FUDGE (For Updating Data and Generating Evaluations) is an open-source code that supports reading, visualizing, checking, modifying, and processing nuclear reaction and decay data. For ease of use the front-end of FUDGE is written in Python while C and C++ routines are employed for computationally intensive calculations. FUDGE has been developed primarily at Lawrence Livermore National Laboratory (LLNL) with contributions from Brookhaven National Laboratory (BNL). It is used by the LLNL Nuclear Data and Theory (NDT) group to deliver high-quality nuclear data libraries to users for a variety of applications. FUDGE is also the world leader in converting data to the Generalized Nuclear Database Structure (GNDS) and working with GNDS data, including processing and visualizing. GNDS is a new extensible hierarchy that has been internationally adopted as the new standard for storing and using nuclear data libraries, replacing the previous standard ENDF-6. A new public release of FUDGE has recently been published on github. This paper gives an overview of nuclear data processing capabilities in FUDGE, as well as describing the latest release, new capabilities, future plans, and basic instructions for users interested in applying FUDGE to their nuclear data workflow.

The AGENT (Arbitrary GEometry Neutron Transport) an open-architecture reactor modeling tool is deterministic neutron transport code for two or three-dimensional heterogeneous neutronic design and analysis of the whole reactor cores regardless of geometry types and material configurations. The AGENT neutron transport methodology is applicable to all generations of nuclear power and research reactors. It combines three theories: (1) the theory of R-functions used to generate real three-dimensional whole-cores of square, hexagonal or triangular cross sections, (2) the planar method of characteristics used to solve isotropic neutron transport in non-homogenized 2D) reactor slices, and (3) the one-dimensional diffusion theory used to couple the planar and axial neutron tracks through the transverse leakage and angular mesh-wise flux values. The R-function-geometrical module allows a sequential building of the layers of geometry and automatic submeshing based on the network of domain functions. The simplicity of geometry description and selection of parameters for accurate treatment of neutron propagation is achieved through the Boolean algebraic hierarchically organized simple primitives into complex domains (both being represented with corresponding domain functions). The accuracy is comparable to Monte Carlo codes and is obtained by following neutron propagation through real geometrical domains that does not require homogenization or simplifications. The efficiency is maintained through a set of acceleration techniques introduced at all important calculation levels. The flux solution incorporates power iteration with two different acceleration techniques: Coarse Mesh Rebalancing (CMR) and Coarse Mesh Finite Difference (CMFD). The stand-alone originally developed graphical user interface of the AGENT code design environment allows the user to view and verify input data by displaying the geometry and material distribution. The user can also view the output data such as three-dimensional maps of the energy-dependent mesh-wise scalar flux, reaction rate and power peaking factor. The AGENT code is in a process of an extensive and rigorous testing for various reactor types through the evaluation of its performance (ability to model any reactor geometry type), accuracy (in comparison with Monte Carlo results and other deterministic solutions or experimental data) and efficiency (computational speed that is directly determined by the mathematical and numerical solution to the iterative approach of the flux convergence). This paper outlines main aspects of the theories unified into the AGENT code formalism and demonstrates the code performance, accuracy and efficiency using few representative examples. The AGENT code is a main part of the so called virtual reactor system developed for numerical simulations of research reactors. Few illustrative examples of the web interface are briefly outlined.

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GIDI+ is a C++ package that reads GNDS formatted nuclear reaction and structure data, and photo-atomic reaction data as needed by radiation transport codes. As of version 3.25, GIDI+ supports reading GNDS 2.0 formatted data. GNDS is a new extensible hierarchy that has been internationally adopted as the new standard for storing nuclear and photo-atomic data, replacing ENDF-6 which has been the standard since the 1960’s. GIDI+ 3.25 supports the GNDS 2.0 map file which can contain a list of all GNDS data needed to make a complete nuclear reaction data library as needed by radiation transport codes. GIDI+ has functions that make it easy to obtain multi-group data from a GNDS file, including summed data (e.g., summing multi-group reaction cross sections to obtain the total multi-group cross section). In addition, the MCGIDI sub-package of GIDI+ can be used by Monte Carlo transport codes to look up data (e.g., cross section, deposition energy) as a function of projectile energy and material temperature. The MCGIDI sub-package can use either GNDS multi-groupor continuous energy data. MCGIDI also has functions for sampling a reaction from a GNDS reactionSuite and functions for sampling a reaction’s outgoing particle energy and angular data. Recent work on GIDI+ includes: supporting GNDS 2.0, speeding up load times and adding support for URR probability tables. GIDI+ is currently implemented in LLNL’s Mercury [4] (Monte Carlo) and Ardra [5] (deterministic) radiation transport codes. An effort to update GEANT4 to use the latest GIDI+ is underway and should be completed by March 2023. GIDI+ is released under the MIT license and can be found at github.com/LLNL/gidiplus.

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