Hakim Ferroukhi

PSI Methodologies for Nuclear Data Uncertainty Propagation with CASMO-5M and MCNPX: Results for OECD/NEA UAM Benchmark Phase I

Capabilities for uncertainty quantification (UQ) with respect to nuclear data have been developed at PSI in the recent years and applied to the UAM benchmark. The guiding principle for the PSI UQ development has been to implement nonintrusive “black box” UQ techniques in state-of-the-art, production-quality codes used already for routine analyses. Two complimentary UQ techniques have been developed thus far: (i) direct perturbation (DP) and (ii) stochastic sampling (SS). The DP technique is, first and foremost, a robust and versatile sensitivity coefficient calculation, applicable to all types of input and output. Using standard uncertainty propagation, the sensitivity coefficients are folded with variance/covariance matrices (VCMs) leading to a local first-order UQ method. The complementary SS technique samples uncertain inputs according to their joint probability distributions and provides a global, all-order UQ method. This paper describes both DP and SS implemented in the lattice physics code CASMO-5MX (a special PSI-modified version of CASMO-5M) and a preliminary SS technique implemented in MCNPX, routinely used in criticality safety and fluence analyses. Results are presented for the UAM benchmark exercises I-1 (cell) and I-2 (assembly).

Advanced calculation methodology for manufacturing and technological parameters" uncertainties propagation at arbitrary level of lattice elements grouping

A new methodology, referred to as manufacturing and technological parameters uncertainty quantification (MTUQ), is under development at Paul Scherrer Institut (PSI). Based on uncertainty and global sensitivity analysis methods, MTUQ aims at advancing state-of-the-art for the treatment of geometrical/material uncertainties in light water reactor computations, using the MCNPX Monte Carlo neutron transport code. The development is currently focused primarily on criticality safety evaluations (CSE). In that context, the key components are a dedicated modular interface with the MCNPX code and a user-friendly interface to model functional relationship between system variables. A unique feature is an automatic capability to parameterize variables belonging to so-called “repeated structures” such as to allow for perturbations of each individual element of a given system modelled with MCNPX. Concerning the statistical analysis capabilities, these are currently implemented through an interface with the ROOT platform to handle the random sampling design. This paper presents the current status of the MTUQ methodology development and a first assessment of an ongoing organisation for economic cooperation and development/nuclear energy agency benchmark dedicated to uncertainty analyses for CSE. The presented results illustrate the overall capabilities of MTUQ and underline its relevance in predicting more realistic results compared to a methodology previously applied at PSI for this particular benchmark.

Testing the Sampling-Based NUSS-RF Tool for the Nuclear Data—Related Global Sensitivity Analysis with Monte Carlo Neutronics Calculations

Abstract NUSS-RF is a tool for nuclear data uncertainty propagation through neutronics calculations with continuous-energy Monte Carlo codes and ACE-formatted nuclear data libraries. Many existing codes, including the original version of NUSS (Nuclear data Uncertainty Stochastic Sampling), are based on simple random sampling algorithms. The NUSS-RF extension now uses a frequency-based sampling algorithm, called the random balance design (RBD), to analyze individual nuclear data uncertainty contributions in regard to the total output (e.g., keff) uncertainty. The implementation of the RBD method into NUSS-RF is initially verified by comparing the computed individual input variance contributions with analytical solutions for two analytical test cases. As well, it is assessed against the alternative approach based on the use of correlation coefficients. NUSS-RF is then used for an analysis of the Jezebel and Godiva fast-spectrum criticality benchmarks: in a first step, the overall effect of the 239Pu(n,f) and 235U(n,f) cross-section uncertainties on keff is evaluated, while in a second step, the contributions from the individual energy groups are quantified. As an additional verification, the NUSS-RF results are assessed against sensitivity and uncertainty analysis based on perturbation theory, showing good agreement between the two solutions. Finally, the capability of NUSS-RF is demonstrated for ranking the input parameters with respect to their influence on the total uncertainty of the output parameters, taking into account possible correlations between input parameters. Possible future improvements for the current computational scheme are discussed in the conclusions.

Peach bottom BWR Turbine Trip Benchmark analyses with RETRAN-3D and CORETRAN

Abstract This paper presents results from Paul Scherrer Institut (PSI) on the three phases of the Peach Bottom Boiling Water Reactor Turbine Trip Benchmark. The first part of the paper presents the PSI analysis using RETRAN-3D of Phase 1, where the system pressure is predicted based on a predefined core power distribution. These calculations elucidate the importance of accurate modeling of the steam separator region and of nonequilibrium effects. In the second part, the CORETRAN results of Phase 2 are summarized and the core 3-D response to the pressurization transient prior to SCRAM is discussed. The CORETRAN results show a slight axial flux redistribution toward the top of the core, while radially a flux redistribution is observed toward core regions with assemblies that are initially moderately voided and where the axial power shape is increasingly top-peaked. The impact of the control rod configuration as well as the assembly coolant inventory dynamics on the 3-D flux redistribution is also discussed. The third part presents results of the Phase 3 calculation using RETRAN-3D, which is a culmination of the analytical work of Phases 1 and 2.

NUSS-RF: stochastic sampling-based tool for nuclear data sensitivity and uncertainty quantification

The “blackbox” approach of stochastic sampling (SS) methods for simultaneous nuclear data uncertainty quantification is powerful except it reveals little of the individual uncertainty contributions. In this work, the SS-based tool “NUSS” (nuclear data uncertainty stochastic sampling) developed at PSI is updated to “NUSS- RF” which estimates individual nuclear data uncertainty contributions to the total output uncertainty. The new capability is based on the Random balance design and Fourier amplitude sensitivity testing methods, both belonging to the so-called global sensitivity analysis. First, the implementation of NUSS-RF is tested using a mathematical function, followed by the sensitivity and uncertainty analysis for 235U(n,γ) and 238U(n,γ) cross sections in Godiva and BWR pincell benchmarks, respectively. The results are compared to the deterministic sensitivity/uncertainty “Sandwich Rule” approach which is local. For uncorrelated inputs, both methods have the equivalent interpretation of the input uncertainty contribution (in terms of variance fraction and sensitivity index), hence producing good agreement in the results. For correlated inputs, the discrepancy between the two methods broadens with the extent of the correlations.

Transition to CASMO-5M and SIMULATE-3K for Stability Analyses of the Swiss Boiling Water Reactors

For stability analyses of the Swiss operating boiling water reactors, the methodology employed and validated so far at the Paul Scherrer Institute (PSI) was based on the RAMONA-3 code with a hybrid upstream static lattice/core analysis approach using CASMO-4 and PRESTO-2. More recently, steps were undertaken toward a new methodology based on the SIMULATE-3K (S3K) code for the dynamical analyses combined with the CMSYS system, which relies on the CASMO/SIMULATE-3 suite of codes and was established at PSI to serve as framework for the development and validation of reference core models of all the Swiss reactors and operated cycles. This paper presents a first validation of the new methodology on the basis of a benchmark recently organized by a Swiss utility and including the participation of several international organizations with various codes/methods. Now in parallel, a transition from CASMO-4E (C4E) to CASMO-5M (C5M) as basis for the CMSYS core models was also recently initiated at PSI. Consequently, it was considered adequate to address the impact of this transition both for the steady-state core analyses as well as for the stability calculations and to achieve thereby an integral approach for the validation of the new S3K methodology. Therefore, a comparative assessment of C4 versus C5M is also presented in this paper, with particular emphasis on the void coefficients and their impact on the downstream stability analysis results.

Towards a Consolidated Approach for the Assessment of Evaluation Models of Nuclear Power Reactors

Abstract The STARS project at the Paul Scherrer Institut (PSI) has adopted the TRACE thermal-hydraulic code. For analyses involving interactions between system and core, a coupling of TRACE with the SIMULATE-3K (S3K) light water reactor (LWR) core simulator has been developed. In this configuration, the codes and associated simulation models play a central role to achieve a comprehensive safety analysis capability. Therefore, efforts have now been undertaken to consolidate the validation strategy by implementing a more rigorous and structured assessment approach for TRACE applications. The principle is to systematically track the evolution of a given set of predicted physical quantities of interest (QoIs) over a multidimensional parametric space. If properly set up, such environment should provide code developers and code users with persistent (less affected by user effect) and quantified information (sensitivity of QoIs) on the applicability of a simulation scheme (codes, methodology, and input models) for steady-state and transient analysis of full LWR systems. Through this, for each given transient/accident, critical paths of the validation process can be identified that could then translate into defining reference schemes to be applied for downstream predictive simulations. To illustrate this approach, this validation strategy is applied to an inadvertent blowdown event that occurred in a Swiss BWR/6. The transient was initiated by the spurious actuation of the automatic depressurization system. Here, the validation approach progresses through a number of dimensions: (a) different versions of the TRACE code; (b) the methodology dimension—in this case imposed power and updated TRACE core models are investigated; and (c) the nodalization dimension, where changes to the input model are assessed. For each step in each validation dimension, a common set of QoIs is investigated. For the steady-state results, these include fuel temperature distributions. For the transient part of the present study, the evaluated QoIs include the system pressure evolution and water carryover into the steam line. It has been seen that the improvements to the model predictions resulted in a small impact on the system pressure gradient, thus confirming a persistency of the downstream mechanical stress estimate, whereas the water carryover could vary by up to 150% as a function of the adopted simulation methodology.