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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 91 records · Page 5
All sourcesNASA NTRSDOE OSTI
DOE OSTI2015

Representability evaluation of fuel debris nuclear characteristics by heterogeneous core of Stacy - 14581

Decommissioning is underway in the Fukushima Daiichi Nuclear Power Station (1FNPS). Criticality control is one of the important safety issues in the decommissioning, especially during fuel debris removal. In 1FNPS, part of fuel debris might be in the form of molten-core-concrete-interaction (MCCI) product. It has been shown that the MCCI product can be critical not only under water but even with water bound in concrete. Further study of criticality characteristics by computation is being conducted to clarify its criticality risk. Furthermore, the risk assessment by computation needs validation by experiment. Japan Atomic Energy Agency is preparing critical experiments with simulated fuel debris in the Static Experiment Critical Facility (STACY), whose solution core will be converted into a new core using fuel rods and light-water moderator. It is difficult to build core configurations composed of water and fuel rods which contain homogeneous mixtures of uranium dioxide (UO{sub 2}) and structural materials of boiling light water reactors. In the modified STACY, it is being planned to simulate fuel debris by arranging UO{sub 2} fuel rods and structural material rods in water. In this study, difference in heterogeneity between the experimental core and possible fuel debris conditions is evaluated by comparing infinite multiplication factors and neutron energy spectra. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2011

C5 Benchmark Problem with Discrete Ordinate Radiation Transport Code DENOVO

The C5 benchmark problem proposed by the Organisation for Economic Co-operation and Development/Nuclear Energy Agency was modeled to examine the capabilities of Denovo, a three-dimensional (3-D) parallel discrete ordinates (S{sub N}) radiation transport code, for problems with no spatial homogenization. Denovo uses state-of-the-art numerical methods to obtain accurate solutions to the Boltzmann transport equation. Problems were run in parallel on Jaguar, a high-performance supercomputer located at Oak Ridge National Laboratory. Both the two-dimensional (2-D) and 3-D configurations were analyzed, and the results were compared with the reference MCNP Monte Carlo calculations. For an additional comparison, SCALE/KENO-V.a Monte Carlo solutions were also included. In addition, a sensitivity analysis was performed for the optimal angular quadrature and mesh resolution for both the 2-D and 3-D infinite lattices of UO{sub 2} fuel pin cells. Denovo was verified with the C5 problem. The effective multiplication factors, pin powers, and assembly powers were found to be in good agreement with the reference MCNP and SCALE/KENO-V.a Monte Carlo calculations.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2015

Experimental study for subcriticality measurement of fuel debris in the Fukushima Daiichi reactor - 14582

Fuel debris defueling in the Fukushima Daiichi reactors is planned. Furthermore, retrieved fuel debris will be packed and stored in casks for a long time. In these processes, from the viewpoint of criticality safety, the subcriticality of the fuel debris should be confirmed. Although there are some subcriticality measurement methods, the applicability of these methods to the fuel debris must be investigated because the elemental composition and shape of the fuel debris are unknown. In this study, we focus on the exponential experiment method which is able to measure a wide range of subcriticality. Basic experiments to validate the methods were carried out using Toshiba Nuclear Critical Assembly (NCA) and {sup 3}He Position Sensitive Detector (PSD). The decay curves of neutron flux distributions were measured by the PSD detector and fitted to a theoretical formula. The elemental compositions and neutron spectrum of the fuel debris have been estimated with a severe accident code and a Monte Carlo code. In the experiment, the neutron spectrum of the fuel debris was simulated in the NCA core. The porosity of the fuel debris was assumed to be 60%, which corresponds to be H/U=7.5 and Volume Ratio=1.5. In order to evaluate the effect of the uncertainty of the compositions on the measured subcriticality, a homogeneous region and a heterogeneous region were configured in an experimental core. These regions consisted of fuel rods with different UO{sub 2} enrichment zone or different gadolinia concentration zone in the axial direction. The effective multiplication factors deduced by the exponential experiment method were a range of 0.94 to 0.98. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Validation and bias quantification of criticality safety codes for SRNS operations

The validation and bias quantification methods employed by the Criticality Safety Engineering group at Savannah River Nuclear Solutions (SRNS) are discussed. Both SCALE and MCNP are validated against benchmark experiments, where these benchmark models have been divided into subgroups representative of the types of fissile material operations at SRNS. A quantification of the bias is made for each subgroup for each code and the three methods employed for quantification are discussed. The bias information is employed directly in the establishment of the Upper Safety Limit (USL) value of the neutron multiplication factor when using standardized documented SRNS methods. A simple example of validation and bias quantification is given and the procedure for setting the USL using the derived bias information is also presented. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

Modeling resonance interference by 0-D slowing-down solution with embedded self-shielding method

The resonance integral table based methods employing conventional multigroup structure for the resonance self-shielding calculation have a common difficulty on treating the resonance interference. The problem arises due to the lack of sufficient energy dependence of the resonance cross sections when the calculation is performed in the multigroup structure. To address this, a resonance interference factor model has been proposed to account for the interference effect by comparing the interfered and non-interfered effective cross sections obtained from 0-D homogeneous slowing-down solutions by continuous-energy cross sections. A rigorous homogeneous slowing-down solver is developed with two important features for reducing the calculation time and memory requirement for practical applications. The embedded self-shielding method (ESSM) is chosen as the multigroup resonance self-shielding solver as an integral component of the interference method. The interference method is implemented in the DeCART transport code. Verification results show that the code system provides more accurate effective cross sections and multiplication factors than the conventional interference method for UO{sub 2} and MOX fuel cases. The additional computing time and memory for the interference correction is acceptable for the test problems including a depletion case with 87 isotopes in the fuel region. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2012

Verification of a Depletion Method in SCALE for the Advanced High Temperature Reactor

This study describes a new method utilizing the Dancoff factor to model a non-standard TRISO fuel form characteristic of the AHTR reactor design concept for depletion analysis using the TRITON sequence of SCALE and the validation of this method by code-to-code comparisons. The fuel used in AHTR has the TRISO particles concentrated along the edges of a slab fuel element. This particular geometry prevented the use of a standard DOUBLEHET treatment, previously developed in SCALE to handle NGNP-designed fuel. The new method permits fuel depletion on complicated geometries that traditionally can be handled only by continuous energy based depletion code systems. The method was initially tested on a fuel design typical of the NGNP, where the DOUBLEHET treatment is available. A more comprehensive study was performed using the VESTA code that uses the continuous energy MCNP5 code as a transport solver and ORIGEN2.2 code for depletion calculations. Comparisons of the results indicate good agreement of whole core characteristics, such as the multiplication factor, and the isotopics, including their spatial distribution. Key isotopes analyzed included 235U, 239Pu, 240Pu and 241Pu. The results from this study indicate that the Dancoff factor method can generate estimates of core characteristics with reasonable precision for scoping studies of configurations where the DOUBLEHET treatment is unavailable.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2012

CASMO5/TSUNAMI-3D spent nuclear fuel reactivity uncertainty analysis

The CASMO5 lattice physics code is used in conjunction with the TSUNAMI-3D sequence in ORNL's SCALE 6 code system to estimate the uncertainties in hot-to-cold reactivity changes due to cross-section uncertainty for PWR assemblies at various burnup points. The goal of the analysis is to establish the multiplication factor uncertainty similarity between various fuel assemblies at different conditions in a quantifiable manner and to obtain a bound on the hot-to-cold reactivity uncertainty over the various assembly types and burnup attributed to fundamental cross-section data uncertainty. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2010

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

A conceptual design study of Gas Cooled Fast Reactors with Modified CANDLE Burn-up has been performed. The objective of this research is to get optimal design parameters of such type reactors. The parameters of nuclear design including the critical condition, conversion ratio, and burn-up level were compared. These parameters are calculated by variation in the fuel fraction 47.5% up to 70%. Two dimensional full core multi groups diffusion calculations was performed by CITATION code. Group constant preparations are performed by using SRAC code system with JENDL-3.2 nuclear data library. In this design the reactor cores with cylindrical cell two dimensional R-Z core models are subdivided into several parts with the same volume in the axial directions. The placement of fuel in core arranged so that the result of plutonium from natural uranium can be utilized optimally for 10 years reactor operation. Modified CANDLE burn-up was established successfully in a core radial width 1.4 m. Total thermal power output for reference core is 550 MW. Study on the effect of fuel to coolant ratio shows that effective multiplication factor (k{sub eff}) is in almost linear relations with the change of the fuel volume to coolant ratio.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Validation of MVP code with HTC critical experiments

In the 1980's, a series of high burnups critical experiments (HTC) was conducted at the Apparatus B experimental facility in Valduc with the support of the IRSN and AREVA NC. Four series of experiments were designed to assess profit associated with actinide-only burn-up credit in the criticality safety analysis for fuel handling (Phases 1 and 2), pool storage (Phase 3) and spent fuel cask (Phase 4) conditions. The HTC rods, specifically fabricated for the experiments, were designed to simulate a typical pressurized water reactor uranium oxide spent fuel. In order to validate the MVP-2.0 code and relatively new nuclear data libraries for the criticality safety analysis, the HTC critical experiments were analyzed using the MVP-2.0 code with JENDL-4.0, ENDF/B-VII.1 or JEFF-3.2. The effective neutron multiplication factor k{sub eff} values were obtained through analyses of all phases of the HTC experiments. It was shown that averaged values of k{sub eff} using JENDL-4.0 and ENDF/B-VII.1 were in good agreement with experimental results for all the phases, while those using JEFF-3.2 showed slight underestimation. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2013

Numerical solution of the time dependent neutron transport equation by the method of the characteristics

This study presents three numerical algorithms to solve the time dependent neutron transport equation by the method of the characteristics. The algorithms have been developed taking into account delayed neutrons and they have been implemented into the novel MCART code, which solves the neutron transport equation for two-dimensional geometry and an arbitrary number of energy groups. The MCART code uses regular mesh for the representation of the spatial domain, it models up-scattering, and takes advantage of OPENMP and OPENGL algorithms for parallel computing and plotting, respectively. The code has been benchmarked with the multiplication factor results of a Boiling Water Reactor, with the analytical results for a prompt jump transient in an infinite medium, and with PARTISN and TDTORT results for cross section and source transients. The numerical simulations have shown that only two numerical algorithms are stable for small time steps.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Verification of the 2D Finite Element SP3 Solutions

Method(FEM) is the free geometry handling. However many of the geometries of the nuclear engineering problems are not unstructured but rather structured, such as a hexahedral, or triangular prism. Thus, the need for the use of FEM has less portion in nuclear engineering. However it is also certain that it is very hard to exclude the possibility for use of an unstructured geometry such as a PBMR core problem, fusion blanket analysis of the source problem, another design issue with an unstructured geometry. In addition, research regarding the acceleration scheme such as Coarse Mesh Finite Difference (CMFD) showed up in nuclear engineering in recent years. By considering these facts, the FEM cannot be totally exclusive. To utilize the FEM for nuclear engineering problems, several types of assistance are needed. The first is parallelization because of its slow calculation speed. The second is an acceleration scheme. With these two types of assistance the FEM can be utilized freely in the nuclear engineering field. In this research, before conducting parallelization and acceleration, the final verification for the implementation of the up-scattering treatment with an arbitrary geometry conducted through a comparison with the McCARD solution through the well-known C5G7 2D benchmark problem. Note that the earlier work for verification of the FEM for the multi-group problem without up-scattering and with a structured geometry including a mathematical formulation has already been done. The main points that will be tested here are as follows. The first is a check of the right implementation for the multi-group problem with an up-scattering cross section from the perspectives of the multiplication factor and power. The second is the transport effect for this problem by comparing the solutions from P{sub 1} and Simplified P{sub 3}(SP{sub 3}). The third is the effectiveness of the higher order basis function for the same number of unknowns. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2012

Development of 3D full-core ERANOS-2.2/MCNPX-2.7.0 models and neutronic analysis of the BFS-2 zero-power facility

The present paper is addressing the development and validation against experimental data of 3D full-core models of the BFS-2 zero-power fast-reactor using both the deterministic system code ERANOS-2.2 and the stochastic code MCNPX-2.7.0. The model configuration of BFS considered for analysis is the BFS-62-3A benchmark. To extend the - deterministic/stochastic - code-to-code comparison, neutronic parameters, i.e. reactivity, neutron spectrum and reaction rates, were also simulated at the cell level with the Monte Carlo code SERPENT-1.1.7 with two modern data libraries, ENDF-B/VII and JEFF-3.1.1. The BFS-2 critical zero-power facility at the Inst. of Physics and Power Engineering (IPPE) was designed for simulations of the core and shielding of sodium-cooled, fast reactors, for neutron data validation and comparison with experimental results. At the BFS-2 facility, the BFS-62-3A critical benchmark experiment was set-up as a mock-up of the BN-600 reactor, with hybrid MOX fuel and stainless steel reflectors. A UO{sub 2} blanket and a large non-homogeneous stainless-steel reflector surround the core. The lattice is hexagonal of pitch 5.1 cm and metallic dowels are used to keep in central position cylindrical rods made of different types of material (fissile, fertile, blanket, plenum, shielding and absorber). A typical subassembly is formed in piling up various pellets of about 1 cm in height and 4.6 cm in diameter, conferring large heterogeneity in the axial direction. The full-core model development was a complex task due to the large number of subassemblies and the axial subassembly heterogeneity. In ERANOS-2.2, it was necessary to homogenize axially per region the pellets used to form the subassembly. The self-shielded macroscopic cross-sections were calculated using the cell code ECCO in association with JEFF-3.1 and ENDF/B-VI.8 data libraries. The core calculations were performed with broad cross-sections data in 33 neutron energy groups with the solver AVNM in the diffusion approximation, mostly. In MCNPX-2.7.0, a step-by-step approach was used, starting with a model in which the fissile rods were simulated on a homogeneous level, to finally integrate the actual heterogeneous description of the subassemblies. The code-to-code cell analysis performed between ECCO, SERPENT and MCNPX with different modern nuclear data library revealed that the results for the infinite multiplication factor between Monte Carlo and deterministic analysis are in good agreement ({Delta}p < 100 pcm). The differences between the results were observed to be larger for the neutron data libraries, with reactivity differences up to 350 pcm. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

On the characterization of biases arising from methods and approximations used for sensitivity analyses

The preliminary assessment of the performance and safety of advanced reactors, as well as the identification of research and development needs, rely on computer simulations due to the lack of operational experience. Those simulations should be based on verified and validated computational tools, calculation schemes, and nuclear data libraries, and should be accompanied of a proper characterization of the involved uncertainties, providing confidence in the computational-based predictions. Thus, sensitivity and uncertainty studies, together with integral experiments, play an essential role in that process. Sensitivities are then a critical element and guidelines about how to produce accurate enough sensitivities depending on the subsequent analyses to be performed are of interest. The present study addresses the sensitivities biases arising from the use of different nuclear data libraries, computational methods and the assumption of modelling simplifications. Not only sensitivities for multiplication factor but also for safety-relevant reactivity responses have been analyzed. In the study, propagated uncertainties are also included for the considered parameters to illustrate the impact of the sensitivity profiles obtained for each case on the final uncertainty. Then, relevant recommendations are given for sensitivity analyses of highly complex systems. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2011

Effect of inflow density on ion diffusion region of magnetic reconnection: Particle-in-cell simulations

We perform a systematic study of the effect of inflow density on reconnection diffusion regions using a 2.5-D particle-in-cell code. The diffusion region structures are analyzed at times when all simulations have reconnected the same amount of magnetic flux. We find that reducing the inflow density from 1 to 1/100th of the current sheet density dramatically increases the diffusion region physical size and the reconnection rate. The width of the diffusion region scales with the upstream ion inertial length systematically. Consistent with the presence of counter-streaming inflowing ion beams near the x-line, the ion meandering width in the diffusion region also scales with the ion inertial length. The aspect ratio of the ion diffusion region remains a constant, independent of the inflow density. The quadrupole Hall magnetic field is reduced. The upstream magnetic field deviates from its asymptotic value by {approx}50% at the lowest simulated inflow density. The downstream ion outflow velocity scales linearly with the upstream Alfven speed with a multiplication factor {approx}0.4 < 1. When applied to magnetic reconnection in the Earth's magnetotail, this factor of 0.4 is a possible explanation as to why bulk flow velocities in the magnetotail are typically on the order of 500 km/s, while the Alfven speeds of inflowing plasmas can exceed 2000 km/s.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2015

The Monte Carlo estimation of an effect of uncertainties in initial data on solving the transport equation by means of the MCU code

A new version of the tally module of the MCU software package is developed in which the approach for taking directly into account the uncertainty in initial data is implemented that is recommended by the international standard on estimating the uncertainty in results of measuring (ISO 13005). The new module makes it possible to evaluate the effect of uncertainty in initial data (caused by technological tolerances in fabrication of structural members of the core) on neutronic characteristics of the reactor. The developed software is adapted to parallel computing with the use of multiprocessor computers, which significantly reduces the computation time: the parallelization coefficient is almost equal to 1. Testing is performed by examples of solving the problem on criticality for the Godiva benchmark experiment and also for the infinite lattice of fuel assemblies of the VVER-440, VVER-1000, and VVER-1200. The results of calculations of the uncertainty in neutronic characteristics (effective multiplication factor, fission reaction rate), which is caused by uncertainties in initial data due to technological tolerances, are compared (in the first case) to the published results obtained using the precision MCNP5 code and (in the second case) to those obtained by means of the RADAR engineering program. A good agreement of results is achieved for all cases.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

Uncertainty analysis of a SFR core with sodium plenum

The new concepts of Sodium-cooled Fast Reactors have to reach the Generation IV safety objectives. In this regard the Sodium Void Effect has to be minimized for the future projects of large-size SFR as well as the uncertainties on it. The Inst. of Radiological Protection and Nuclear Safety (IRSN) as technological support of French public authorities is in charge of safety assessment of operating and under construction reactors, as well as future projects. In order to state about the safety of new SFR designs the IRSN must be able to evaluate core parameters and their uncertainties. In this frame a sensitivity and uncertainty study has been performed to evaluate the impact of nuclear data uncertainty on sodium void effect, for the benchmark model of large SFR BN-800. The benchmark parameters (effective multiplication factor and sodium void effect) have been evaluated using two codes, the deterministic code ERANOS and the Monte Carlo code SCALE, while the S/U analysis has been performed only with SCALE. The results of the these studies point out the most relevant cross section uncertainties that affect the SVE and how efforts should be done in increasing the existing nuclear data accuracies. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2012

Impact of Fuel Failure on Criticality Safety of Used Nuclear Fuel

Commercial used nuclear fuel (UNF) in the United States is expected to remain in storage for considerably longer periods than originally intended (e.g., <40 years). Extended storage (ES) time and irradiation of nuclear fuel to high-burnup values (>45 GWd/t) may increase the potential for fuel failure during normal and accident conditions involving storage and transportation. Fuel failure, depending on the severity, can result in changes to the geometric configuration of the fuel, which has safety and regulatory implications. The likelihood and extent of fuel reconfiguration and its impact on the safety of the UNF is not well understood. The objective of this work is to assess and quantify the impact of fuel reconfiguration due to fuel failure on criticality safety of UNF in storage and transportation casks. This effort is primarily motivated by concerns related to the potential for fuel degradation during ES periods and transportation following ES. The criticality analyses consider representative UNF designs and cask systems and a range of fuel enrichments, burnups, and cooling times. The various failed-fuel configurations considered are designed to bound the anticipated effects of individual rod and general cladding failure, fuel rod deformation, loss of neutron absorber materials, degradation of canister internals, and gross assembly failure. The results quantify the potential impact on criticality safety associated with fuel reconfiguration and may be used to guide future research, design, and regulatory activities. Although it can be concluded that the criticality safety impacts of fuel reconfiguration during transportation subsequent to ES are manageable, the results indicate that certain configurations can result in a large increase in the effective neutron multiplication factor, k eff . Future work to inform decision making relative to which configurations are credible, and therefore need to be considered in a safety evaluation, is recommended.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

Neutronic analysis of a fusion hybrid reactor

In a PHYSOR 2010 paper(1) we introduced a fusion hybrid reactor whose fusion component is the gasdynamic mirror (GDM), and whose blanket was made of thorium - 232. The thrust of that study was to demonstrate the performance of such a reactor by establishing the breeding of uranium - 233 in the blanket, and the burning thereof to produce power. In that analysis, we utilized the diffusion equation for one-energy neutron group, namely, those produced by the fusion reactions, to establish the power distribution and density in the system. Those results should be viewed as a first approximation since the high energy neutrons are not effective in inducing fission, but contribute primarily to the production of actinides. In the presence of a coolant, however, such as water, these neutrons tend to thermalize rather quickly, hence a better assessment of the reactor performance would require at least a two group analysis, namely the fast and thermal groups. We follow that approach and write an approximate set of equations for the fluxes of these groups. From these relations we deduce the all-important quantity, k{sub eff}, which we utilize to compute the multiplication factor, and subsequently, the power density in the reactor. We show that k{sub eff} can be made to have a value of 0.99, thus indicating that 100 thermal neutrons are generated per fusion neutron, while allowing the system to function as 'subcritical.' Moreover, we show that such a hybrid reactor can generate hundreds of megawatts of thermal power per cm of length depending on the flux of the fusion neutrons impinging on the blanket. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗