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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.

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¯.

The CASL reactor simulation package VERA has been adapted to provide high-fidelity simulation capabilities for modeling source range detector response during subcritical reactor configurations. New...

Delayed neutrons, which are described by kinetic parameters, are significant for nuclear reactor operation as they make nuclear reactors controllable. Modern lattice physics codes such as TRITON and Polaris in SCALE code system, generate kinetic parameters for a given material composition. The calculation is performed by summation of the delayed neutron data for each precursor isotope and then weighting is performed by real and/or adjoint neutron spectrum. Quantifying uncertainties in the weighted kinetic parameters is important for assembly and core calculations. Understanding uncertainty in modeling scenarios involve kinetic parameters (e.g. transients), requires propagating uncertainty in the weighted kinetic parameters due to uncertainties in the fundamental nuclear data libraries, including delayed neutron data. In this work, uncertainty analysis of the homogenized (also called weighted or macroscopic) kinetic parameters has been performed using Sampler, a module in SCALE code system, to investigate the effect of fundamental nuclear and delayed neutron data uncertainties on the weighted kinetic parameters. Two major sources of uncertainties were considered: (1) fundamental nuclear data (i.e. cross-sections, fission yield, decay data) and (2) nuclide-dependent group-wise delayed neutron data based on reported experimental measurements. In this study, a new capability was developed through SCALE code system to allow propagation of delayed neutron data uncertainties. Preliminary analysis demonstrated 7% uncertainty in β eff at Beginning of Life (BOL) and increased to 15% after fuel burnup. Decay constant groups showed lower uncertainty than delayed neutron fraction groups, both at BOL and End of Life (EOL). Delayed neutron fraction responses showed high correlation to each other which is expected to be due to the cross-section covariances reported in SCALE data libraries as well as the normalization condition of the nuclide-dependent DNF. Different sources of U-235 delayed neutron data were compared, and the results showed that the uncertainty calculated by Tuttle data was bounded by other sources. The current study can be extended to calculate kinetic parameters and their uncertainties for more advanced LWR applications.

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.

A new depletion capability has been developed in the Exnihilo radiation transport code suite. This capability enables massively parallel domain-decomposed coupling between the Shift continuous-energy Monte Carlo solver and the nuclide depletion solvers in ORIGEN to perform high-performance Monte Carlo depletion calculations. This paper describes this new depletion capability and discusses its various features, including a multi-level parallel decomposition, high-order transport-depletion coupling, and energy-integrated power renormalization. Several test problems are presented to validate the new capability against other Monte Carlo depletion codes, and the parallel performance of the new capability is analyzed.

This paper describes the methodology developed and implemented in the Virtual Environment for Reactor Applications Core Simulator (VERA-CS) to perform high-fidelity, pressurized water reactor (PWR), multicycle, core physics calculations. Depletion of the core with pin-resolved power and nuclide detail is a significant advance in the state of the art for reactor analysis, providing the level of detail necessary to address the problems of the U.S. Department of Energy Nuclear Reactor Simulation Hub, the Consortium for Advanced Simulation of Light Water Reactors (CASL). VERA-CS has three main components: the neutronics solver MPACT, the thermal-hydraulic (T-H) solver COBRA-TF (CTF), and the nuclide transmutation solver ORIGEN. This paper focuses on MPACT and provides an overview of the resonance self-shielding methods, macroscopic-cross-section calculation, two-dimensional/one-dimensional (2-D/1-D) transport, nuclide depletion, T-H feedback, and other supporting methods representing a minimal set of the capa...

The discharge rod internal pressure (RIP) and cladding hoop stress (CHS) distributions are quantified for Watts Bar Nuclear Unit 1 (WBN1) fuel rods by modeling core cycle design data, operation dat...

Polaris is a 2-dimensional multigroup lattice physics capability in the SCALE code system for the analysis of light water reactor fuel designs. The goal of light water reactor lattice physics codes is to generate few-group homogenized cross sections for downstream full-core nodal diffusion calculations. Lattice physics calculations contain three primary components: the cross section processing calculation, the 2D transport calculation, and the depletion calculation. This paper summarizes the calculational methods and their implementation into Polaris, with an emphasis on implementation of the embedded self-shielding method. The accuracy of the embedded self-shielding method depends on the procedure used to generate self-shielding factors on the multigroup library. Numerical benchmarks calculations reveal that the accuracy of Polaris eigenvalue predictions is enhanced by (1) using heterogeneous unit cell models to generate the self-shielding factors on the library and (2) using self-shielding factors for within-group scattering cross section.

To more accurately represent the desired power in a constant power burnup calculation, the depletion steps of the calculation can be divided into substeps and the neutron flux renormalized on each substep to match the desired power. This paper explores how such renormalization should be performed, how large a difference it makes, and whether using renormalization affects results regarding the relative performance of different neutronics–depletion coupling schemes. When used with older coupling schemes, renormalization can provide a considerable improvement in overall accuracy. With previously published higher order coupling schemes, which are more accurate to begin with, renormalization has a much smaller effect. While renormalization narrows the differences in the accuracies of different coupling schemes, their order of accuracy is not affected.

The SCALE code system’s ability to address stochastic distributions of fuel particles within a graphite matrix has been revisited in both multigroup (MG) features and continuous-energy (CE) Monte Carlo methods. This paper describes current and emergent SCALE capabilities within the CSAS sequence to address double-heterogeneous systems and presents verification and validation studies of these methods and data. Good agreement was obtained for a high temperature gas-cooled reactor (HTGR) fuel pebble model between CSAS MG eigenvalue calculations and corresponding CE reference solutions. Code-to-code comparisons for this HTGR pebble model showed good agreement of CSAS-KENO and CSAS-Shift CE calculations and the Serpent and MCNP codes in terms of eigenvalues and reaction rate ratios. Validation studies based on two HTGR experiments resulted in good agreement between MG and CE results, as well as between experiment and calculation, although the level of agreement was significantly influenced by the applied ENDF/B nuclear data library.

We have developed a formulation for oxygen transport in uranium dioxide nuclear fuel that accounts for the effects of irradiation. The overall simulation combines the evolving isotopic composition, thermochemistry, and oxygen transport in irradiated fuel. The driving forces for oxygen transport are computed from local thermodynamic equilibrium calculations and account for the effects of temperature gradients and composition, including fission products. The proposed method provides a mechanism for including complex thermodynamic models of nuclear fuel in modeling of mass redistribution, and alleviates difficulties associated with the common thermodiffusion formulation. The transport model has been implemented within the nuclear fuel performance code BISON utilizing the thermochemistry code Thermochimica, with burnup calculations provided by the ORIGEN isotopic transmutation code.

A method for including external feed with polynomial time dependence in depletion calculations with the Chebyshev Rational Approximation Method (CRAM) is presented and the implementation of CRAM to the ORIGEN module of the SCALE suite is described. In addition to being able to handle time-dependent feed rates, the new solver also adds the capability to perform adjoint calculations. Results obtained with the new CRAM solver and the original depletion solver of ORIGEN are compared to high precision reference calculations, which shows the new solver to be orders of magnitude more accurate. Furthermore, in most cases, the new solver is up to several times faster due to not requiring similar substepping as the original one.

We present a quasidiffusion (QD) method for solving neutral particle transport problems in Cartesian XY geometry on unstructured quadrilateral meshes, including local refinement capability. Neutral particle transport problems are central to many applications including nuclear reactor design, radiation safety, astrophysics, medical imaging, radiotherapy, nuclear fuel transport/storage, shielding design, and oil well-logging. The primary development is a new discretization of the low-order QD (LOQD) equations based on cell-local finite differences. The accuracy of the LOQD equations depends on proper calculation of special non-linear QD (Eddington) factors from a transport solution. In order to completely define the new QD method, a proper discretization of the transport problem is also presented. The transport equation is discretized by a conservative method of short characteristics with a novel linear approximation of the scattering source term and monotonic, parabolic representation of the angular flux on incoming faces. Analytic and numerical tests are used to test the accuracy and spatial convergence of the non-linear method. All tests exhibit O(h^2) convergence of the scalar flux on orthogonal, random, and multi-level meshes.

Nuclear data uncertainty propagation based on stochastic sampling ( SS) is becoming more attractive while leveraging modern computer power. Two variants of the SS approach are compared in this paper. The Total Monte Carlo (TMC) method by the Nuclear Research and Consultancy Group (NRG) generates perturbed ENDF-6-formatted nuclear data by varying nuclear reaction model parameters. At Paul Scherrer Institute (PSI) the Nuclear data Uncertainty Stochastic Sampling (NUSS) system generates perturbed ACE-formatted nuclear data files by applying multigroup nuclear data covariances onto pointwise ACE-formatted nuclear data. Uncertainties of Pu-239 and U-235 from ENDF/B-VII.1, ZZ-SCALE6/COVA-44G and TENDL covariance libraries are considered in NUSS and propagated in MCNPX calculations for well-studied Jezebel and Godiva fast spectrum critical benchmarks. The corresponding uncertainty results obtained by TMC are compared with NUSS results and the deterministic Sensitivity/Uncertainty method of TSUNAMI-3D from SCALE6 package is also applied to serve as a separate verification. The discrepancies in the propagated Pu-239 and U-235 uncertainties due to method and covariance differences are discussed.

The paper presents recent activities conducted at the Paul Scherrer Institut (PSI) in relation to the development and validation of an integral calculation methodology based on CASMO-4/SIMULATE-3/MCNPX for accurate estimations of the fast neutron fluence (FNF) accumulated on reactor pressure vessels and internals of the operating Swiss BWRs. With this computational scheme, the default neutron source is set up at the pin-by-pin level with realistic spectrum specifications based on the actual reactor cycle-specific data from validated reference CASMO-4/SIMULATE-3 core analysis models. On this basis, MCNPX models are then applied for optimized calculations of the fast neutron flux at the RPV or at any other location of interest including e.g. at surveillance dosimeters. In that framework, the validation studies conducted so far have included one dosimeter set irradiated in a BWR/6 reactor during two relatively recent operating cycles. Although this first analysis revealed a satisfactory performance when comparing the calculation results to measured data, it was considered necessary to proceed with further sensitivity/optimization studies combined with an enlarged validation basis (i.e. using additional dosimeter sets) in order to strengthen the overall confidence in the scheme both at the qualitative and quantitative level. A summary of the recent progress achieved in these directions is presented in this paper. To start, recalling that BWRs are characterized by very complex and heterogeneous fuel assembly and core designs (e.g. pins with different enrichments and burnable absorber loading, partial length rods, fuel assemblies of different types in the core), the impact of such heterogeneities on FNF estimations is under investigation in order to determine the level of modeling details required for accurate computational schemes to be used for long-term evaluations of modern BWR core designs. Next, additional validation studies based on experimental dosimeter data obtained from the same BWR/6 reactor are presented. These enlarged validation studies involve the analysis of four dosimeter sets, each irradiated during one cycle (including the 3 first reactor operation cycles), and subsequently analyzed at the PSI Hot Lab shortly after the dosimeters extraction. All these additional validation studies are conducted using both the JEFF-3.1.1 and the ENDF/B-VII.0 continuous-energy neutron data libraries in order to assess the sensitivity of the PSI BWR computational scheme also upon the employed nuclear data.

SCALE’s general depletion, activation, and spent fuel source terms analysis capabilities are enabled through a family of modules related to the main ORIGEN depletion/irradiation/decay solver. The nuclide tracking in ORIGEN is based on the principle of explicitly modeling all available nuclides and transitions in the current fundamental nuclear data for decay and neutron-induced transmutation and relies on fundamental cross section and decay data in ENDF/B VII. Cross section data for materials and reaction processes not available in ENDF/B-VII are obtained from the JEFF-3.0/A special purpose European activation library containing 774 materials and 23 reaction channels with 12,617 neutron-induced reactions below 20 MeV. Resonance cross section corrections in the resolved and unresolved range are performed using a continuous-energy treatment by data modules in SCALE. All nuclear decay data, fission product yields, and gamma-ray emission data are developed from ENDF/B-VII.1 evaluations. Decay data include all ground and metastable state nuclides with half-lives greater than 1 millisecond. Using these data sources, ORIGEN currently tracks 174 actinides, 1149 fission products, and 974 activation products. The purpose of this chapter is to describe the stand-alone capabilities and underlying methodology of ORIGEN—as opposed to the integrated depletion capability it provides in all coupled neutron transport/depletion sequencesmore » in SCALE, as described in other chapters.« less

Source terms and spent nuclear fuel (SNF) storage pool decay heat load analyses for operating nuclear power plants require a large number of Oak Ridge Isotope Generation and Depletion (ORIGEN) calculations. SNF source term calculations also require a significant amount of bookkeeping to track quantities such as core and assembly operating histories, spent fuel pool (SFP) residence times, heavy metal masses, and enrichments. The ORIGEN Assembly Isotopics (ORIGAMI) module in the SCALE code system provides a simple scheme for entering these data. However, given the large scope of the analysis, extensive scripting is necessary to convert formats and process data to create thousands of ORIGAMI input files (one per assembly) and to process the results into formats readily usable by follow-on analysis tools. This primer describes a project within the SCALE Fulcrum graphical user interface (GUI) called ORIGAMI Automator that was developed to automate the scripting and bookkeeping in large-scale source term analyses. The ORIGAMI Automator enables the analyst to (1) easily create, view, and edit the reactor site and assembly information, (2) automatically create and run ORIGAMI inputs, and (3) analyze the results from ORIGAMI. ORIGAMI Automator uses the standard ORIGEN binary concentrations files produced by ORIGAMI, with concentrationsmore » available at all time points in each assembly’s life. The GUI plots results such as mass, concentration, activity, and decay heat using a powerful new ORIGEN Post-Processing Utility for SCALE (OPUS) GUI component. This document includes a description and user guide for the GUI, a step-by-step tutorial for a simplified scenario, and appendices that document the file structures used.« less

The US Department of Energy’s Nuclear Energy Advanced Modeling and Simulation (NEAMS) Program has contributed significantly to the development of the PROTEUS neutron transport code at Argonne National Laboratory and to the Oak Ridge Isotope Generation and Depletion Code (ORIGEN) depletion/decay code at Oak Ridge National Laboratory. PROTEUS’s key capability is the efficient and scalable (up to hundreds of thousands of cores) neutron transport solver on general, unstructured, three-dimensional finite-element-type meshes. The scalability and mesh generality enable the transfer of neutron and power distributions to other codes in the NEAMS toolkit for advanced multiphysics analysis. Recently, ORIGEN has received considerable modernization to provide the high-performance depletion/decay capability within the NEAMS toolkit. This work presents a description of the initial integration of ORIGEN in PROTEUS, mainly performed during FY 2015, with minor updates in FY 2016.

This report documents work performed supporting the Department of Energy (DOE) Office of Nuclear Energy (NE) Fuel Cycle Technologies Used Fuel Disposition Campaign (UFDC) under work breakdown structure element 1.02.08.10, ST Analysis. In particular, this report fulfills the M4 milestone M4FT- 15OR0810036, Quantify effects of power uncertainty on fuel assembly characteristics, within work package FT-15OR081003 ST Analysis-ORNL. This research was also supported by the Consortium for Advanced Simulation of Light Water Reactors (http://www.casl.gov), an Energy Innovation Hub (http://www.energy.gov/hubs) for Modeling and Simulation of Nuclear Reactors under U.S. Department of Energy Contract No. DE-AC05-00OR22725. The discharge rod internal pressure (RIP) and cladding hoop stress (CHS) distributions are quantified for Watts Bar Nuclear Unit 1 (WBN1) fuel rods by modeling core cycle design data, operation data (including modeling significant trips and downpowers), and as-built fuel enrichments and densities of each fuel rod in FRAPCON-3.5. A methodology is developed which tracks inter-cycle assembly movements and assembly batch fabrication information to build individual FRAPCON inputs for each evaluated WBN1 fuel rod. An alternate model for the amount of helium released from the zirconium diboride (ZrB2) integral fuel burnable absorber (IFBA) layer is derived and applied to FRAPCON output data to quantify the RIP andmore » CHS for these types of fuel rods. SCALE/Polaris is used to quantify fuel rodspecific spectral quantities and the amount of gaseous fission products produced in the fuel for use in FRAPCON inputs. Fuel rods with ZrB2 IFBA layers (i.e., IFBA rods) are determined to have RIP predictions that are elevated when compared to fuel rod without IFBA layers (i.e., standard rods) despite the fact that IFBA rods often have reduced fill pressures and annular fuel pellets. The primary contributor to elevated RIP predictions at burnups less than and greater than 30 GWd/MTU is determined to be the total fuel rod void volume and the amount of released fission gas in the fuel rod, respectively. Cumulative distribution functions (CDFs) are prepared from the distribution of RIP and CHS predictions for all standard and IFBA rods. The provided CDFs allow for the determination of the portion of WBN1 fuel rods that exceed a specified RIP or CHS limit. Results are separated into IFBA and standard rods so that the two groups may be analyzed individually. FRAPCON results are provided in sufficient detail to enable the recalculation of the RIP while considering any desired plenum gas temperature, total void volume, or total amount of gas present in the void volume. A method to predict the CHS from a determined or assumed RIP is also proposed, which is based on the approximately linear relationship between the CHS and the RIP. Finally, improvements to the computational methodology of FRAPCON are proposed.« less

This is a project in the field of fundamental research on numerical methods for solving the particle transport equation. Numerous practical problems require to use unstructured meshes, for example, detailed nuclear reactor assembly-level calculations, large-scale reactor core calculations, radiative hydrodynamics problems, where the mesh is determined by hydrodynamic processes, and well-logging problems in which the media structure has very complicated geometry. Currently this is an area of very active research in numerical transport theory. main issues in developing numerical methods for solving the transport equation are the accuracy of the numerical solution and effectiveness of iteration procedure. The problem in case of unstructured grids is that it is very difficult to derive an iteration algorithm that will be unconditionally stable.

To model Generation IV reactor systems in detail, INEEL is currently developing a new thermal hydraulic analysis tool coupling RELAP5-3D / ATHENA© and the computational fluid dynamics (CFD) software, Fluent. One of the first steps in this endeavor is extensive validation and verification (V&V) of Fluent for various situations of interest, such as the abrupt expansion of a gas entering a gas-cooled reactor core. Fluent results were compared to validation data provided by Baughn, et al. on turbulent air flow through an axisymmetric pipe expansion with constant wall heat flux [1] and uniform wall temperature [2]. Fluent peak Nusselt numbers varied 25% from validation data—well outside experimental uncertainties of 5%. However, non-peak Nusselt numbers varied only 10% from validation data and fully-developed Nusselt numbers were in good agreement with widely-accepted empirical relations such as the Dittus-Boelter Correlation.Copyright © 2002 by ASME