SCALE Code System Research Articles

Numerical simulation of axial and radial decay heat distribution in spent RBMK-1500 fuel assembly

The RBMK-1500 reactor is a boiling water reactor. Thus, the coolant density and the neutron moderation efficiency vary axially over the fuel assembly height influencing the distribution of decay heat in the discharged fuel. The underestimation of this effect on the decay heat may lead to biases in safety assessments of the spent fuel management. Axial coolant density and power profiles were used in this work to determine axial and radial decay heat distributions in the RBMK-1500 fuel assembly over the first 100 years of cooling using the SCALE code system. It was found that the application of axial profiles influences the increase of estimated total decay heat by 3 % just 2 years after the irradiation is finished. However, the estimation of total decay heat is lower by 2 % after 100 years of cooling compared to the case with no applied axial profiles. The maximum of the normalized decay heat profile also varies between 1.31 and 1.40 over the cooling time. The estimated decay heat in outer pins over the cooling time is higher by 10 ÷ 14 % compared to the decay heat produced in inner pins. The variation of results over the cooling phase and reasons behind the difference in estimated decay heat was explained by analysing nuclides which mainly contribute to the total decay heat.

Equilibrium core modeling of a pebble bed reactor similar to the Xe-100 with SCALE

As the nuclear industry moves towards licensing and constructing advanced reactors, new attention has been focused on the advanced reactor designs that have past operational experience, such as pebble-bed high-temperature gas-cooled reactors (PB-HTGRs). Pebble-bed reactor designs have many advantages, such as their higher operating temperatures and online refueling capabilities. However, high-fidelity computational modeling of pebble-bed reactor designs, from reactor startup to operation at equilibrium, is more challenging compared to conventionally fueled reactors due to the continuous movement of the fuel pebbles through the reactor during operation. In previous work at Oak Ridge National Laboratory (ORNL), the SCALE Leap-In method for Cores at Equilibrium (SLICE) was developed around tools within the SCALE code system. This iterative method can effectively generate pebble-bed reactor zone-wise fuel inventories at equilibrium core operation within a reasonable computational time. The objective of this work was to further verify the ORNL SLICE method and to investigate the impact of considering temperature profiles during the application of the method. The SLICE method was applied to a modular high-temperature gas-cooled reactor design based upon publicly available design specifications of the Xe-100 pebble-bed reactor. Upon comparing the results from the SLICE method to published literature, the differences in the eigenvalue keffective were on the order of several hundred pcm (percent millirho). To investigate one possible cause of these differences, a study looking at the sensitivity of the full-core equilibrium keffective and discharge nuclide inventory to temperature was performed by developing equilibrium cores of two additional temperature profiles. From this temperature study, differences on the order of hundreds of pcm for the full-core equilibrium keffective, and up to 15% difference for the discharge inventories were found. These results indicated the strong dependence on temperature that needs to be considered for future work in equilibrium modeling of PB-HTGRs.

Key nuclear data for non-LWR reactivity analysis

An assessment of nuclear data performance for non-light-water reactor (non-LWR) reactivity calculations was performed at Oak Ridge National Laboratory that involved a thorough literature review to collect related observations made across different research institutions, an interrogation of the latest ENDF/B evaluated nuclear data libraries, and propagation of nuclear data uncertainties to key figures of merit associated with reactor safety for six non-LWR benchmarks. The outcome of this comprehensive study was published in a technical report issued by the US Nuclear Regulatory Commission. This paper provides a summary of the study’s key observations and conclusions and demonstrates with two examples how the various methods available in the SCALE code system were used to identify key cross section uncertainties for non-LWR reactivity analyses.

Analyses of the bias and uncertainty of SNF decay heat calculations using Polaris and ORIGEN

The bias and uncertainty of calculated decay heat from spent nuclear fuel (SNF) are essential for code validation. Also, predicting these quantities is crucial for deriving decay heat safety margins, influencing the design and safety of facilities at the back end of the nuclear fuel cycle. This paper aims to analyze the calculated spent nuclear fuel decay heat biases, uncertainties, and correlations. The calculations are based on the Polaris and ORIGEN codes of the SCALE code system. Stochastically propagated uncertainties of inputs and nuclear data into calculated decay heats are compared. Uncertainty propagation using the former code is straightforward. In contrast, the counterpart of ORIGEN necessitated the pre-generation of perturbed nuclear cross-section libraries using TRITON, followed by coincident perturbations in the ORIGEN calculations. The decay heat uncertainties and correlations have shown that the observed validation biases are insignificant for both Polaris and ORIGEN. Also, similarities are noted between the calculated decay heat uncertainties and correlations of both codes. The fuel assembly burnup and cooling time significantly influence uncertainties and correlations, equivalently expressed in both Polaris and ORIGEN models. The analyzed decay heat data are highly correlated, particularly the fuel assemblies having either similar burnup or similar cooling time. The correlations were used in predicting the validation bias using machine learning models (ML). The predictive performance was analyzed for machine learning models weighting highly correlated benchmarks. The application of random forest models has resulted in promising variance reductions and predicted biases significantly similar to the validation ones. The machine learning results were verified using the MOCABA algorithm (a general Monte Carlo-Bayes procedure). The bias predictive performance of the Bayesian approach is examined on the same validation data. The study highlights the potential of neighborhood-based models, using correlations, in predicting the bias of spent nuclear fuel decay heat calculations and identifying influential and highly similar benchmarks.

Uncertainty Propagation in UAM Time-Dependent Neutronics PWR Studies

A subset of pressurized water reactor (PWR) studies from Exercise II-2 of the Uncertainty Analysis in Modeling for Light Water Reactors (UAM-LWR) benchmark study has been performed to quantify the importance of both nuclear data uncertainties and manufacturing uncertainties in an assembly depletion calculation and a mini-core rod movement transient. The depletion study of the 15 × 15 PWR assembly using the SAMPLER and TRITON modules of the SCALE code system revealed a maximum uncertainty in keff of 0.49% for fresh fuel, decreasing to 0.38% at the midpoint of the fuel cycle and rising back to 0.48% at the end of the fuel cycle. Uncertainties and correlations of various homogenized cross sections and other group constant data, such as keff, have been determined, and the effect of randomly applied manufacturing uncertainties was found to be largely negligible relative to nuclear data uncertainties for bulk lattice parameters. However, for local parameters, such as the pin power factors, assembly discontinuity factors, and diffusion coefficients, the effects from manufacturing uncertainties were appreciable and sometimes dominant. Nuclear data uncertainties were found to be the dominant contributors to uncertainty in the isotopic composition of the overall assembly, with the exception of very early in the fuel cycle, where manufacturing uncertainties such as perturbations to the fuel density and pin radius made nonnegligible contributions to total uncertainty. The contribution of manufacturing uncertainties to isotopic uncertainties was nonnegligible at a pin-by-pin level, but still smaller than the contributions from nuclear data uncertainty. Studies of the PWR mini-core rod movement transient using homogenized data from the SCALE models in the PARCS diffusion code showed little difference between the tested modeling approaches and demonstrated that nuclear data uncertainties dominated the manufacturing uncertainties in the global figures of merit considered, such as the equilibrium core boron concentration, the maximum core power factor, and the maximum reactivity insertion. For local effects, such as maximum pin power during the transient, the randomly applied manufacturing uncertainties were dominant. It was found in general that for global system properties, nuclear data uncertainties made significantly larger contributions to total uncertainty, whereas for local parameters the impact of manufacturing uncertainties was at least nonnegligible, and for some parameters, dominant.

An Assessment of Heterogeneous Effects on System Reactivity for Criticality Safety Analyses with LEU+ and HALEU Materials

To develop the criticality safety basis for any system, process, or package, the worst-case configuration of materials resulting in the maximum system reactivity must be determined. It is commonly accepted that in terms of the maximum system reactivity, at the lower enrichments used in current commercial practice (i.e., 5 wt% 235U), a heterogeneous configuration is bounding of a homogeneous mixture of fissile and moderating materials. However, a common assumption made is that with increasing enrichment, a homogeneous system can be bounding. With increased industry interest in utilizing higher enrichments for commercial applications with low-enriched uranium (LEU+) (≤10 wt% 235U), and high assay low-enriched uranium (HALEU) (≤20 wt% 235U) materials, it has become increasingly important to verify any assumptions and to have a better understanding of the expected system behavior at these higher enrichments. The SCALE code system was used to assess the effects of heterogeneity on system reactivity with varying enrichments and system configurations for a UO2 and water system, typical of a transportation package criticality analysis. The purpose of this assessment was to provide insight on the effect of material heterogeneity on system reactivity with increasing enrichment. The results of this study confirm that for systems with a higher hydrogen-to–fissile material (H/X) ratio, the homogeneous mixture of material may be bounding for HALEU materials. However, for systems with a lower hydrogen-to–fissile material ratio (H/X ≤ 200), a heterogeneous configuration of contents is expected to be bounding for most LEU materials. Overall, for any LEU system, including HALEU material, heterogeneous reactivity effects should always be considered.

Processing of JEFF nuclear data libraries for the SCALE Code System and testing with criticality benchmark experiments

In the last years, a new version of the Joint Evaluated Fission and Fusion File (JEFF) data library, namely JEFF-3.3, has been released with relevant updates in the neutron reaction, thermal neutron scattering and covariance sub-libraries. In the frame of the EU H2020 SANDA project, severale efforts have been made to enable the use of JEFF nuclear data libraries with the extensively tested and verified SCALE Code System. With this purpose, AMPX processing code has been applied to enable such application, allowing to provide insight into the interaction between the code and the new versions of JEFF data file. This paper provides an overview about the processing of JEFF-3.3 nuclear data library with AMPX for its application within the SCALE package. The AMPX-formatted cross-section library has been widely verified and tested using a comprehensive set of criticality benchmarks from ICSBEP, by comparing both with results provided by other processing and neutron transport codes and experimental. Processing of JEFF-3.3 covariances is also addressed along with their corresponding verification using covariances processed with NJOY. This work paves the way towards a successful future interaction between JEFF libraries and SCALE.

Nuclear Data Sensitivity and Uncertainty Study for the Pressurized Water Reactor (PWR) Benchmark Using RMC and SCALE

In order to improve the safety and economy of nuclear reactors, it is necessary to analyze the sensitivity and uncertainty (S/U) of the nuclear data. The capabilities of S/U analysis has been developed in the Reactor Monte Carlo code RMC, using the iterated fission probability (IFP) method and the superhistory method. In this paper, the S/U capabilities of RMC are applied to a typical PWR benchmark B&W’s Core XI, and compared with the multigroup and continuous-energy S/U capabilities in the SCALE code system. The S/U results of the RMC-IFP method and the RMC-superhistory method are compared with TSUNAMI-CE/MG in SCALE. The sensitivity results and the uncertainty results of major nuclides that contribute a lot to the uncertainties in keff are in good agreement in both RMC and SCALE. The RMC-superhistory method has the same precision as the IFP method, but it reduces the memory footprint by more than 95% and only doubles the running time. The superhistory method has obvious advantages when there are many nuclides and reaction types to be analyzed. In addition, the total uncertainties in the keff of the first-order uncertainty quantification method are compared with the stochastic sampling method, and the maximum relative deviation of total uncertainties in the keff is 8.53%. Verification shows that the capabilities of S/U analysis developed in the RMC code has good accuracy.

Continuous-Energy ENDF/B-VIII.0 Cross Section and SCALE 6.2.4 Performance for Nuclear Criticality Safety Applications: 1H, C, 58,60Ni, 182,183,184,186W, 235,238U, 239Pu

As part of the nuclear data evaluation and validation cycle, the ENDF/B-VIII.0 cross-section library released in 2018 requires testing to determine areas of improvement and deterioration. Previous work by the authors investigated the performance of 16O, 56Fe, and 63,65Cu cross sections, with this study acting as an extension of the prior work. In addition to the isotopes and nuclear criticality safety benchmarks of interest to the prior work, benchmarks from the International Criticality Safety Benchmark Evaluation Project Handbook were selected for their keff sensitivity to 1H, C, 58,60Ni, 182,183,184,186W, 235,238U, or 239Pu cross sections and were modeled in the SCALE code system maintained by Oak Ridge National Laboratory. In total, 253 benchmark configurations were selected for their sensitivities and modeled using SCALE 6.2.4 Criticality Safety Analysis Sequences (CSAS) continuous-energy Monte Carlo keff calculations. This collection includes and expands upon the 99 benchmarks in the prior work. The AMPX-processed ENDF/B-VIII.0 library was decomposed into individual ENDF/B-VIII.0 datum libraries for each isotope of interest. Doing so allowed for the individual substitution of an ENDF/B-VIII.0 cross section in the place of ENDF/B-VII.1, determining isotope-specific effects of ENDF/B-VIII.0 relative to ENDF/B-VII.1. Full library calculations with entirely ENDF/B-VII.1 data or entirely ENDF/B-VIII.0 data were also executed. As a measure of performance, the average relative deviation was determined as the ratio of the deviation between calculated and experimental keff to the propagated calculational and experimental uncertainty. With calculated full library and isotope-specific ENDF/B-VIII.0 keff’s, an optimized combination of data libraries was estimated and confirmed with SCALE calculations. This showed that reverting 239Pu, 58Ni, 16O, and 65Cu cross sections to ENDF/B-VII.1 resulted in improved performance relative to the full ENDF/B-VIII.0 library. Across all 253 benchmarks, the average relative deviation was 1.29σ for the full ENDF/B-VII.1 library, 1.17σ for the full ENDF/B-VIII.0 library, and 0.97σ for the optimized combination. The reversion of 239Pu, 58Ni, 16O, and 65Cu cross sections to ENDF/B-VII.1 in the 99 benchmarks of the prior work resulted in further improved experimental agreement compared to the previously reported improvement from 16O and 65Cu alone. Therefore, it is suggested that applications with significant sensitivities to 239Pu, 58Ni, 16O, and 65Cu consider their choice of nuclear data library.

Nuclide Inventory Benchmark for BWR Spent Nuclear Fuel: Challenges in Evaluation of Modeling Data Assumptions and Uncertainties

This work discusses challenges and approaches to uncertainty analyses associated with the development of a nuclide inventory benchmark for fuel irradiated in a boiling water reactor. The benchmark under consideration is being developed based on experimental data from the SFCOMPO international database. The focus herein is on how to address missing data in fuel design and operating conditions that are important for adequately simulating the time-dependent changes in fuel during irradiation in the reactor. The effects of modeling assumptions and uncertainties in modeling parameters on the calculated nuclide inventory were analyzed and quantified through computational models developed using capabilities in the SCALE code system. Particular attention was given to the impact of the power history and water coolant density on the calculated nuclide inventory, as well as to the effect of geometry modeling considerations not usually addressed in a nuclide inventory benchmark. These considerations include gap closure, channel bow, and channel corner radius, which do not usually apply to regular reactor operation but are relevant for assessing impacts of potential anomalous operating scenarios.

Uncertainty quantification of light-water-moderated and light-water-reflected cores with highly-enriched uranium fuel at Kyoto University Critical Assembly

ABSTRACT A series of uncertainty quantifications are successfully conducted by the eigenvalue calculations, the uncertainty of criticality and reactivity, and the reduction of uncertainty using the KENO-VI code, the TSUNAMI-3D, the TSAR and the TSURFER modules of the SCALE code system, respectively. The eigenvalue calculations with KENO-VI provide noteworthy the discrepancy between experimental and numerical results of effective multiplication factor (keff), and the neutron spectrum dependence of keff in the light-water-moderated and light-water-reflected cores (C-cores) at the Kyoto University Critical Assembly (KUCA). In terms of TSUNAMI-3D, the discrepancy of keff is determined mainly by the uncertainty of uranium-235 (235U) fission spectrum (χ value) induced by the cross-section data, due to the use of highly-enriched uranium fuel in the KUCA C-cores. Also, the neutron spectrum dependence of keff is adequately understood from the effects of the uncertainty of hydrogen-1 capture, oxygen-16 elastic scattering, and aluminum-27 elastic and inelastic scattering reactions, in addition to the uncertainty of 235U fission spectrum. The reduction in the uncertainty of criticality and reactivity is interestingly achieved by the cross-section adjustment method provided in the TSURFER module, demonstrating the contribution of relevant nuclides and reactions of core components in the KUCA C-cores.

Validation of spent nuclear fuel decay heat calculations using Polaris, ORIGEN and CASMO5

Decay heat calculations of spent nuclear fuel (SNF) using Polaris and ORIGEN codes in the SCALE code system, and CASMO5 code, are validated using measurements from the Clab and GE-Morris facilities. Multiple hypothesis testing, relying on permutations and bootstrapping, is conducted to simultaneously analyze the significance of differences between calculations and measurements – namely the biases. The biases are small for the Clab data, and larger and have more variance for the GE-Morris data. Then, statistical tests are applied separately based on the measurement laboratory and the SNF origin. The calculations on the Clab PWRs show to be systematically different from the measurements. The remaining benchmarks show to be insignificantly different from the measurements. The benchmarks are useful to validate decay heat calculations of PWR and BWR fuel assemblies at various cooling times. Nevertheless, the choice of the data and the codes could affect the significance of the obtained biases.

Modeling of the Molten Salt Reactor Experiment with SCALE

A SCALE model was developed for the Molten Salt Reactor Experiment (MSRE) benchmark that was recently added to the International Handbook of Evaluated Reactor Physics Benchmark Experiments. This SCALE model served as a basis for criticality calculations and nuclear data sensitivity and uncertainty analyses with the Monte Carlo code Shift and the TSUNAMI computational capabilities in the SCALE code system. The focus of this work is the assessment of the impact of nuclear data on the calculated eigenvalue results in support of the discussion of differences between the calculated and the experimental eigenvalue result. The differences in the eigenvalues obtained using the ENDF/B-VII.0, ENDF/B-VII.1, and ENDF/B-VIII.0 nuclear data libraries cover a relatively small range of 230 pcm. Since eigenvalue sensitivity of the MSRE is dominated by the neutron multiplicity and neutron capture of 235U and elastic scattering in graphite, relevant changes in the ENDF/B libraries for nuclear reactions (such as carbon capture) that caused large differences in other graphite-moderated systems did not have a significant impact. Propagation of nuclear data uncertainty results in an eigenvalue uncertainty of pcm with the major contributors being U neutron multiplicity, graphite elastic scattering, and 7Li neutron capture. All calculations resulted in large differences of 2000 pcm in eigenvalue compared to the benchmark experimental value. Several potential contributors to this difference—including uncertainties and gaps in the knowledge of the material, geometry, and nuclear data—were identified. Simplified models of the full MSRE core were developed, and similarity assessments were conduced with the full MSRE core model. It was found that simplified models can serve as adequate surrogates of the full-core model such that they can be used for performing selected nuclear data performance assessments with a lower computational burden.

Nuclear Data Sensitivity Study for the EBR-II Fast Reactor Benchmark Using SCALE with ENDF/B-VII.1 and ENDF/B-VIII.0

The EBR-II benchmark, which was recently included in the International Handbook of Evaluated Reactor Physics Benchmark Experiments, served as a basis for assessing the performance of the SCALE code system for fast reactor analyses. A reference SCALE model was developed based on the benchmark specifications. Great agreement was observed between the eigenvalue calculated with this SCALE model and the benchmark eigenvalue. To identify potential gaps and uncertainties of nuclear data for the simulation of various quantities of interest in fast spectrum systems, sensitivity and uncertainty analyses were performed for the eigenvalue, reactivity effects, and the radial power profile of EBR-II using the two most recent ENDF/B nuclear data library releases. While the nominal results are consistent between the calculations with the different libraries, the uncertainties due to nuclear data vary significantly. The major driver of observed uncertainties is the uncertainty of the 235U (n,γ) reaction. Since the uncertainty of this reaction is significantly reduced in the ENDF/B-VIII.0 library compared to ENDF/B-VII.1, the obtained output uncertainties tend to be smaller in ENDF/B-VIII.0 calculations, although the decrease is partially compensated by increased uncertainties in 235U fission and ν¯.

Comparison of Irradiated TRISO Fuel Radioactivity from Multiple Advanced Reactor Designs

The SCALE code system was used to model, deplete, and compare several different tri-structural isotropic (TRISO)–fueled reactor designs: a helium-cooled prismatic reactor, a helium-cooled pebble-bed reactor (PBR), and a fluoride-lithium-beryllium (FLIBE) molten-salt-cooled PBR. The purpose of this comparison was to understand how differences in the reactor designs affect the radioactivity of the fuel after discharge and whether those differences are significant. First, the various reactor designs were built and depleted in the TRITON module for each design and fuel enrichment. Then, the TRITON outputs were used to create burnup-dependent reactor libraries. These libraries were then used by ORIGEN to determine the activities of discharged fuel for each reactor, which were compared to generic Westinghouse 17 × 17 fuel. Overall, the results showed that short-term activities are dominated by reactors with higher operating powers, and the reactor type, initial fuel enrichment, and maximum burnup are of only secondary importance. Although this analysis only focuses on activities in Becquerels, these dependencies are consistent with the expected behavior of decay heat. However, analysis of long-term time periods post irradiation shows that the reactor type and maximum burnup have strong impacts on the activities; initial fuel enrichment has a secondary impact while operating power is inconsequential. These results would be useful for analyses, such as dose assessment and modeling in postrelease scenarios, normal fuel handling operations, and spent fuel transport, storage, and disposal. Of particular interest, the results in this technical note show that analyses that focus on spent nuclear fuel of advanced reactors need to consider each parameter carefully. Unsurprisingly, if the correct operating power is not used in short-term analyses, the results will not be correct. Perhaps unexpectedly, however, if the correct reactor type is not used, then the long-term results will also be incorrect, especially for areas such as permanent disposal. Even though this technical note focuses on the total activity of nuclear fuel, it provides initial results on the effects of various input parameters and also provides a framework to extend the work into other analyses of spent fuel from advanced reactors, especially those employing TRISO fuel.

Extension of SCALE/Sampler’s sensitivity analysis

Nuclear data are a major source of uncertainties in reactor physics calculations. The propagation of nuclear data uncertainties to important system responses is instrumental when determining appropriate safety margins in reactor safety analyses. It is also important to understand the major contributors to the observed uncertainties to make recommendations for further measurements and evaluations and aid in the understanding of the studied system.The SCALE code system allows for nuclear data uncertainty analysis based on the random sampling approach as implemented in SCALE’s Sampler sequence. Sampler was recently extended by a sensitivity analysis in terms of the calculation of two correlation-based sensitivity indices. This analysis allows for the identification of the top contributing nuclear reactions to any analyzed output uncertainty. This paper presents the sensitivity indices, along with their interpretation and limitations. It demonstrates the application in an eigenvalue and decay heat analysis for a boiling water reactor fuel assembly.

Investigating Possibilities of fuelling NIRR-1 with Low Enriched Uranium Silicide-Aluminum Dispersion Fuels

In the last quarter of 2018, low enriched uranium dioxide fuel with zirconium alloy cladding was used successfully to convert the core of NIRR-1 from HEU to LEU fuel and the removed core returned to the country of origin. The objective of this study was to investigate the possibility of fueling the same system with alternate LEU fuel for future replacement of the current fuel, without any unacceptable compromise in reactor performance. Having more than one fuel options available for the same reactor system will present Nigeria an opportunity of making good economic decisions at the end of the cycle of the current LEU fuel. The performance of low enriched uranium silicide aluminum dispersion fuels in the core of NIRR-1 has been investigated and the results were identical with that of similar studies conducted elsewhere for generic MNSR system. Some of the calculated reactor parameters using this alternate LEU fuel were closely identical with that of the old HEU core. The computer software selected for this studies were the SCALE code system and the VENTURE PC. While the SCALE code system was employed to generate a properly averaged multigroup cross section library for the investigated LEU core models for NIRR-1 system, the VENTURE PC was utilized to give criticality information, few group fluxes and power density distributions within the core of the modeled system.Keywords— Reactor, Reactivity, Fuel, Enriched, Silicide

Uncertainty analyses of spent nuclear fuel decay heat calculations using SCALE modules

Decay heat residuals of spent nuclear fuel (SNF), i.e., the differences between calculations and measurements, were obtained previously for various spent fuel assemblies (SFA) using the Polaris module of the SCALE code system. In this paper, we compare decay heat residuals to their uncertainties, focusing on four PWRs and four BWRs. Uncertainties in nuclear data and model inputs are propagated stochastically through calculations using the SCALE/Sampler super-sequence. Total uncertainties could not explain the residuals of two SFAs measured at GE-Morris. The combined z-scores for all SFAs measured at the Clab facility could explain the resulting deviations. Nuclear-data-related uncertainties contribute more in the high burnup SFAs. Design and operational uncertainties tend to contribute more to the total uncertainties. Assembly burnup is a relevant variable as it correlates significantly with the SNF decay heat. Additionally, burnup uncertainty is a major contributor to decay heat uncertainty, and assumptions relating to these uncertainties are crucial. Propagation of nuclear data and design and operational uncertainties shows that the analyzed assemblies respond similarly with high correlation. The calculated decay heats are highly correlated in the PWRs and BWRs, whereas lower correlations were observed between decay heats of SFAs that differ in their burnups.

Neutron Transport Simulations of RBMK Fuel Assembly Using Multigroup and Continuous Energy Data Libraries within the SCALE Code

The neutron transport simulations of RBMK-1500 fuel assembly were performed using both multigroup and continuous energy data libraries available within the SCALE code system in order to validate its suitability for the estimation of RBMK neutronic characteristics. The resonance processing of cross section, involved in the preparation of the multigroup data library, has a significant impact on neutron transport calculations. Standard Dancoff factors (DFs) used for the heterogeneous geometry of RBMK fuel assembly are insufficient for the accurate estimation of resonance self-shielding. Thus, the SCALE module MCDancoff was used in this study to determine location-specific DFs. The results of RBMK-1500 fuel assembly simulations using standard and user-defined DFs were compared. In addition, the continuous energy (CE) cross-section data library was applied for the benchmark calculations. The impact of different nuclear data libraries on neutron transport simulations was tested as well. It was found out that the usage of the multigroup data libraries generates some deviation from the reference simulations obtained with CE libraries. The CE library based on the estimated ENDF/B-VII.1 data proved to be the best alternative for neutron transport simulations of RBMK fuel assembly.

Skyshine Calculations for a Large Spent Nuclear Fuel Storage Facility with SCALE 6.2.3

The SCALE code system developed at Oak Ridge National Laboratory includes state-of-the-art capabilities for radiation source term and radiation transport simulations that can be used in numerous applications, including dose rate analyses of complex consolidated interim storage facilities (CISFs). A licensed CISF could be used to store tens of thousands of tonnes of spent nuclear fuel discharged from commercial power reactors using various cask and storage pad designs. A CISF design must comply with the regulatory requirements provided in 10 CFR Part 72, including requirements related to annual dose limits applicable to real individuals located beyond the area controlled by the licensee. Therefore, calculating a dose to the public is a necessary part of the licensing process for the construction of a CISF. These calculations are very challenging because of the complexity of the CISF design and the low magnitude of dose rate at large distances from the facility. This paper describes detailed far-field dose rate calculations performed for a proposed CISF using MAVRIC, the Monte Carlo radiation shielding sequence in SCALE 6.2.3, with automated variance reduction based on discrete ordinates calculations. The method presented in this paper uses a detailed Monte Carlo radiation transport simulation in one step from source to dose rate. A series of independent simulations was made using the complete site geometry (all casks present), but with only one cask containing radiation sources to obtain the dose rate maps produced by each storage cask. The CISF dose rate map was obtained by adding the dose rate maps produced by the independent individual cask simulations. Ample volumes of air and soil extending beyond the location of interest for dose rate calculation were included in the calculation model to properly simulate important radiation attenuation and scattering events that affect far-field dose rates. A comprehensive sensitivity study is included in this paper to illustrate the importance of selecting appropriate air volume, mass density, and composition for CISF skyshine dose rate calculations. Dry soil and soil containing water were analyzed to determine their effects on groundshine radiation.