GOKHAN YESILYURT

MISCELLANEOUS

ABOUT SCALE

NATURE OF PROBLEM SOLVED

The SCALE system was developed for the Nuclear Regulatory Commission to satisfy a need for a standardized method of analysis for the evaluation of nuclear fuel facility and package designs. In its present form, the system has the capability to perform criticality, shielding, radiation source term, spent fuel depletion/decay, reactor physics, and sensitivity/uncertainty analyses using well-established functional modules tailored to the SCALE system. The CSAS5 control module contains criticality safety analysis sequences that calculate the neutron multiplication factor for one-dimensional (XSDRNPM) and multidimensional (KENO V.a) system models. The CSAS5 module also has the capability to perform criticality searches (optimum, minimum, or specified values of k-eff) on geometry dimensions or nuclide concentrations in KENO V.a. The CSAS6 control module contains criticality safety analysis sequences using the KENO-VI module for multidimensional models with more complex geometries, including hexagonal arrays. Sequences that provide problem-dependent multigroup cross sections for use in stand-alone codes are also available in the CSAS5 module. Both KENO modules can perform continuous energy calculations in SCALE 6. In addition, sensitivity and uncertainty (S/U) analysis capabilities for criticality safety are included in SCALE. Both 1-D and 3-D sequences plus several auxiliary codes have been developed into a suite of sensitivity and uncertainty analysis codes called TSUNAMI (Tools for Sensitivity and Uncertainty Analysis Methodology Implementation). TSUNAMI contains a number of codes that were developed primarily to assess the degree of applicability of benchmark experiments for use in criticality code validations. However, the sensitivity and uncertainty data produced by these codes can be used in a wide range of studies. Sensitivity coefficients produced by the TSUNAMI sensitivity analysis sequences predict the relative changes in a system’s calculated keff value due to changes in the neutron cross-section data. Both TSUNAMI-1D and TSUNAMI-3D fold the sensitivity data with cross-section covariance data to calculate the uncertainty in the calculated keffvalue due to tabulated uncertainties in the cross-section data. The applicability of benchmark experiments to the criticality safety validation of a given application can be assessed using S/U-based integral indices. The TSUNAMI-IP (Indices and Parameters) code utilizes sensitivity data and cross-section covariance data to produce a number of relational integral indices that can be used to assess system similarity. Two-dimensional (2-D) spent fuel depletion is available in the TRITON control module. TRITON couples ORIGEN-S depletion calculations with the 2-D flexible mesh discrete ordinates code NEWT. TRITON supports branch calculations that allow calculation of cross sections and their first derivatives with respect to fuel and moderator temperature, moderator density, soluble boron concentration, and control rod insertion, as a function of burnup. These cross sections are stored in a database format that can be retrieved and processed as appropriate for use by core analysis codes. The rigor of the NEWT solution in estimating angular flux distributions combined with the world-recognized accuracy of ORIGEN-S depletion gives TRITON the capability to perform rigorous burnup-dependent physics calculations with few implicit approximations. Three-dimensional (3-D) Monte Carlo spent fuel depletion is available in SCALE via the TRITON and TRITON6 control modules. TRITON couples ORIGEN-S depletion calculations with KENO V.a, while TRITON6 uses KENO-VI. ORIGEN-ARP is an automated depletion decay sequence for both Windows and Unix/Linux systems. It includes a Windows graphical user interface (GUI) for ORIGEN-S and ARP (Automated Rapid Processing), which automatically interpolates cross sections on enrichment, burnup, and optionally moderator density using a set of standard basic cross-section libraries for LWR and MOX fuel assembly designs. The interpolated cross sections are passed to ORIGEN-S. Utility codes are provided so users can generate their own ORIGEN-ARP basic cross-section libraries via TRITON. Other automated criticality safety related sequences include the STARBUCS 3-D burnup credit sequence (combining ORIGEN-ARP with KENO V.a or KENO-VI) and the SMORES 1-D material optimization sequence for criticality safety. A new general purpose 3-D radiation shielding sequence has been developed for SCALE 6. The MAVRIC control module uses the new Monaco Monte Carlo shielding module to perform analyses with the automated 3-D variance reduction CADIS methodology using module xkba (eXecutable Koch-Baker-Alcouffe) of the new Denovo 3-D discrete ordinates code system. This automated scheme generates 3-D Monte Carlo biasing parameters that enable MAVRIC to calculate accurate doses with outstanding efficiency. The Monaco geometry input is identical to KENO-VI. In addition, the capability to perform criticality accident alarm system (CAAS) analysis using KENO-VI coupled with MAVRIC is provided. Two other shielding analysis sequences are provided in SCALE. SAS1 analyzes general 1-D shielding problems via XSDRNPM-S. The QADS module analyzes 3-D gamma-ray shielding problems via the point kernel code, QAD-CGGP.

METHOD OF SOLUTION

The SCALE system consists of easy-to-use analytical sequences which are automated to perform the necessary data processing and manipulation of well-established computer codes required by the sequence. Thus the user is able to select an analytical sequence characterized by the type of analysis (criticality, shielding, or heat transfer) to be performed and the geometric complexity of the system being analyzed. The user then prepares a single set of input for the control module corresponding to this analytical sequence. The control module input is in terms of easily visualized engineering parameters specified in a simplified, free-form format. The control modules use this information to derive additional parameters and prepare the input for each of the functional modules in the analytical sequence. Provisions have also been made to allow the user to execute the functional modules on a stand-alone basis. The radiation transport codes employ either discrete ordinates or Monte Carlo methods.

Gokhan YesilyurtLead Nuclear Engineer @ X Energy, LLC in Greenbelt, MD 20770