SCALE Code System Research Articles

Sensitivity and Uncertainty Analysis of Commercial Reactor Criticals for Burnup Credit

This paper provides insights into the neutronic similarities between a representative high-capacity rail-transport cask containing typical pressurized water reactor (PWR) spent nuclear fuel assemblies and critical reactor state-points, referred to as commercial reactor critical (CRC) state-points. Forty CRC state-points from five PWRs were analyzed, and the characteristics of CRC state-points that may be applicable for validation of burnup-credit criticality safety calculations for spent fuel transport/storage/disposal systems were identified. The study employed cross-section sensitivity and uncertainty analysis methods developed at Oak Ridge National Laboratory and the TSUNAMI set of tools in the SCALE code system as a means to investigate neutronic similarity on an integral and nuclide-reaction-specific level. The results indicate that except for the fresh-fuel-core configuration, all analyzed CRC state-points are either highly similar, similar, or marginally similar to the representative high-capacity cask containing spent nuclear fuel assemblies with burnups ranging from 10 to 60 GWd/tU in terms of their shared uncertainty in keff due to cross-section uncertainties. On a nuclide-reaction-specific level, the CRC state-points provide significant coverage, in terms of neutronic similarity, for most of the actinides and fission products relevant to burnup credit. Hence, in principle, the evaluated CRC state-points could serve as part of a set of benchmark experiments for determining a bias and bias uncertainty to be applied to the calculated keff of a spent fuel transport/storage/disposal system to correct for approximations in computational methods and errors and uncertainties in nuclear data. Note, however, that an evaluation to quantify the uncertainties associated with various CRC modeling parameters (e.g., fuel isotopic compositions, physical characteristics of reactor core components, and reactor operating history information), which has relevance to the use of these critical configurations for bias determination, was not performed as part of this study.

Recent Use of Covariance Data for Criticality Safety Assessment

The TSUNAMI codes of the Oak Ridge National Laboratory SCALE code system were applied to a burnup credit application to demonstrate the use of sensitivity and uncertainty analysis with recent cross section covariance data for criticality safety code and data validation. The use of sensitivity and uncertainty analysis provides for the assessment of a defensible computational bias, bias uncertainty, and gap analysis for a complex system that otherwise could be assessed only through the use of expert judgment and conservative assumptions.

SCALE analysis of CLAB decay heat measurements for LWR spent fuel assemblies

A set of decay heat measurements for spent fuel assemblies recently carried out at the Swedish central interim storage facility for spent fuel, CLAB, was analyzed with the SCALE code system. The measurements include a variety of light water reactor assemblies that cover a large burnup range – up to 51 GWd/MTU – and a cooling time domain of interest to spent fuel storage and transportation applications. The results of the analysis show a good agreement between measured and predicted decay heat, with the calculated decay heat in general within the range of the uncertainty of the measured value. The effect of various assembly data on the calculated decay heat is analyzed and discussed. Uncertainties that may arise from various approaches and assumptions in the computational model are identified and examined.

Comparison of dose rate calculations with measurements at CASTOR HAW 20/28 CG cask

AbstractThis paper presents first results of a comparison of dose rates calculated by SCALE 4.4a with measured dose rates at a CASTOR HAW 20/28 CG cask. This type of CASTOR cask is used for the transport and interim storage of vitrified high active waste (HAW) from reprocessing. The calculations were done using the SCALE code system which was developed by Oak Ridge National Laboratory (ORNL). The measurements were carried out in January 2003. The results show that the SCALE code system is a suitable tool for shielding assessment tasks to verify compliance with dose rate limits. However, this comparison shows that additional work is necessary to obtain a better correlation between calculation and measurement.

Application of the Pitzer Method for Modeling Densities of Actinide Solutions in the SCALE Code System

A new approach to modeling fissile solution densities is developed and applied to arbitrary solutions of UO22+, Th4+, Pu4+, H+, and NO3- in water. The model is based on the semiempirical method of Pitzer for describing electrolyte solution thermodynamics. This new density model has been included in the recent release of the SCALE 5 code system as the default for criticality calculations of fissile solutions.

Use of Sensitivity and Uncertainty Analysis to Select Benchmark Experiments for the Validation of Computer Codes and Data

Sensitivity and uncertainty analysis methodologies under development at Oak Ridge National Laboratory were applied to determine whether existing benchmark experiments adequately cover the area of applicability for the criticality code and data validation of PuO2 and mixed-oxide (MOX) powder systems. The study examined three PuO2 powder systems and four MOX powder systems that would be useful for establishing mass limits for a MOX fuel fabrication facility. Using traditional methods to choose experiments for criticality analysis validation, 46 benchmark critical experiments were identified as applicable to the PuO2 powder systems. However, only 14 experiments were thought to be within the area of applicability for dry MOX powder systems.The applicability of 318 benchmark critical experiments, including the 60 experiments initially identified, was assessed. Each benchmark and powder system was analyzed using the Tools for Sensitivity and UNcertainty Analysis Methodology Implementation (TSUNAMI) one-dimensional (TSUNAMI-1D) or TSUNAMI three-dimensional (TSUNAMI-3D) sensitivity analysis sequences, which will be included in the next release of the SCALE code system. This sensitivity data and cross-section uncertainty data were then processed with TSUNAMI-IP to determine the correlation of each application to each experiment in the benchmarking set. Correlation coefficients are used to assess the similarity between systems and determine the applicability of one system for the code and data validation of another.The applicability of most of the experiments identified using traditional methods was confirmed by the TSUNAMI analysis. In addition, some PuO2 and MOX powder systems were determined to be within the area of applicability of several other benchmarks that would not have been considered using traditional methods. Therefore, the number of benchmark experiments useful for the validation of these systems exceeds the number previously expected. The TSUNAMI analysis also emphasized some areas where more benchmark data are needed, indicating the need for further evaluation of existing experiments, or possibly the completion of new experiments to fill these gaps. This lack of evaluated data is particularly important for very dry and dense MOX powder systems.

Calculation of the void reactivity of CANDU lattices using the scale code system

The reactivity effect of coolant voiding in CANDU-type fuel lattices has been calculated with different methods using the scale code system. The known positive void reactivity coefficient of the original lattice was correctly obtained. A modified fuel bundle containing dysprosium and slightly enriched uranium to eliminate the positive reactivity effect was also calculated. Owing to the increased heterogeneity of this modified fuel the one-dimensional cylindrical calculation with XSDRN proved to be inadequate. Code options allowing bundle geometry were successfully used for the calculation of the strongly space dependent flux and spectrum changes which determine the void reactivity.

Validation of SCALE-4 for Burnup Credit Applications

In the past, criticality analysis of pressurized water reactor (PWR) fuel stored in racks and casks has assumed that the fuel is fresh with the maximum allowable initial enrichment. If credit is allowed for fuel burnup in the design of casks that are used in the transport of spent light water reactor fuel to a repository, the increase in payload can lead to a significant reduction in the cost of transport and a potential reduction in the risk to the public. A portion of the work has been performed at Oak Ridge National Laboratory (ORNL) in support of the U.S. Department of Energy (DOE) efforts to demonstrate a validation approach of criticality safety methods to be used in burnup credit cask design. The date, the SCALE code system developed at ORNL has been the primary computational tool used by DOE to investigate technical issues related to burnup credit. The SCALE code package is a well-established code system that has been widely used in away from reactor applications. Criticality safety analyses are performed via the criticality safety analysis sequences (CSAS) and spent-fuel characterization via the shielding analysis sequence (QSAS) and spent-fuel characterization via the shielding analysis sequence (SAS2H). The SCALE 27-group burnup library containing ENDF/B-IV (actinides) and ENDF/B-V (fission products) data has been used for all calculations. The American National Standards Institute/American Nuclear Society (ANSI/ANS)-8.1 criticality safety standard requires validation and benchmarking of the calculational methods used in evaluating criticality safety limits for applications outside reactors of correlation against critical experiments that are applicable. Numerous critical experiments for fresh PWR-type fuel in storage and transport configurations exist and can be used as part of a validation database. However, there are no critical experiments with burned PWR-type fuel in storage and transport configurations. As an alternative, commercial reactors offer an excellent source of measured critical configurations. Part of the work that has been performed to date to validate the SCALE-4 code system for burnup credit applications using measured critical configurations includes: 1. fresh fuel critical experiments having geometric and nuclear characteristics similar to PWR spent fuel in storage and transport configurations 2. commercial PWR hot-zero-power and hot-full-power reactor critical configurations. The ability to closely predict reactor critical conditions is important in the validation of a methodology for spent-fuel applications because input data are determined based on relatively little detail of reactor core operation. Such limited information is expected to be representative of data available when burnup credit calculations are being performed in the determination of optimum cask loadings. The results reported demonstrate the ability of the ORNL SCALE-4 methodology to predict a value of k eff very close to the known value of 1.0, both for fresh fuel criticals and for the more complex reactor criticals. Beyond these results, additional work in the determination of biases and uncertainties is necessary prior to use in burnup credit applications

モンテカルロ法による表面線量当量率分布の計算手法

A technique for Monte Carlo codes to obtain surface dose rate profiles of cask surfaces within a reasonable CPU time is presented in this paper. This technique does not use a huge number of actual small detectors defined by additional geometry input data, but uses imaginary detectors. To demonstrate this technique, SAS4 in the SCALE code system is chosen. The estimator is a surface crossing estimator (SXE). The technique is very simple and easy. In the Monte Carlo calculation, the contribution to the dose by a particle across the surface estimator is stored separately according to the coordinate of the surface crossing point of the particle as if there were a series of small surface estimators. After the Monte Carlo calculation has been done, the post processing procedure is made to produce dose rate profiles at the cask surface using the data from a series of small surface estimators. The additional CPU time required for introducing the new technique to the original CPU time is very small. The proposed technique using SXE is confirmed to be a very efficient and cost-saving method for getting dose rate profiles around a cask.

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Summary of session

In this paper a newly developed thermal hydraulics (TH) code coupled with the state-of-the-art neutronics deterministic code AGENT (Arbitrary Geometry Neutron Transport) is presented. This coupled system is called the AGENT-TH scheme. The neutronics AGENT code uniquely combines the method of characteristics (MOC) and R-function geometrical modeling to solve the 2D/3D neutron transport in nuclear reactors of current or future designs. Commonly, the cross sections are prepared using the SCALE code system. The implementation of the TH code with AGENT enables coupled thermal-hydraulic and neutronics analysis in providing the average reactor fuel temperatures. This allows for an accurate overall power profiling within the reactor system, which affects reactor coolant temperature, neutron flux, and other associated reactor parameters.The two main types of LWRs are boiling water reactors (BWRs) and pressurized water reactors (PWRs). The BWRs operate at lower pressures than PWRs, which results in a higher void fraction, or fraction of coolant flow which is vapor versus liquid. Performances of PWRs, with a higher operating pressure, may be obtained using the homogeneous equilibrium mixture (HEM) model, which assumes a slip relation equal to one between the coolant liquid and vapor phases. To account for the higher void fraction in the BWR coolant, the BWRs properties are assessed based on the drift flux model. The thermal hydraulics drift flux model is developed as a part of the TH code, to address the two-phase flow associated with two-phase heat transfer (including subcooled nucleation and saturated boiling) as found in BWR systems. This is accomplished based on the drift flux equations in 1D steady-state conditions, which permits fast computation times without sacrificing final accuracy.The TH code is benchmarked against the reactor simulation code TRACE for a single pin to assess the thermal-hydraulics capabilities of the drift flux model for the fluid void fraction, flow quality, pressure, saturation temperature, and mixture temperature along the axial height of the pin. The drift flux thermal hydraulics module is coupled with AGENT using a radial heat transfer model taking into account heat transfer from the fuel centerline to the bulk fluid. Fuel temperatures calculated in the TH code are used to update cross sections from the SCALE code system to account for thermal feedback. Reasonable agreements are obtained between the AGENT-TH and TRACE for the presented benchmark examples.