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Ilas, Dan

Publications and source records attributed to Ilas, Dan.

At least 19 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2022

Neutron irradiation of tungsten in hydrogen environment at HFIR

Neutron irradiation of tungsten with and without the presence of hydrogen is needed to understand the influence of hydrogen on microstructure development under fusion reactor conditions. However, there is a risk of ignition if air ingress occurs during seal welding of irradiation capsules in a pressurized hydrogen environment. Therefore, an irradiation capsule was designed that contains several disks of vanadium hydrides at a 30% hydrogen-to-metal atomic ratio. During irradiation, the hydrogen is released from the hydrides as the capsule temperature increases, so the irradiation capsule environment is mostly hydrogen when the capsule reaches its 400 °C design temperature. In this study, we describe the design and operating characteristics of this first-of-a-kind irradiation capsule.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2016

Phenomena Identification and Ranking Table (PIRT) for Fluoride High-Temperature Reactor (FHR) Neutronics

The deployment of Fluoride (salt-cooled) High Temperature Reactors (FHRs) has many potential benefits such as improved safety, improved economics, and a diversification of the energy portfolio because of its capability to supply industrial process heat. However, significant challenges remain before this class of reactors can be deployed, mostly related to its technology readiness. A recent study identified several areas in need of further development in order to advance the technology. These include the verification and validation of neutronics and thermal hydraulics modeling and simulation tools in support of design, safety analysis and licensing. The focus of this paper is to summarize the issues for verification and validation of neutronics tools. Current tools for neutronics analysis of LWRs are inadequate for the simulation of salt cooled reactors. The methodologies developed for the high temperature gas cooled reactors (HTGRs) have some relevance for FHRs because of the similarities between the two reactors such as graphite block structure, encapsulated fuel particles, and high operating temperatures. However, they are not sufficient for the accuracy desired of a code for licensing as they have a number of important differences neutronically, such as the use of liquid salt containing beryllium as a coolant and their unique fuel configuration.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Phenomena Identification and Ranking Table (PIRT) for Fluoride High-Temperature Reactor (FHR) Neutronics

The deployment of Fluoride (salt-cooled) High Temperature Reactors (FHRs) has many potential benefits such as improved safety, improved economics, and a diversification of the energy portfolio because of its capability to supply industrial process heat. However, significant challenges remain before this class of reactors can be deployed, mostly related to its technology readiness. A recent study identified several areas in need of further development in order to advance the technology. These include the verification and validation of neutronics and thermal hydraulics modeling and simulation tools in support of design, safety analysis and licensing. The focus of this paper is to summarize the issues for verification and validation of neutronics tools. Current tools for neutronics analysis of LWRs are inadequate for the simulation of salt cooled reactors. The methodologies developed for the high temperature gas cooled reactors (HTGRs) have some relevance for FHRs because of the similarities between the two reactors such as graphite block structure, encapsulated fuel particles, and high operating temperatures. However, they are not sufficient for the accuracy desired of a code for licensing as they have a number of important differences neutronically, such as the use of liquid salt containing beryllium as a coolant and their unique fuel configuration. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2014

SmAHTR-CTC Neutronic Design

Building on prior experience for the 2010 initial SmAHTR neutronic design and on 2012 neutronic design for the advanced high temperature reactor (AHTR), this paper presents the main results of the neutronic design effort for the newly re-purposed SmAHTR-CTC reactor concept. The results are obtained based on full-core simulations performed with SCALE6.1. The dimensionality of the SmAHTR design space is reduced by using constraints originating in material fabricability, fuel licensing, molten salt chemistry, thermal-hydraulic and mechanical considerations. The new design represents in many regards a substantial improvement from the neutronic performance standpoint over the 2010 SmAHTR concept. Among other results, it is shown that fuel cycle length of over 2 years or discharged fuel burnup of 40GWd/MTHM are possible with a low, 8% fuel enrichment in a once-through fuel cycle, while 8-year once-through fuel cycle lengths are possible at higher fuel enrichments.

Ilas, Dan↗
DOE OSTI2013

Impact of HFIR LEU Conversion on Beryllium Reflector Degradation Factors

An assessment of the impact of low enriched uranium (LEU) conversion on the factors that may cause the degradation of the beryllium reflector is performed for the High Flux Isotope Reactor (HFIR). The computational methods, models, and tools, comparisons with previous work, along with the results obtained are documented and discussed in this report. The report documents the results for the gas and neutronic poison production, and the heating in the beryllium reflector for both the highly enriched uranium (HEU) and LEU HFIR configurations, and discusses the impact that the conversion to LEU may have on these quantities. A time-averaging procedure was developed to calculate the isotopic (gas and poisons) production in reflector. The sensitivity of this approach to different approximations is gauged and documented. The results show that the gas is produced in the beryllium reflector at a total rate of 0.304 g/cycle for the HEU configuration; this rate increases by ~12% for the LEU case. The total tritium production rate in reflector is 0.098 g/cycle for the HEU core and approximately 11% higher for the LEU core. A significant increase (up to ~25%) in the neutronic poisons production in the reflector during the operation cycles is observed for the LEU core, compared to the HEU case, for regions close to the core s horizontal midplane. The poisoning level of the reflector may increase by more than two orders of magnitude during long periods of downtime. The heating rate in the reflector is estimated to be approximately 20% lower for the LEU core than for the HEU core. The decrease is due to a significantly lower contribution of the heating produced by the gamma radiation for the LEU core. Both the isotopic (gas and neutronic poisons) production and the heating rates are spatially non-uniform throughout the beryllium reflector volume. The maximum values typically occur in the removable reflector and close to the midplane.

Beryllium↗
DOE OSTI2013

Hybrid Technique in SCALE for Fission Source Convergence Applied to Used Nuclear Fuel Analysis

The new hybrid SOURCE ConveRgence accelERator (SOURCERER) sequence in SCALE deterministically computes a fission distribution and uses it as the starting source in a Monte Carlo eigenvalue criticality calculation. In addition to taking the guesswork out of defining an appropriate, problem-dependent starting source, the more accurate starting source provided by the deterministic calculation decreases the probability of producing inaccurate tally estimates associated with undersampling problems caused by inadequate source convergence. Furthermore, SOURCERER can increase the efficiency of the overall simulation by decreasing the number of cycles that has to be skipped before the keff accumulation. SOURCERER was applied to a representative example for a used nuclear fuel cask utilized at the Maine Yankee storage site {Scaglione and Ilas}. Because of the time constraints of the Used Fuel Research, Development, and Demonstration project, it was found that using more than 30,000 neutrons per cycle will lead to inaccurate Monte Carlo calculation of keff due to the inevitable decrease in the number of skipped and active cycles used with this problem. For a fixed uncertainty objective and by using 30,000 neutron per cycle, the use of SOURCERER increased the efficiency of the keff calculation by 60%compared to a Monte Carlo calculation that used a starting source distributed uniformly in fissionable regions, even with the inclusion of the extra computational time required by the deterministic calculation. Additionally, the use of SOURCERER increased the reliability of keff calculation using any number of skipped cycles below 350.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

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

This study describes a new approach employing the Dancoff correction method to model the TRISO-based fuel form used by the Advanced High-Temperature Reactor (AHTR) reactor design concept. The Dancoff correction method is used to perform isotope depletion analysis using the TRITON sequence of SCALE and is verified by code-to-code comparisons. The current AHTR fuel design has TRISO particles concentrated along the edges of a slab fuel element. This geometry prevented the use of the DOUBLEHET treatment, previously developed in SCALE to model spherical and cylindrical 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 configuration typical of the Next Generation Nuclear Plant (NGNP), where DOUBLEHET treatment is possible. A confirmatory study was performed on the AHTR reference core geometry using the VESTA code, which 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 DOUBLEHET treatment cannot be performed.

AHTR↗
DOE OSTI2013

SCALE Code Validation for Prismatic High-Temperature Gas-Cooled Reactors

Using experimental data published in the International Handbook of Evaluated Reactor Physics Benchmark Experiments for the fresh cold core of the High Temperature Engineering Test Reactor, a comprehensive validation study has been carried out to assess the performance of the SCALE code system for analysis of high-temperature gas-cooled reactor configurations. This paper describes part of the results of that effort. The studies performed included criticality evaluations for the full core and for the annular cores realized during the fuel loading, as well as calculations and comparisons for excess reactivity, shutdown margin, control rod worths, temperature coefficient of reactivity, and reaction rate distributions. Comparisons of the SCALE results with both experimental values and MCNP-calculated values are presented. The comparisons show that the SCALE calculated results, obtained with both multigroup and continuous energy cross sections, are in reasonable agreement with the experimental data and the MCNP predictions.

code validation↗
DOE OSTI2013

Neutronics and Depletion Methods for Parametric Studies of Fluoride Salt Cooled High Temperature Reactors with Slab Fuel Geometry and Multi-Batch Fuel Management Schemes

The Advanced High-Temperature Reactor (AHTR) is a 3400 MWth fluoride-salt-cooled high-temperature reactor (FHR) that uses TRISO particle fuel compacted into slabs rather than spherical or cylindrical fuel compacts. Simplified methods are required for parametric design studies such that analyzing the entire feasible design space for an AHTR is tractable. These simplifications include fuel homogenization techniques to increase the speed of neutron transport calculations in depletion analysis and equilibrium depletion analysis methods to analyze systems with multi-batch fuel management schemes. This paper presents three elements of significant novelty. First, the Reactivity-Equivalent Physical Transformation (RPT) methodology usually applied in systems with coated-particle fuel in cylindrical and spherical geometries has been extended to slab geometries. Secondly, based on this newly developed RPT method for slab geometries, a methodology that uses Monte Carlo depletion approaches was further developed to search for the maximum discharge burnup in a multi-batch system by iteratively estimating the beginning of equilibrium cycle (BOEC) composition and sampling different discharge burnups. This Iterative Equilibrium Depletion Search (IEDS) method fully defines an equilibrium fuel cycle (keff, power, flux, and composition evolutions) but is computationally demanding, although feasible on single-processor workstations. Finally, an analytical method, the Non-Linear Reactivity Model, was developed by expanding the linear reactivity model to include an arbitrary number of higher order terms so that single-batch depletion results could be extrapolated to estimate the maximum discharge burnup and BOEC keff in systems with multi-batch fuel management schemes. Results from this method were benchmarked against equilibrium depletion analysis results using the IEDS.

fluoride-salt-cooled high-temperature reactor (FHR↗
DOE OSTI2012

Validation of SCALE for High Temperature Gas-Cooled Reactors Analysis

This report documents verification and validation studies carried out to assess the performance of the SCALE code system methods and nuclear data for modeling and analysis of High Temperature Gas-Cooled Reactor (HTGR) configurations. Validation data were available from the International Handbook of Evaluated Reactor Physics Benchmark Experiments (IRPhE Handbook), prepared by the International Reactor Physics Experiment Evaluation Project, for two different HTGR designs: prismatic and pebble bed. SCALE models have been developed for HTTR, a prismatic fuel design reactor operated in Japan and HTR-10, a pebble bed reactor operated in China. The models were based on benchmark specifications included in the 2009, 2010, and 2011 releases of the IRPhE Handbook. SCALE models for the HTR-PROTEUS pebble bed configuration at the PROTEUS critical facility in Switzerland have also been developed, based on benchmark specifications included in a 2009 IRPhE draft benchmark. The development of the SCALE models has involved a series of investigations to identify particular issues associated with modeling the physics of HTGRs and to understand and quantify the effect of particular modeling assumptions on calculation-to-experiment comparisons.

Ilas, Germina↗
DOE OSTI2012

AHTR Mechanical, Structural, And Neutronic Preconceptual Design

This report provides an overview of the mechanical, structural, and neutronic aspects of the Advanced High Temperature Reactor (AHTR) design concept. The AHTR is a design concept for a large output Fluoride salt cooled High-temperature Reactor (FHR) that is being developed to enable evaluation of the technology hurdles remaining to be overcome prior to FHRs becoming a commercial reactor class. This report documents the incremental AHTR design maturation performed over the past year and is focused on advancing the design concept to a level of a functional, self-consistent system. The AHTR employs plate type coated particle fuel assemblies with rapid, off-line refueling. Neutronic analysis of the core has confirmed the viability of a 6-month 2-batch cycle with 9 weight-percent enriched uranium fuel. Refueling is intended to be performed automatically under visual guidance using dedicated robotic manipulators. The present design intent is for used fuel to be stored inside of containment for at least 6 months and then transferred to local dry wells for intermediate term, on-site storage. The mechanical and structural concept development effort has included an emphasis on transportation and constructability to minimize construction costs and schedule. The design intent is that all components be factory fabricated into rail transportable modules that are assembled into subsystems at an on-site workshop prior to being lifted into position using a heavy-lift crane in an open-top style construction. While detailed accident identification and response sequence analysis has yet to be performed, the design concept incorporates multiple levels of radioactive material containment including fully passive responses to all identified design basis or non-very-low frequency beyond design basis accidents. Key building design elements include: 1) below grade siting to minimize vulnerability to aircraft impact, 2) multiple natural circulation decay heat rejection chimneys, 3) seismic base isolation, and 4) decay heat powered back-up electricity generation. The report provides a preconceptual design of the manipulators, the fuel transfer system, and the salt transfer loops. The mechanical handling of the fuel and how it is accomplished without instrumentation inside the salt is described within the report. All drives for the manipulators reside outside the reactor top flange. The design has also taken into account the transportability of major components and how they will be assembled on site

Varma, Venugopal Koikal↗
DOE OSTI2012

AHTR Mechanical, Structural, and Neutronic Preconceptual Design

This report provides an overview of the mechanical, structural, and neutronic aspects of the Advanced High Temperature Reactor (AHTR) design concept. The AHTR is a design concept for a large output Fluoride salt cooled High-temperature Reactor (FHR) that is being developed to enable evaluation of the technology hurdles remaining to be overcome prior to FHRs becoming an option for commercial reactor deployment. This report documents the incremental AHTR design maturation performed over the past year and is focused on advancing the design concept to a level of a functional, self-consistent system. The reactor concept development remains at a preconceptual level of maturity. While the overall appearance of an AHTR design is anticipated to be similar to the current concept, optimized dimensions will differ from those presented here. The AHTR employs plate type coated particle fuel assemblies with rapid, off-line refueling. Neutronic analysis of the core has confirmed the viability of a 6-month two-batch cycle with 9 wt. % enriched uranium fuel. Refueling is intended to be performed automatically under visual guidance using dedicated robotic manipulators. The report includes a preconceptual design of the manipulators, the fuel transfer system, and the used fuel storage system. The present design intent is for used fuel to be stored inside of containment for at least six months and then transferred to local dry wells for intermediate term, on-site storage. The mechanical and structural concept development effort has included an emphasis on transportation and constructability to minimize construction costs and schedule. The design intent is that all components be factory fabricated into rail transportable modules that are assembled into subsystems at an on-site workshop prior to being lifted into position using a heavy-lift crane in an open-top style construction. While detailed accident identification and response sequence analysis has yet to be performed, the design concept incorporates fully passive responses to all identified design basis or non-very-low frequency beyond design basis accidents as well as multiple levels of radioactive material containment. Key building design elements include (1) below grade siting to minimize vulnerability to aircraft impact, (2) multiple natural circulation decay heat rejection chimneys, (3) seismic base isolation, and (4) decay heat powered back-up electricity generation.

01 COAL, LIGNITE, AND PEAT↗
DOE OSTI2012

SCALE Code Validation for Prismatic High-Temperature Gas-Cooled Reactors

Using experimental data published in the International Handbook of Evaluated Reactor Physics Benchmark Experiments for the fresh cold core of the High Temperature Engineering Test Reactor, a comprehensive validation study has been carried out to assess the performance of the SCALE code system for analysis of High Temperature Gas-Cooled Reactor (HTGR) configurations. This paper describes part of the results of this effort. The studies performed included criticality evaluations for the full core and for the annular cores realized during the fuel loading, as well as calculations and comparisons for excess reactivity, shutdown margin, control rod worths, temperature coefficient of reactivity, and reaction rate distributions. Comparisons of the SCALE results with both the experimental values and MCNP-calculated values are presented. The comparisons show that the SCALE calculated results, obtained with both multigroup and continuous energy cross sections, are in reasonable agreement with the experimental data. The agreement with the MCNP predictions is, in general, very good.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

Development of a Scale Model for High Flux Isotope Reactor Cycle 400

The development of a comprehensive SCALE computational model for the High Flux Isotope Reactor (HFIR) is documented and discussed in this report. The SCALE model has equivalent features and functionality as the reference MCNP model for Cycle 400 that has been used extensively for HFIR safety analyses and for HFIR experiment design and analyses. Numerical comparisons of the SCALE and MCNP models for the multiplication constant, power density distribution in the fuel, and neutron fluxes at several locations in HFIR indicate excellent agreement between the results predicted with the two models. The SCALE HFIR model is presented in sufficient detail to provide the users of the model with a tool that can be easily customized for various safety analysis or experiment design requirements.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2012

A Preliminary Analysis of Dose Rates Associated with ITER CVCS Equipment/Area Location

A preliminary analysis of the ITER Chemical and Volume Control System (CVCS) Area was performed to assess dose rates outside the walls and ceiling of the facility after 1.5 years of operation at shutdown, 2 days, and 10 days after shutdown. For this purpose a simplified Monte Carlo computer model was developed using the MCNP (MCNP5 Ver. 1.51) code. Two components are included: the smaller filter tank and the larger ion exchanger. These pieces of equipment are associated with the Integrated Blanket ELM Divertor Primary Heat Transfer System, which will have the largest dose rates associated with activated corrosion products during operation in comparison with other systems. The ion exchanger contained two source regions, a 1.2-m-thick resin bed above a 0.55 m-thick skirt, and a 0.8-m-thick water region. The filter constituted an additional source. Thus the model consisted of three sources (filter, resin, water), homogeneously distributed within the appropriate source regions. However, much of the results (that address individual isotopes) are presented with the two sources in the ion exchanger combined. In these cases the sources are referred to as the 'ion exchanger source' and the 'filter source.' Dimensions for the facility and components, as well as source isotopes and strengths, and material densities, were supplied by US ITER. Because of its simplification, the model does not contain pipes. Consequently, radiation streaming through pipe penetrations, radiation emanating from the pipes, and shielding from the pipes were not considered in this analysis. Dose rates on the outside of two walls and the ceiling were calculated. The two walls are labeled as the 'long' wall (aligned with the X-axis) and the 'short' wall (aligned with the Y-axis). These walls and ceiling were nominally set to 30-cm-thick concrete. In the original analysis, standard concrete (2.3 g/cc density) was used. In addition to the shielding walls/ceiling, a floor and an additional wall opposite the long wall were added for photon scattering contributions. These were both 10-cm-thick, standard concrete structures. Other components (tanks, pipes, etc.), that were not included in the model, would potentially add additional scattering and shielding. Possibly these additional effects will be addressed in a later, more detailed analysis. The room was 29.6 m in length (X-axis), but was limited to 15 m in the model. The inside width (Y axis) and height (Z axis) were 4 m and 3.4 m, respectively. The origin for the model was located inside the room at the corner opposite the long wall and adjacent to the short wall at the floor level. The room was filled with air at standard temperature and pressure. The stainless steel (SS304) wall thicknesses for the ion exchanger and filter were 2.2 cm and 0.8 cm, respectively. The axial center of the filter was located 140 cm from the short wall and 100 cm from the long wall (outer surface). The axial center of the ion exchanger was located 440 cm from the short wall and 250 cm from the long wall (inner surface). The resin was assumed to be a homogeneous mixture of equal atom density fractions of hydrogen and carbon* at a specified density of 1.136 g/cc. The filter material was assumed to be homogeneous carbon at a specified density of 1.8 g/cc. If the filter media were stainless steel and the accumulated activity were the same, the dose rate outside the filter would be lower, provided the density of the stainless steel sintered filter material is significantly higher than 1.8g/cc. The densities of the water and air were assumed to be 1.0 g/cc and 1.096E-3 g/cc, respectively. The model included 10 small volumes placed outside the model adjacent to the wall at locations where the dose rates were expected to be highest and which showed to a degree the fall-off of the dose rate with distance along a particular wall or ceiling. These 'tally cells' are shown in Fig. 2 and are described in Table 1. Each cell had dimensions of 50 cm x 50 cm x 1 cm and was oriented so that the 1-cm-thick dimension was perpendicular to the wall against which the cell was located. Also, smaller tally volumes were placed at 5 cm (updated model), 30 cm, 50 cm, and 100 cm from each source, both radially from the outside of each component at the height of each source mid-plane, and vertically, along the axial centerline, above the ion-exchanger and below the filter. (For the ion exchanger, the vertical distance was limited to 75 cm because of interference from the ceiling.) These tally regions can be seen in Fig. 3. Each cell had dimensions of 5 cm x 5 cm x 1 cm, oriented so that the 1-cm-thick dimension was perpendicular to the X, Y or Z axis.

08 HYDROGEN↗