper considers potential ways to overcome this restriction. Transport container TK-6 for spent fuel assemblies was designed quite a long time ago and, as shown in this paper, the requirement on the maximally permissible neutron multiplication factor of the loaded container for individual states to be analyzed in compliance with Ukrainian regulations is not met. First of all, this concerns the container criticality analysis in optimal neutron slow-down (container filling with water-air mixture with optimal density). The paper shows potential ways for TK-6 burnup-credit loading with the maximum number of fuel assemblies and partial container loading (Authors)

Benchmark for the qualification of gamma shielding calculation methods for light-water type reactor spent fuels

International Nuclear Information System (INIS)

Blum, P.; Cagnon, R.; Nimal, J.C.

1982-01-01

This report gives the results of a campaign of gamma dose rates measurement in the vicinity of a transport package loaded with 12 PWR spent fuel assemblies, so that the characteristics of the package and the fuel. It describes the measuring methods, and gives the accuracy of the data which will be usefull, as benchmarks, to the control of the calculation methods used to verify the gamma shielding of the packages. It shows how to calculate gamma dose rates from the data given on the package and the fuel, and gives the results of a calculation with the Mecure IV code and compares them to the measurements

Assessment of TRAC-PF1/MOD3 Mark-22 assembly model using SRL ''A'' tank single-assembly flow experiments

International Nuclear Information System (INIS)

Fischer, S.R.; Lam, K.; Lin, J.C.

1991-01-01

This paper summarizes the results of an assessment of our TRAC-PF1/MOD3 Mark-22 prototype fuel assembly model against single-assembly data obtained from the ''A'' Tank single-assembly tests that were performed at the Savannah River Laboratory. We felt the data characterize prototypic assembly behavior over a range of air-water flow conditions of interest for loss-of-coolant accident (LOCA) calculations. This study was part of a benchmarking effort performed to evaluate and validate a multiple-assembly, full-plant model that is being developed by Los Alamos National Laboratory to study various aspects of the Savannah River plant operating conditions, including LOCA transients, using TRAC-PF1/MOD3 Version 1.10. The results of this benchmarking effort demonstrate that TRAC-PF1/MOD3 is capable pf calculating plenum conditions and assembly flows during conditions thought to be typical of the Emergency Cooling System (ECS) phase of a LOCA. 10 refs., 12 fig

Vectorization and parallelization of Monte-Carlo programs for calculation of radiation transport

International Nuclear Information System (INIS)

Seidel, R.

1995-01-01

The versatile MCNP-3B Monte-Carlo code written in FORTRAN77, for simulation of the radiation transport of neutral particles, has been subjected to vectorization and parallelization of essential parts, without touching its versatility. Vectorization is not dependent on a specific computer. Several sample tasks have been selected in order to test the vectorized MCNP-3B code in comparison to the scalar MNCP-3B code. The samples are a representative example of the 3-D calculations to be performed for simulation of radiation transport in neutron and reactor physics. (1) 4πneutron detector. (2) High-energy calorimeter. (3) PROTEUS benchmark (conversion rates and neutron multiplication factors for the HCLWR (High Conversion Light Water Reactor)). (orig./HP) [de

Calculation of reactivity of control rods in graphite moderated reactors

International Nuclear Information System (INIS)

Nakata, H.

1978-01-01

A study about the method of calculation for the reactivity of control rods in graphite-moderated critical assemblies, is presented. The result of theoretical calculation, developed by super celles and Nordheim-Scalettar methods are compared with experimental results for the critical Assembly of General Atomic. The two methods are then applicable to reactivity calculation of the control rods of graphite moderated critical assemblies [pt

Reactivity Impact of Difference of Nuclear Data Library for PWR Fuel Assembly Calculation by Using AEGIS Code

International Nuclear Information System (INIS)

Ohoka, Yasunori; Tatsumi, Masahiro; Sugimura, Naoki; Tabuchi, Masato

2011-01-01

In 2010, the latest version of the Japanese Evaluated Nuclear Data Library (JENDL-4.0) has been released by JAEA. JENDL-4.0 is major update from JENDL- 3.3, and confirmed to give good accuracy by integral test for fission reactor systems such as fast neutron system and thermal neutron system. In this study, we evaluated the reactivity impact due to difference between ENDF/B-VII.0 and JENDL-4.0 for PWR fuel assembly burnup calculation using AEGIS code which has been developed by Nuclear Engineering, Ltd. in cooperation with Nuclear Fuel Industries, Ltd. and Nagoya University

Design analysis of various transportation package options for BN-350 SNF in terms of nuclear radiation safety in planning for long-terms dry storage

International Nuclear Information System (INIS)

Aisabekov, A.Z.; Mukenova, S.A.; Tur, E.S.; Tsyngaev, V.M.

2004-01-01

Full text: This effort is performed under the BN-350 reactor facility decommissioning project. One of the project tasks - spent nuclear fuel handling - includes the following: fuel packaging into sealed canisters, transportation of the canisters in multi-seat metallo-concrete containers and placement of the containers for a long-term dry storage. The goal of this effort is to computationally validate nuclear and radiation safety of the SNF containers placed for storage both under normal storage conditions and probable accident situations. The basic unit structure and design configurations are presented: assemblies, canisters, transportation containers. The major factors influencing nuclear and radiation safety are presented: fuel burn-up, enrichment, fabrication tolerance, types of fuel assemblies, configuration of assemblies in the canister and canisters in the container, background of assemblies placed in the reactor and cooling pool. Conditions under which the SNF containers will be stored are described and probable accident situations are listed. Proceeding from the conservatism principle, selection of the assemblies posing the greatest nuclear hazard is validated. A neutron effective multiplication factor is calculated for the SNF containers under the normal storage conditions and for the case of emergency. The effective multiplication factor is shown to be within a standard value of 0.95 in any situation. Based on the experimental data on assembly and canister dose rates, canisters posing the highest radiation threat are selected. Activities of sources and gamma-radiation spectral composition are calculated. Distribution of the dose rate outside the containers both under the normal storage conditions and accident situations are calculated. The results obtained are analyzed

Refuelling design and core calculations at NPP Paks: codes and methods

International Nuclear Information System (INIS)

Pos, I.; Nemes, I.; Javor, E.; Korpas, L.; Szecsenyi, Z.; Patai-Szabo, S.

2001-01-01

This article gives a brief review of the computer codes used in the fuel management practice at NPP Paks. The code package consist of the HELIOS neutron and gamma transport code for preparation of few-group cross section library, the CERBER code to determine the optimal core loading patterns and the C-PORCA code for detailed reactor physical analysis of different reactor states. The last two programs have been developed at the NPP Paks. HELIOS gives sturdy basis for our neutron physical calculation, CERBER and C-PORCA programs have been enhanced in great extent for last years. Methods and models have become more detailed and accurate as regards the calculated parameters and space resolution. Introduction of a more advanced data handling algorithm arbitrary move of fuel assemblies can be followed either in the reactor core or storage pool. The new interactive WINDOWS applications allow easier and more reliable use of codes. All these computer code developments made possible to handle and calculate new kind of fuels as profiled Russian and BNFL fuel with burnable poison or to support the reliable reuse of fuel assemblies stored in the storage pool. To extend thermo-hydraulic capability, with KFKI contribution the COBRA code will also be coupled to the system (Authors)

Integral transport computation of gamma detector response with the CPM2 code

International Nuclear Information System (INIS)

Jones, D.B.

1989-12-01

CPM-2 Version 3 is an enhanced version of the CPM-2 lattice physics computer code which supports the capabilities to (1) perform a two-dimensional gamma flux calculation and (2) perform Restart/Data file maintenance operations. The Gamma Calculation Module implemented in CPM-2 was first developed for EPRI in the CASMO-1 computer code by Studsvik Energiteknik under EPRI Agreement RP2352-01. The gamma transport calculation uses the CPM-HET code module to calculate the transport of gamma rays in two dimensions in a mixed cylindrical-rectangular geometry, where the basic fuel assembly and component regions are maintained in a rectangular geometry, but the fuel pins are represented as cylinders within a square pin cell mesh. Such a capability is needed to represent gamma transport in an essentially transparent medium containing spatially distributed ''black'' cylindrical pins. Under a subcontract to RP2352-01, RPI developed the gamma production and gamma interaction library used for gamma calculation. The CPM-2 gamma calculation was verified against reference results generated by Studsvik using the CASMO-1 program. The CPM-2 Restart/Data file maintenance capabilities provide the user with options to copy files between Restart/Data tapes and to purge files from the Restart/Data tapes

Development of a database system for the calculation of indicators of environmental pressure caused by transport

Energy Technology Data Exchange (ETDEWEB)

Giannouli, Myrsini; Samaras, Zissis [Aristotle University of Thessaloniki, Laboratory of Applied Thermodynamics, Mechanical Engineering Department, GR 54124, Thessaloniki, P.O. Box 458 (Greece); Keller, Mario; De Haan, Peter [INFRAS, Muhlemattstrasse 45 CH-3007, Bern (Switzerland); Kallivoda, Manfred [psiA-Consult, Environmental Research and Engineering GmbH, Lastenstrasse 38/1, 1230 Wien (Austria); Sorenson, Spencer; Georgakaki, Aliki [DTU: Technical University of Denmark, Nils Koppels Alle, Building 403, DK 2800 Kgs. Lyngby (Denmark)

2006-03-15

The scope of this paper is to summarise a methodology developed for TRENDS (TRansport and ENvironment Database System-TRENDS). The main objective of TRENDS was the calculation of environmental pressure indicators caused by transport. The environmental pressures considered are associated with air emissions from the four main transport modes, i.e. road, rail, ships and air. In order to determine these indicators a system for calculating a range of environmental pressures due to transport was developed within a PC-based MS Access environment. Emphasis is given on the latest features incorporated in the model and their applications. One of the recently developed features of the software provides an option for simple scenario analysis including vehicle dynamics (such as turnover and evolution) for all EU15 member states. This feature is called the Transport Activity Balance module (TAB) and enables the production of collective results for all transport modes as well as a comparative assessment of air emissions produced by the various modes. Traffic activity and emission data obtained according to a basic (reference) scenario are displayed for the time period 1970-2020. In addition, a detailed assessment of the results produced by TRENDS was conducted by means of comparison with data found in the literature. Finally, vehicle emissions produced by the model for the EU15 member states were spatially disaggregated for the base year, 1995 and GIS maps were generated. Examples of these maps are displayed in this document, for the various modes of transport considered in the study. (author)

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

International Nuclear Information System (INIS)

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

2001-01-01

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

Investigation of the structural anisotropy in a self-assembling glycinate layer on Cu(100) by scanning tunneling microscopy and density functional theory calculations

Science.gov (United States)

Kuzmin, Mikhail; Lahtonen, Kimmo; Vuori, Leena; Sánchez-de-Armas, Rocío; Hirsimäki, Mika; Valden, Mika

2017-07-01

Self-assembling organic molecule-metal interfaces exhibiting free-electron like (FEL) states offers an attractive bottom-up approach to fabricating materials for molecular electronics. Accomplishing this, however, requires detailed understanding of the fundamental driving mechanisms behind the self-assembly process. For instance, it is still unresolved as to why the adsorption of glycine ([NH2(CH2)COOH]) on isotropic Cu(100) single crystal surface leads, via deprotonation and self-assembly, to a glycinate ([NH2(CH2)COO-]) layer that exhibits anisotropic FEL behavior. Here, we report on bias-dependent scanning tunneling microscopy (STM) experiments and density functional theory (DFT) calculations for glycine adsorption on Cu(100) single crystal surface. We find that after physical vapor deposition (PVD) of glycine on Cu(100), glycinate self-assembles into an overlayer exhibiting c(2 × 4) and p(2 × 4) symmetries with non-identical adsorption sites. Our findings underscore the intricacy of electrical conductivity in nanomolecular organic overlayers and the critical role the structural anisotropy at molecule-metal interface plays in the fabrication of materials for molecular electronics.

Radionuclide transport calculations from high-level long-lived radioactive waste disposal in deep clayey geologic formation toward adjacent aquifers

International Nuclear Information System (INIS)

Genty, A.; Le Potier, C.

2007-01-01

In the context of high-level nuclear waste repository safety calculations, the modeling of radionuclide migration is of first importance. Three dimensional radionuclide transport calculations in geological repository need to describe objects of the meter scale embedded in geologic layer formations of kilometer extension. A complete and refined spatial description would end up with at least meshes of hundreds of millions to billions elements. The resolution of this kind of problem is today not reachable with classical computers due to resources limitations. Although parallelized computation appears as potential tool to handle multi-scale calculations, to our knowledge no attempt have been yet performed. One emerging solution for repository safety calculations on very large cells meshes consists in using a domain decomposition approach linked to massive parallelized computer calculation. In this approach, the repository domain is divided in small elementary domains and transport calculation are performed independently on different processor for each elementary domain. Before to develop this possible solution, we performed some preliminary test in order to access the order of magnitude of cells needed to perform converged calculation on one elementary disposal domain and to check if Finite Volume (FV) based on Multi Point Flux Approximation (MPFA) spatial scheme or more classical Mixed Hybrid Finite Element (MHFE) spatial scheme were adapted for those calculations in highly heterogeneous porous media. Our preliminary results point out that MHFE and VF schemes applied on non-parallelepiped hexahedral cells for flow and transport calculations in highly heterogeneous media gave satisfactory results. Nevertheless further investigations and additional calculations are needed in order to exhibit the mesh discretization level needed to perform converged calculations. (authors)

3D heterogeneous transport calculations of CANDU fuel with EVENT/HELIOS

International Nuclear Information System (INIS)

Rahnema, F.; Mosher, S.; Ilas, D.; De Oliveira, C.; Eaton, M.; Stamm'ler, R.

2002-01-01

The applicability of the EVENT/HELIOS package to CANDU lattice cell analysis is studied in this paper. A 45-group cross section library is generated using the lattice depletion transport code HELIOS. This library is then used with the 3-D transport code EVENT to compute the pin fission densities and the multiplication constants for six configurations typical of a CANDU cell. The results are compared to those from MCNP with the same multigroup library. Differences of 70-150 pcm in multiplication constant and 0.08-0.95% in pin fission density are found for these cases. It is expected that refining the EVENT calculations can reduce these differences. This gives confidence in applying EVENT to transient analyses at the fuel pin level in a selected part of a CANDU core such as the limiting bundle during a loss of coolant accident (LOCA). (author)

Radiation transport calculations for the ANS [Advanced Neutron Source] beam tubes

International Nuclear Information System (INIS)

Engle, W.W. Jr.; Lillie, R.A.; Slater, C.O.

1988-01-01

The Advanced Neutron Source facility (ANS) will incorporate a large number of both radial and no-line-of-sight (NLS) beam tubes to provide very large thermal neutron fluxes to experimental facilities. The purpose of this work was to obtain comparisons for the ANS single- and split-core designs of the thermal and damage neutron and gamma-ray scalar fluxes in these beams tubes. For experimental locations far from the reactor cores, angular flux data are required; however, for close-in experimental locations, the scalar fluxes within each beam tube provide a credible estimate of the various signal to noise ratios. In this paper, the coupled two- and three-dimensional radiation transport calculations employed to estimate the scalar neutron and gamma-ray fluxes will be described and the results from these calculations will be discussed. 6 refs., 2 figs

Comparison of Non-overlapping and Overlapping Local/Global Iteration Schemes for Whole-Core Deterministic Transport Calculation

International Nuclear Information System (INIS)

Yuk, Seung Su; Cho, Bumhee; Cho, Nam Zin

2013-01-01

In the case of deterministic transport model, fixed-k problem formulation is necessary and the overlapping local domain is chosen. However, as mentioned in, the partial current-based Coarse Mesh Finite Difference (p-CMFD) procedure enables also non-overlapping local/global (NLG) iteration. In this paper, NLG iteration is combined with p-CMFD and with CMFD (augmented with a concept of p-CMFD), respectively, and compared to OLG iteration on a 2-D test problem. Non-overlapping local/global iteration with p-CMFD and CMFD global calculation is introduced and tested on a 2-D deterministic transport problem. The modified C5G7 problem is analyzed with both NLG and OLG methods and the solutions converge to the reference solution except for some cases of NLG with CMFD. NLG with CMFD gives the best performance if the solution converges. But if fission-source iteration in local calculation is not enough, it is prone to diverge. The p-CMFD global solver gives unconditional convergence (for both OLG and NLG). A study of switching scheme is in progress, where NLG/p-CMFD is used as 'starter' and then switched to NLG/CMFD to render the whole-core transport calculation more efficient and robust. Parallel computation is another obvious future work

Monte Carlo method for neutron transport problems

International Nuclear Information System (INIS)

Asaoka, Takumi

1977-01-01

Some methods for decreasing variances in Monte Carlo neutron transport calculations are presented together with the results of sample calculations. A general purpose neutron transport Monte Carlo code ''MORSE'' was used for the purpose. The first method discussed in this report is the method of statistical estimation. As an example of this method, the application of the coarse-mesh rebalance acceleration method to the criticality calculation of a cylindrical fast reactor is presented. Effective multiplication factor and its standard deviation are presented as a function of the number of histories and comparisons are made between the coarse-mesh rebalance method and the standard method. Five-group neutron fluxes at core center are also compared with the result of S4 calculation. The second method is the method of correlated sampling. This method was applied to the perturbation calculation of control rod worths in a fast critical assembly (FCA-V-3) Two methods of sampling (similar flight paths and identical flight paths) are tested and compared with experimental results. For every cases the experimental value lies within the standard deviation of the Monte Carlo calculations. The third method is the importance sampling. In this report a biased selection of particle flight directions discussed. This method was applied to the flux calculation in a spherical fast neutron system surrounded by a 10.16 cm iron reflector. Result-direction biasing, path-length stretching, and no biasing are compared with S8 calculation. (Aoki, K.)

On calculating phase shifts and performing fits to scattering cross sections or transport properties

International Nuclear Information System (INIS)

Hepburn, J.W.; Roy, R.J. Le

1978-01-01

Improved methods of calculating quantum mechanical phase shifts and for performing least-squares fits to scattering cross sections or transport properties, are described. Their use in a five-parameter fit to experimental differential cross sections reduces the computer time by a factor of 4-7. (Auth.)

Generalized Coarse-Mesh Rebalance Method for Acceleration of Neutron Transport Calculations

International Nuclear Information System (INIS)

Yamamoto, Akio

2005-01-01

This paper proposes a new acceleration method for neutron transport calculations: the generalized coarse-mesh rebalance (GCMR) method. The GCMR method is a unified scheme of the traditional coarse-mesh rebalance (CMR) and the coarse-mesh finite difference (CMFD) acceleration methods. Namely, by using an appropriate acceleration factor, formulation of the GCMR method becomes identical to that of the CMR or CMFD method. This also indicates that the convergence property of the GCMR method can be controlled by the acceleration factor since the convergence properties of the CMR and CMFD methods are generally different. In order to evaluate the convergence property of the GCMR method, a linearized Fourier analysis was carried out for a one-group homogeneous medium, and the results clarified the relationship between the acceleration factor and the spectral radius. It was also shown that the spectral radius of the GCMR method is smaller than those of the CMR and CMFD methods. Furthermore, the Fourier analysis showed that when an appropriate acceleration factor was used, the spectral radius of the GCMR method did not exceed unity in this study, which was in contrast to the results of the CMR or the CMFD method. Application of the GCMR method to practical calculations will be easy when the CMFD acceleration is already adopted in a transport code. By multiplying a suitable acceleration factor to a coefficient (D FD ) of a finite difference formulation, one can improve the numerical instability of the CMFD acceleration method

Monte Carlo modeling and analyses of YALINA- booster subcritical assembly Part II: pulsed neutron source

International Nuclear Information System (INIS)

Talamo, A.; Gohar, M.Y.A.; Rabiti, C.

2008-01-01

One of the most reliable experimental methods for measuring the kinetic parameters of a subcritical assembly is the Sjoestrand method applied to the reaction rate generated from a pulsed neutron source. This study developed a new analytical methodology for characterizing the kinetic parameters of a subcritical assembly using the Sjoestrand method, which allows comparing the analytical and experimental time dependent reaction rates and the reactivity measurements. In this methodology, the reaction rate, detector response, is calculated due to a single neutron pulse using MCNP/MCNPX computer code or any other neutron transport code that explicitly simulates the fission delayed neutrons. The calculation simulates a single neutron pulse over a long time period until the delayed neutron contribution to the reaction is vanished. The obtained reaction rate is superimposed to itself, with respect to the time, to simulate the repeated pulse operation until the asymptotic level of the reaction rate, set by the delayed neutrons, is achieved. The superimposition of the pulse to itself was calculated by a simple C computer program. A parallel version of the C program is used due to the large amount of data being processed, e.g. by the Message Passing Interface (MPI). The new calculation methodology has shown an excellent agreement with the experimental results available from the YALINA-Booster facility of Belarus. The facility has been driven by a Deuterium-Deuterium or Deuterium-Tritium pulsed neutron source and the (n,p) reaction rate has been experimentally measured by a 3 He detector. The MCNP calculation has utilized the weight window and delayed neutron biasing variance reduction techniques since the detector volume is small compared to the assembly volume. Finally, this methodology was used to calculate the IAEA benchmark of the YALINA-Booster experiment

Experimental validation of radial reconstructed pin-power distributions in full-scale BWR fuel assemblies with and without control blade

Energy Technology Data Exchange (ETDEWEB)

Giust, Flavio, E-mail: flavio.giust@axpo.c [Paul Scherrer Institute, CH-5232 Villigen PSI (Switzerland); Ecole Polytechnique Federale de Lausanne, CH-1015 Lausanne (Switzerland); Axpo Kernenergie AG, Parkstrasse 23, CH-5401 Baden (Switzerland); Grimm, Peter [Paul Scherrer Institute, CH-5232 Villigen PSI (Switzerland); Chawla, Rakesh [Paul Scherrer Institute, CH-5232 Villigen PSI (Switzerland); Ecole Polytechnique Federale de Lausanne, CH-1015 Lausanne (Switzerland)

2010-12-15

Total fission rate measurements have been performed on full-size BWR fuel assemblies of type SVEA-96+ in the zero power reactor PROTEUS at the Paul Scherrer Institute. This paper presents comparisons of reconstructed 2D pin fission rates from nodal diffusion calculations to the experimental results in two configurations: one 'regular' (I-1A) and the other 'controlled' (I-2A). Both configurations consist of an array of 3 x 3 SVEA-96+ fuel assemblies moderated with light water at 20 {sup o}C. In configuration I-2A, an L-shaped hafnium control blade (half of a real cruciform blade) is inserted adjacent to the north-west corner of the central fuel assembly. To minimise the impact of the surroundings, all measurements were done in fuel pins belonging to the central assembly. The 3 x 3 experimental configuration (test zone) was modelled using the core monitoring and design tools that are applied at the Leibstadt Nuclear Power Plant (KKL). These are the 2D transport code HELIOS, used for the cross-section generation, and the 3D, 2-group nodal diffusion code PRESTO-2. The exterior is represented, in the axial and radial directions, by 2-group partial current ratios (PCRs) calculated at the test zone boundary using a 3D Monte Carlo (MCNPX) model of the whole PROTEUS reactor. Sensitivity cases are analysed to show the impact of changes in the 2D lattice modelling on the calculated fission rate distribution and reactivity. Further, the effects of variations in the test zone boundary PCRs and their behaviour in energy are investigated. For the test zone configuration without control blade, the pin-power reconstruction methodology delivers the same level of accuracy as the 2D transport calculations. On the other hand, larger deviations that are inherent to the use of reflected geometry in the lattice calculations are observed for the configuration with the control blade inserted. In the basic (reference) simulation cases, the calculated-to-experimental (C

Architecture and roles of periplasmic adaptor proteins in tripartite efflux assemblies.

Directory of Open Access Journals (Sweden)

Vassiliy N. Bavro

2015-05-01

Full Text Available Recent years have seen major advances in the structural understanding of the different components of tripartite efflux assemblies, which encompass the multidrug efflux (MDR pumps and type I secretion systems. The majority of these investigations have focused on the role played by the inner membrane transporters and the outer membrane factor (OMF, leaving the third component of the system – the Periplasmic Adaptor Proteins (PAPs - relatively understudied. Here we review the current state of knowledge of these versatile proteins which, far from being passive linkers between the OMF and the transporter, emerge as active architects of tripartite assemblies, and play diverse roles in the transport process. Recognition between the PAPs and OMFs is essential for pump assembly and function, and targeting this interaction may provide a novel avenue for combating multidrug resistance. With the recent advances elucidating the drug-efflux and energetics of the tripartite assemblies, the understanding of the interaction between the OMFs and PAPs is the last piece remaining in the complete structure of the tripartite pump assembly puzzle.

Linear-chain assemblies of iron oxide nanoparticles

Energy Technology Data Exchange (ETDEWEB)

Dhak, Prasanta; Kim, Min-Kwan; Lee, Jae Hyeok; Kim, Miyoung; Kim, Sang-Koog, E-mail: sangkoog@snu.ac.kr

2017-07-01

Highlights: • Hydrothermal synthesis of pure phase 200 nm Fe{sub 3}O{sub 4} nanoparticles. • Studies of linear-chain assemblies of iron oxide nanosphere by FESEM. • Micromagnetic simulations showed the presence of 3D vortex states. • The B.E. for different numbers of particles in linear chain assemblies were calculated. - Abstract: We synthesized iron oxide nanoparticles using a simple hydrothermal approach and found several types of segments of their linear-chain self-assemblies as observed by field emission scanning electron microscopy. X-ray diffraction and transmission electron microscopy measurements confirm a well-defined single-phase FCC structure. Vibrating sample magnetometry measurements exhibit a ferromagnetic behavior. Micromagnetic numerical simulations show magnetic vortex states in the nanosphere model. Also, calculations of binding energies for different numbers of particles in the linear-chain assemblies explain a possible mechanism responsible for the self-assemblies of segments of the linear chains of nanoparticles. This work offers a step towards linear-chain self-assemblies of iron oxide nanoparticles and the effect of magnetic vortex states in individual nanoparticles on their binding energy.

Systematic assembly homogenization and local flux reconstruction for nodal method calculations of fast reactor power distributions

International Nuclear Information System (INIS)

Dorning, J.J.

1991-01-01

A simultaneous pin lattice cell and fuel bundle homogenization theory has been developed for use with nodal diffusion calculations of practical reactors. The theoretical development of the homogenization theory, which is based on multiple-scales asymptotic expansion methods carried out through fourth order in a small parameter, starts from the transport equation and systematically yields: a cell-homogenized bundled diffusion equation with self-consistent expressions for the cell-homogenized cross sections and diffusion tensor elements; and a bundle-homogenized global reactor diffusion equation with self-consistent expressions for the bundle-homogenized cross sections and diffusion tensor elements. The continuity of the angular flux at cell and bundle interfaces also systematically yields jump conditions for the scaler flux or so-called flux discontinuity factors on the cell and bundle interfaces in terms of the two adjacent cell or bundle eigenfunctions. The expressions required for the reconstruction of the angular flux or the 'de-homogenization' theory were obtained as an integral part of the development; hence the leading order transport theory angular flux is easily reconstructed throughout the reactor including the regions in the interior of the fuel bundles or computational nodes and in the interiors of the pin lattice cells. The theoretical development shows that the exact transport theory angular flux is obtained to first order from the whole-reactor nodal diffusion calculations, done using the homogenized nuclear data and discontinuity factors, is a product of three computed quantities: a ''cell shape function''; a ''bundle shape function''; and a ''global shape function''. 10 refs

EFTF cobalt test assembly results

International Nuclear Information System (INIS)

Rawlins, J.A.; Wootan, D.W.; Carter, L.L.; Brager, H.R.; Schenter, R.E.

1988-01-01

A cobalt test assembly containing yttrium hydride pins for neutron moderation was irradiated in the Fast Flux Test Facility during Cycle 9A for 137.7 equivalent full power days at a power level fo 291 MW. The 36 test pins consisted of a batch of 32 pins containing cobalt metal to produce Co-60, and a set of 4 pins with europium oxide to produce Gd-153, a radioisotope used in detection of the bone disease Osteoporosis. Post-irradiation examination of the cobalt pins determined the Co-60 produced with an accuracy of about 5 %. The measured Co-60 spatially distributed concentrations were within 20 % of the calculated concentrations. The assembly average Co-60 measured activity was 4 % less than the calculated value. The europium oxide pins were gamma scanned for the europium isotopes Eu-152 and Eu-154 to an absolute accuracy of about 10 %. The measured europium radioisotpe anc Gd-153 concentrations were within 20 % of calculated values. In conclusion, the hydride assembly performed well and is an excellent vehicle for many Fast Flux Test Facility isotope production applications. The results also demonstrate that the calculational methods developed by the Westinghouse Hanford Company are very accurate. (author)

Development of a dry transport and storage cask for spent LWR fuel assemblies in Spain

International Nuclear Information System (INIS)

Melches, C.; Uriarte, A.; Espallardo, J.A.

1982-01-01

One of the advantages of the cask storage concept is its flexibility which makes it specially attractive in the case of the Spanish circumstances. For these reasons the Empresa Nacional del Uranio, S.A. (ENUSA), Junta de Energia Nuclear (JEN) and Equipos Nucleares, S.A. (ENSA) initiated in 1981 a joint program for the development of a prototype cask for the dry transport and storage of spent fuel assemblies. This program includes as main steps the analysis of the conceptual design, the detailed design and experimental tests, the fabrication of a prototype and its licencing and safety testing. The mentioned program, which started in the early 1981, is scheduled to be completed at the end of 1984

An approximate framework for quantum transport calculation with model order reduction

Energy Technology Data Exchange (ETDEWEB)

Chen, Quan, E-mail: quanchen@eee.hku.hk [Department of Electrical and Electronic Engineering, The University of Hong Kong (Hong Kong); Li, Jun [Department of Chemistry, The University of Hong Kong (Hong Kong); Yam, Chiyung [Beijing Computational Science Research Center (China); Zhang, Yu [Department of Chemistry, The University of Hong Kong (Hong Kong); Wong, Ngai [Department of Electrical and Electronic Engineering, The University of Hong Kong (Hong Kong); Chen, Guanhua [Department of Chemistry, The University of Hong Kong (Hong Kong)