Showing papers in "Nuclear Science and Engineering in 2005"
TL;DR: In this paper, relative entropy has been applied to a slightly asymmetric two-fissile-component problem with a dominance ratio (DR) of 0.993 and the numerical results are mostly satisfactory but also show the possibility of the occasional occurrence of unnecessarily strict stationarity diagnostics.
Abstract: In Monte Carlo criticality calculations, source error propagation through the stationary (active) cycles and source convergence in the settling (inactive) cycles are both dominated by the dominance ratio (DR) of fission kernels. For symmetric two-fissile-component systems with the DR close to unity, the extinction of fission source sites can occur in one of the components even when the initial source is symmetric and the number of histories per cycle is more than 1000. When such a system is made slightly asymmetric, the neutron effective multiplication factor at the inactive cycles does not reflect the convergence to stationary source distribution. To overcome this problem, relative entropy has been applied to a slightly asymmetric two-fissile-component problem with a DR of 0.993. The numerical results are mostly satisfactory but also show the possibility of the occasional occurrence of unnecessarily strict stationarity diagnostics. Therefore, a criterion is defined based on the concept of data compression limit in information theory. Numerical results for a pressurized water reactor fuel storage facility with a DR of 0.994 strongly support the efficacy of relative entropy in both the posterior and progressive stationarity diagnostics.
72 citations
TL;DR: In this paper, two possibilities of mixing Pu and enriched U in the same assembly are presented (homogeneously and heterogeneously) and a variant of the CORAIL concept in which the MOX rods are substituted with inert matrix fuel rods (PuO{sub 2}-CeO(sub 2}) was also studied.
Abstract: If it becomes necessary to stabilize the Pu inventory before the advent of Gen IV fast reactors, then it must be multirecycled in thermal neutron reactors like pressurized water reactors (PWRs). However, because of the neutron physics characteristics of Pu, it is difficult to multirecycle it in mixed-oxide (MOX)-fueled PWRs. Indeed, since there are fewer and fewer fissile isotopes in Pu, it is necessary to compensate by increasing its content, causing it to quickly reach values where the void coefficient is positive (above 12% Pu). To avoid this, Pu must be used together with enriched U so that its degradation is compensated by an increase of {sup 235}U enrichment. Two possibilities of mixing Pu and enriched U in the same assembly are presented (homogeneously and heterogeneously). In the first, called MOX-UE, all the fuel rods are made of PuO{sub 2}-U{sub enriched}O{sub 2}, whereas the second, called CORAIL, contains approximately one-third of standard MOX rods (PuO{sub 2}-U{sub tail}O{sub 2}) and two-thirds of UO{sub 2} rods. A variant of the CORAIL concept in which the MOX rods are substituted with inert matrix fuel rods (PuO{sub 2}-CeO{sub 2}) was also studied. These assemblies allow Pu to be multirecycled in standard PWRs, thus stabilizingmore » the Pu inventory between 200 and 400 t heavy metal (for a nuclear electricity production of 400 TW.h(electric)/yr, i.e., typical of a country such as France). The number of reactors loaded with Pu depends on the performances of each concept in terms of Pu burning, and it represents between 80% (CORAIL with the MOX rods) and 30% (MOX-UE with 12% Pu) of the total power. There is only a small difference regarding the needs in natural U between the Pu monorecycling option and the different Pu multirecycling options. Hence, it appears that saving U should not be offered as an incentive for multirecycling Pu in PWRs.« less
53 citations
TL;DR: In this article, the generalized coarse-mesh rebalance (GCMR) method was proposed for neutron transport calculations, which is a unified scheme of the traditional coarse-and-means scheme.
Abstract: 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-...
49 citations
TL;DR: In this paper, the ability of point kinetics to describe dynamic processes in accelerator-driven systems (ADSs) is investigated and full three-dimensional energy-space-time-dependent calculations, coupled with thermodynamic properties, are presented.
Abstract: The ability of point kinetics to describe dynamic processes in accelerator-driven systems (ADSs) is investigated. Full three-dimensional energy-space-time-dependent calculations, coupled with therm ...
43 citations
TL;DR: The solution in this study was judged to be the use of Monte Carlo techniques coupled with robust variance reduction to accelerate problem convergence.
Abstract: Certain reactor transients cause a reduction in moderator temperature and, hence, increased attenuation of neutrons and decreased response of excore detectors. This decreased detector response is of concern because of the credit assumed for detector-initiated reactor trip to terminate the transient. Explicit modeling of this phenomenon presents the analyst with a difficult problem because of the dense and optically thick neutron absorption media, given the constraint that precise response characteristics must be known in order to account for this phenomenon. The solution in this study was judged to be the use of Monte Carlo techniques coupled with robust variance reduction to accelerate problem convergence. A fresh discussion on the motivation for variance reduction is included, followed by separate accounts of manual and automated applications of variance reduction techniques. Finally, the results of both manual and automated variance reduction techniques are presented and compared.
43 citations
TL;DR: This work presents an asynchronous message-passing algorithm that performs sweeps simultaneously in multiple ordinate directions, a simple geometric heuristic to prioritize the computational tasks that a processor works on, and an algorithm for detecting and eliminating cycles that sometimes exist in unstructured grids and can prevent sweeps from successfully completing.
Abstract: The method of discrete ordinates is commonly used to solve the Boltzmann transport equation. The solution in each ordinate direction is most efficiently computed by sweeping the radiation flux across the computational grid. For unstructured grids this poses many challenges, particularly when implemented on distributed-memory parallel machines where the grid geometry is spread across processors. We present several algorithms relevant to this approach: (a) an asynchronous message-passing algorithm that performs sweeps simultaneously in multiple ordinate directions, (b) a simple geometric heuristic to prioritize the computational tasks that a processor works on, (c) a partitioning algorithm that creates columnar-style decompositions for unstructured grids, and (d) an algorithm for detecting and eliminating cycles that sometimes exist in unstructured grids and can prevent sweeps from successfully completing. Algorithms (a) and (d) are fully parallel; algorithms (b) and (c) can be used in conjunction with (a) to achieve higher parallel efficiencies. We describe our message-passing implementations of these algorithms within a radiation transport package. Performance and scalability results are given for unstructured grids with up to 3 million elements (500 million unknowns) running on thousands of processors of Sandia National Laboratories' Intel Tflops machine and DEC-Alpha CPlant cluster.
39 citations
TL;DR: It is shown theoretically and then proved by numerical examples that the separation of the eigenvalues of the mathematical operator defining the problem can be taken as a good indicator of the importance of space effects in time-dependent conditions.
Abstract: The paper considers some physical aspects of the neutron space kinetics of critical and source-driven subcritical systems. The possibility of introducing some indicators to qualify the spatial natu...
38 citations
TL;DR: In this article, an oxide-dispersion-strengthened (ODS) martensitic steel 9Cr-ODS was irradiated with 5-MeV Ni ions at 500 C at a dose rate of 1.4? 10-3 dpa/s to doses of 5, 50, and 150 dpa.
Abstract: An oxide-dispersion-strengthened (ODS) martensitic steel 9Cr-ODS was irradiated with 5-MeV Ni ions at 500 C at a dose rate of 1.4 ? 10-3 dpa/s to doses of 5, 50, and 150 dpa. The ODS steel has been designed for use in higher-temperature energy systems. However, the radiation effects are not fully characterized, particularly to high doses. Dense dislocations, precipitates, and yttrium-titanium oxide particles dominated the microstructure of 9Cr-ODS for both the unirradiated and irradiated cases with no dislocation loops observed. No voids were detected for doses up to 150 dpa. The average size of the oxide particles, whose size is approximately described by a lognormal distribution, slightly decreased with dose from {approx}12 nm for the unirradiated case to {approx}9 nm at 150 dpa. The decrease in size follows a square root of dose dependency, indicating the effect is radiation induced. The decrease in size is not expected to have a detrimental effect on high-temperature strength, even to extremely high dose.
38 citations
TL;DR: The main focus is on accurately approximating the effects that neighboring assemblies have on the few-group cross sections, assembly discontinuity factors, form factors, and other transport parameters of a given assembly, by using albedo boundary conditions that are estimated with low computational cost.
Abstract: We present recent improvements in assembly-level calculations for reactor analysis, including modifications that support core-level analysis by quasi-diffusion. Our main focus is on accurately approximating the effects that neighboring assemblies have on the few-group cross sections, assembly discontinuity factors, form factors, and other transport parameters of a given assembly. We show that we can do this by using albedo boundary conditions that are estimated with low computational cost. We also present an efficient way to tabulate these effects to permit accurate interpolation by the core-level algorithm. We describe our algorithms and present results from several difficult test problems containing mixed-oxide and UO{sub 2} assemblies. Our methodology significantly reduces the largest errors made by present-day methodology. For example, in our test problems it reduces the maximum pin-power error by a factor of {approx}5.
38 citations
TL;DR: In this article, an improvement of the lattice code component related to resonance self-shielding calculations is described, which is based on a subgroup flux equation with probability tablature.
Abstract: Improvement of the lattice code component related to resonance self-shielding calculations is described. The proposed self-shielding model is based on a subgroup flux equation with probability tabl...
35 citations
TL;DR: In this paper, the gamma-ray emission probabilities from the beta decay of 239Np were also analyzed, and a value of 2.683 +- 0.012 barns was derived for the thermal capture cross section of 238U.
Abstract: The precise value of the thermal capture cross section of238U is uncertain, and evaluated cross sections from various sourcesdiffer by more than their assigned uncertainties. A number of theoriginal publications have been reviewed to assess the discrepant data,corrections were made for more recent standard cross sections andotherconstants, and one new measurement was analyzed. Due to the strongcorrelations in activation measurements, the gamma-ray emissionprobabilities from the beta decay of 239Np were also analyzed. As aresult of the analysis, a value of 2.683 +- 0.012 barns was derived forthe thermal capture cross section of 238U. A new evaluation of thegamma-ray emission probabilities from 239Np decay was alsoundertaken.
TL;DR: The CIAU method has been recently extended to evaluate the uncertainty of coupled three-dimensional neutronics/thermal-hydraulics calculations and the result is CIAU-TN, which gives an idea of the errors expected from the present computational tools.
Abstract: The best-estimate calculation results from complex system codes are affected by approximations that are unpredictable without the use of computational tools that account for the various sources of uncertainty. The code with (the capability of) internal assessment of uncertainty (CIAU) has been previously proposed by the University of Pisa to realize the integration between a qualified system code and an uncertainty methodology and to supply proper uncertainty bands each time a nuclear power plant (NPP) transient scenario is calculated. The derivation of the methodology and the results achieved by the use of CIAU are discussed to demonstrate the main features and capabilities of the method. In a joint effort between the University of Pisa and The Pennsylvania State University, the CIAU method has been recently extended to evaluate the uncertainty of coupled three-dimensional neutronics/ thermal-hydraulics calculations. The result is CIAU-TN. The feasibility of the approach has been demonstrated, and sample results related to the turbine trip transient in the Peach Bottom NPP are shown. Notwithstanding that the full implementation and use of the procedure requires a database of errors not available at the moment, the results give an idea of the errors expected from the present computational tools.
TL;DR: This paper aims at presenting several extensions of the classical TPD, in which additional modeling capabilities are progressively introduced, and sketches a discretized approach of these problems.
Abstract: The theory of probabilistic dynamics (TPD) offers a framework capable of modeling the interaction between the physical evolution of a system in transient conditions and the succession of branchings...
TL;DR: In this article, the authors describe the creation of a comprehensive cross-section database for high-energy heavy-ion radiation fields in more than one dimension, with a wide range of energies and in three dimensions.
Abstract: To correctly specify the composition and spectra of transmitted heavy-ion radiation fields, such as those encountered in space radiation protection studies, accurate values of the total, elastic scattering, reaction cross sections, and spectral and angular distributions of all emitted particles (nucleons, light ions, and heavy ions) from the nuclear interactions of propagating high-energy heavy-ion particles with target nuclei are required. For space radiation protection studies, this means that double-differential (energy and angle) isotope production cross sections must be known for all stable nuclear isotopes with mass numbers from 1 to about 60 colliding with any target nucleus at energies from tens of mega-electron-volts per nucleon up to several giga-electron-volts per nucleon. Currently there are several radiation transport codes that transport high-energy nucleons, light ions, heavy ions, or some combination of them. None, however, transport all of these particles in more than one dimension. In order to make a comprehensive tool for space applications that transports all of these particles, with a wide range of energies and in three dimensions, the database described above is needed, particularly for light and heavy ions. This paper discusses the creation of this comprehensive cross-section database.
TL;DR: In this paper, the criticality analysis for a pebble bed reactor, HTR-10, is performed with Monte Carlo simulations with ENDF/B-VI continuous energy cross sections.
Abstract: In this study, the criticality analysis for a pebble bed reactor, HTR-10, is performed with Monte Carlo simulations. The MCNP4B code package is utilized in the analysis with ENDF/B-VI continuous energy cross sections. The full core with the initial loading case is considered in simulations. The variation of the effective multiplication factor as a function of core loading height is also analyzed. Three different geometrical models are employed to see the effect of geometrical detail on the criticality calculations. Results are compared with diffusion calculations as well as the experimental data. Results show that the use of the homogenized fuel zone model does not yield acceptable results and underestimates the core criticality. However, the results obtained by using models with uniform and randomly distributed coated fuel particles in the fuel zone are in quite good agreement and there is not any systematic difference. Furthermore, criticality values do not change significantly with different random arrangements of coated fuel particles in fuel spheres. However, the random and irregular arrangements of pebbles may result in statistically different criticality values at least due to varying streaming effect.
TL;DR: In this paper, the authors have developed the Contributon and Point-wise Cross Section Driven (CPXSD) methodology for constructing effective fine and broad-group cross-section structures.
Abstract: Multigroup cross sections are one of the major factors that cause uncertainties in the results of deterministic transport calculations. Thus, it is important to prepare effective cross-section libraries that include an appropriate group structure and are based on an appropriate spectrum. There are several multigroup cross-section libraries available for particular applications. For example, the 47-neutron, 20-gamma group BUGLE library that is derived from the 199-neutron, 42-gamma group VITAMIN-B6 library is widely used for light water reactor (LWR) shielding and pressure vessel dosimetry applications. However, there is no publicly available methodology that can construct problem-dependent libraries. Thus, the authors have developed the Contributon and Point-wise Cross Section Driven (CPXSD) methodology for constructing effective fine- and broad-group structures. In this paper, new fine-group structures were constructed using the CPXSD, and new fine-group cross-section libraries were generated. Th...
TL;DR: In this paper, a statistically sound criterion using Shannon and relative entropies is defined based on the inequality with a penalty term for the minimum descriptive length of instantaneously decodable encoding.
Abstract: The criterion of information-theoretic stationarity diagnostics for the Monte Carlo simulation of nuclear criticality has been extended to undersampling diagnostics. Here, undersampling diagnostics means the posterior checking of the number of neutron histories per cycle. A statistically sound criterion using Shannon and relative entropies is defined based on the inequality with a penalty term for the minimum descriptive length of instantaneously decodable encoding. An alternative criterion based on a large sample property of particle population is defined within the information-theoretic framework of the asymptotic equipartition property and the method of types. An auxiliary criterion is proposed using the concave property of Shannon entropy. Numerical results are presented for the 'k-effective of the world' problem by Whitesides. The results indicate that the estimation bias of the neutron effective multiplication factor will be reduced to a practically negligible level if these criteria are satisfied. It can be concluded that equilibrium is a stronger condition than stationarity concerning the source distribution in the Monte Carlo simulation.
TL;DR: In this article, the oxidation properties of different grades of nuclear graphite (PAEB, PCEB, PPEA, and IG-11) were studied thermogravimetrically at 400, 800, and 1200 deg.
Abstract: The oxidation behaviors of different grades of nuclear graphite - PAEB, PCEB, PPEA, and IG-11 - were studied thermogravimetrically at 400, 800, and 1200 deg. C as a part of work to select one grade of nuclear graphite for use in a gas turbine-modular helium reactor (GT-MHR). The results showed that all grades of nuclear graphite resist oxidation at 400 deg. C. The difference in oxidation between different grades of nuclear graphite was greater at 800 deg. C than at 400 deg. C and 1200 deg. C. At 800 deg. C, for the same grade of nuclear graphite, when the centerline of the specimen is parallel to the axis of extrusion (with grain), the oxidation rate is greater than that of the graphite specimen with the centerline perpendicular to the axis of extrusion (against grain). The experimental results revealed that PPEA had the best oxidation resistance, and IG-11 had the worst due to high impurities. Moreover, the oxidation experiment exhibited that there were some oxidizable materials in unclear nuclear graphite.
TL;DR: In this paper, a numerical methodology of sodium-water reaction (SWR) and a coupling method of SWR and multiphase flow analysis are proposed. But the method is not applicable to the coupling phenomena in SWR.
Abstract: A numerical methodology of sodium-water reaction (SWR) and a coupling method of SWR and multiphase flow analysis are proposed. Two SWR models are considered. One is a surface reaction model, which assumes that water vapor reacts with liquid sodium at the gas-liquid interface. The surface reaction is likely to be dominant in the initial phase of SWR. The analogy between mass and heat transfers is assumed to evaluate the diffusion-controlled reaction rate. The other is a gas-phase reaction model. If chemical reaction heating due to the surface reaction is large enough to vaporize the liquid sodium, it turns over in the gas-phase reaction. In the gas-phase reaction, water vapor reacts with sodium gas. The reaction mechanisms in the gas-phase reaction are investigated using an ab initio molecular orbital method. The reaction rate of the gas-phase reaction described by the Arrhenius law is obtained from the transition-state theory or the capture theory. The reaction models are employed in a compressible multifluid and one-pressure model using the Highly Simplified Marker and Cell method for multiphase flow analysis. As numerical examples, surface reaction with multiphase flow analysis and simplified gas-phase reaction analyses are carried out. It is confirmed that the present method ismore » practically applicable to the coupling phenomena of SWR and multiphase flow.« less
TL;DR: In this paper, a modified nodal integral method (MNIM) for two-dimensional, time-dependent Navier-Stokes equations is extended to three dimensions, based on local transverse integrations over finite size cells that reduce each partial differential equation to a set of ordinary differential equations.
Abstract: A modified nodal integral method (MNIM) for two-dimensional, time-dependent Navier-Stokes equations is extended to three dimensions. The nodal integral method is based on local transverse integrations over finite size cells that reduce each partial differential equation to a set of ordinary differential equations (ODEs). Solutions of these ODEs in each cell for the transverse-averaged dependent variables are then utilized to develop the difference schemes. The discrete variables are scalar velocities and pressure, averaged over the faces of bricklike cells. The development of the MNIM is different from the conventional nodal method in two ways: (a) it is Poisson-type pressure equation based and (b) the convection terms are retained on the left side of the transverse-integrated equations and thus contribute to the homogeneous part of the solution. The first feature leads to a set of symmetric transverse-integrated equations for all the velocities, and the second feature yields distributions of constant + linear + exponential form for the transverse-averaged velocities. The scheme is tested on three-dimensional lid-driven cavity problems in cube- and prism-shaped cavities. Results obtained using the MNIM on fairly coarse meshes are comparable with reference solutions obtained using much finer meshes.
TL;DR: In this article, the impact of higher PWR fuel burnup is examined from the perspective of its impact on spent-fuel radioactivity, decay heat, and plutonium content, and the necessary fresh fuel enrichments to achieve high burnup in PWRs with the same three-batch operation scheme are first computed; then, characteristics of the spent fuel are determined.
Abstract: Reducing the burden of management of spent nuclear fuel is important to the future of nuclear energy. The impact of higher pressurized water reactor (PWR) fuel burnup is examined in this paper from the perspective of its impact on spent-fuel radioactivity, decay heat, and plutonium content. The necessary fresh fuel enrichments to achieve high burnup in PWRs with the same three-batch operation scheme are first computed; then, characteristics of the spent fuel are determined. The increase in decay heat with burnup is found to be generally less than linear. Although each high-burnup fuel assembly would be hotter and more radioactive, the total decay heat to be removed or accommodated in storage is less for the same electricity production. If the time window before 150 yr after discharge can be excluded from impacting a repository, significant savings in its capacity can be realized with high-burnup fuel. The high-burnup fuel is more proliferation resistant because of reduced total plutonium production per kilowatt hour and because of higher content of less desirable plutonium isotopes, such as 238 Pu. The fuel cycle cost can be slightly reduced by increasing burnup until it reaches a shallow minimum near 70 MWd/kg. Higher burnups would require one-time changes to the limits on enrichments that can be handled in most com mercial fuel fabrication facilities. Changing the waste fee to base it on the amount of radioactivity in the spent fuel would enhance the economic benefit of high burnup.
TL;DR: In this paper, the authors simulate the interaction between hyperalkaline solutions derived from the degradation of cement and potential host rocks for repositories for low-and intermediate-level radioactive waste in two different cases: (a) the planned repository at Wellenberg and (b) the modeling of the GTS-HPF experiment at the Grimsel Test Site.
Abstract: Reactive transport calculations simulating the interaction between hyperalkaline solutions derived from the degradation of cement and potential host rocks for repositories for low- and intermediate-level radioactive waste have been performed. Two different cases are shown: (a) The example of the planned repository at Wellenberg and (b) the modeling of the GTS-HPF experiment at the Grimsel Test Site. The GIMRT code has been used for the simulations. Mineral reactions are described by kinetic rate laws. The reaction rates for the primary minerals are based on experimentally determined rates published in the literature and geometric considerations combined with measurements regarding mineral surface areas. Relatively fast rates for the secondary minerals have been used, so the results resemble the local equilibrium solution for these minerals. In both cases, the alteration of the rock and the precipitation of secondary phases cause a reduction in the permeability of the system, which would actually be beneficial for the performance of a repository. Mineral surface area controls, to a large extent, the amount of mineral alteration and the change in permeability.
TL;DR: The feasibility of using the detection of electron antineutrinos produced in fission to monitor the time dependence of the plutonium content of nuclear power reactors is discussed in this paper.
Abstract: The feasibility of using the detection of electron antineutrinos produced in fission to monitor the time dependence of the plutonium content of nuclear power reactors is discussed. If practical, such a scheme would allow worldwide, automated monitoring of reactors and, thereby, the detection of certain proliferation scenarios. For GW(electric) power reactors, the count rates and the sensitivity of the antineutrino spectrum (to the core burnup) suggest that monitoring of the gross operational status of the reactor from outside the containment vessel is feasible. As the plutonium content builds up in a given burn cycle, the total number of antineutrinos steadily drops; and this variation is quite detectable, assuming fixed reactor power. The average antineutrino energy also steadily drops, and a measurement of this variation would be very useful to help offset uncertainties in the total reactor power. However, the expected change in the antineutrino signal from the diversion of a significant quantity of plutonium, which would typically require the diversion of as little as a single fuel assembly in a GW(electric) reactor, would be very difficult to detect.
TL;DR: In this paper, a computationally efficient single event Monte Carlo approach for calculating dose from electrons is presented. Analog elastic scattering and inelastic energy-loss differential cross sections for ele...
Abstract: We present a computationally efficient single event Monte Carlo approach for calculating dose from electrons. Analog elastic scattering and inelastic energy-loss differential cross sections for ele...
TL;DR: In this paper, a spatial discretization method was developed for low-order quasi-diffusion equations on coarse grids and corresponding homogenization procedure for full-core reactor calculations.
Abstract: Spatial discretization methods have been developed for the low-order quasi-diffusion equations on coarse grids and corresponding homogenization procedure for full-core reactor calculations. The proposed methods reproduce accurately the complicated large-scale behavior of the transport solution within assemblies. The developed discretization is spatially consistent with a fine-mesh discretization of the transport equation in the sense that it preserves a set of spatial moments of the fine-mesh transport solution over either coarse-mesh cells or its subregions, as well as the surface currents and eigenvalue. To demonstrate accuracy of the proposed methods, numerical results are presented of calculations of test problems that simulate the interaction of mixed-oxide and uranium assemblies.
TL;DR: In this article, a lumped linear discontinuous spatial discretization for Sn calculations on tetrahedral meshes is proposed for applications such as thermal radiative transfer, where resi...
Abstract: A lumped, linear discontinuous spatial discretization for Sn calculations on tetrahedral meshes is described. This method is designed for applications such as thermal radiative transfer, where resi...
TL;DR: Explicit relations were developed to estimate isotope enrichment factors for iQF6 vapors diluted in a carrier gas G, which are isotope selectively laser-excited and flow subsonically through a wall-... as mentioned in this paper.
Abstract: Explicit relations are developed to estimate isotope enrichment factors for iQF6 vapors diluted in a carrier gas G, which are isotope selectively laser-excited and flow subsonically through a wall-...
TL;DR: In this article, a model based on the point reactor kinetic equations has been developed to investigate these reactivity effects and validated against experi- mental results from the Los Alamos National Laboratory uranyl fluoride Solution High-Energy Burst Assembly (SHEBA).
Abstract: One of the primary methods to produce medical isotopes, such as 99 Mo, is by irradiation of uranium targets in heterogeneous reactors. Solution reactors present a potential alternative to produce medical isotopes. The Medical Isotope Production Reactor (MIPR) concept has been proposed to produce medical isotopes with lower uranium consumption and waste than those in heterogeneous reactors. Com- mercial production of medical isotopes in solution reactors requires steady-state operation at ;200 kW. At this power regime, fuel-solution temperature increase and radiolytic-gas bubble formation introduce a negative reactivity feedback that has to be mitigated. A model based on the point reactor kinetic equations has been developed to investigate these reactivity effects. This model has been validated against experi- mental results from the Los Alamos National Laboratory uranyl fluoride Solution High-Energy Burst Assembly (SHEBA) and shows the feasibility of solution reactors for the commercial production of medical isotopes.
TL;DR: The results produced by this three-scale approximation of the point-kinetics equations are shown to be practically as accurate as the numerical resultsproduced by the Kaganove-type algorithms used in production codes, yet at significantly less cost in computational time and resources.
Abstract: The derivation of a closed-form expression is presented for a three-timescale approximation of the point-kinetics equations with two effective groups of delayed neutrons. The results produced by this three-scale approximation are shown to be practically as accurate as the numerical results produced by the Kaganove-type algorithms used in production codes, yet at significantly less cost in computational time and resources. Potential uses of this approximation for increasing the efficiency of production codes for computing the space-time distribution of neutrons in reactors are also indicated.
TL;DR: Both analytical models of this parallel algorithm and performance analysis are presented and the message-passing model is used to communicate the local solutions between processes participating in solving a problem.
Abstract: Recent advances in parallel software development for solving three-dimensional (3-D) neutron transport problems using the characteristics method are presented. The characteristics method solves the transport equation by collecting local angular fluxes along neutron paths. In order to be able to solve large 3-D transport problems in a reasonable time frame, the characteristics solver needs to be accelerated. After applying adequate numerical acceleration techniques, the only issue is to parallelize the solver. The parallelization of this solver is based on distributing a group of tracks, generated by a ray-tracing procedure, on several processors. Different distributing schemes and load-balancing techniques based on a calculation load model are presented. A message-passing model is used to communicate the local solutions between processes participating in solving a problem. Both analytical models of this parallel algorithm and performance analysis are presented and illustrated by several examples.