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DOE OSTI2018

Study on neutronics calculation method of SCWR with annular fuel assembly - 26804

Many Supercritical Water Reactor designs have been proposed all over the world. Annular fuel was employed in SCWR by Xi'an JiaoTong University to achieve lower fuel center temperature and lower cladding surface temperature. Different analysis codes have been developed for these designs. In this article, a neutronics calculation code for SCWR with annular fuel is developed. This code is compared with Monte-Carlo code doing both assembly and core calculation. The results showed that the spectra difference of PWRs and SCWRs is the main reason for assembly calculation errors and conventional assembly homogenization method is not suitable for SCWRs. Thus the group-collapsing and spatial-averaging method were studied to improve the accuracy. The few-group structure which contains 14 energy groups is used. The 'super-assembly' model is employed to consider effect of reflector. The results showed that the accuracy of effective multiplication factor and radial power was improved. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2015

AMPX: A Modern cross section processing system for generating nuclear data libraries - 14619

Nuclear data libraries are essential for nuclear analysis code systems like SCALE. AMPX has been continuously developed since the early 1970's and is used to generate these data libraries for SCALE. Recently, Oak Ridge National Laboratory (ORNL) initiated a major modernization effort to position AMPX for continued development and deployment on modern computing platforms. An important step toward modernization has been to merge AMPX under the SCALE continuous integration (CI) development framework. As a result, AMPX is now developed under the SCALE software quality assurance plan, providing increased confidence in the quality of data libraries developed and deployed by AMPX. This merge allowed us to develop novel methods to generate self-shielding factors that combine SCALE's resonance self-shielding capabilities with the AMPX library-generation capabilities. AMPX processes Evaluated Nuclear Data Format (ENDF/B) files. The ENDF community is working to develop a more modern format that will replace the existing ENDF/B format. Due to strict space and memory requirements when AMPX was first developed, the original AMPX processing and reading of ENDF data were tightly integrated. On newer computer systems, this is no longer necessary. Now there is a separation between reading and processing which allows for simultaneous support of current and new ENDF formats. As a result, AMPX is positioned to adopt any new ENDF format when available. For end users, the modernization effort has resulted in improved processing capabilities for generating nuclear data libraries. The objective of the paper is to document the latest AMPX cross-section processing capabilities and provide the current status of the AMPX modernization effort. (authors)

42 ENGINEERING↗

DOE OSTI2022

Implementation of hybrid finite element method based transport solver in GRIFFIN

A new transport solver option based on the hybrid FEM (HFEM) was implemented in GRIFFIN, the MOOSE-based reactor analysis code, as an effort to support routine core design calculations for advanced reactor applications. The HFEM formulation with P{sub N} (spherical harmonics expansion), akin to the variational nodal method, is effective for solving a spatially homogenized problem with strong transport effect. The residual and Jacobian evaluations of the HFEM weak form were derived and successfully implemented in GRIFFIN, having the diffusion and the PN options available in the new HFEM based transport solver. The performance was tested with the simplified ABTR benchmark problems. The results indicate that the HFEM-based transport solver is a feasible option for solving problems with spatially homogenized and strong streaming by providing superior accuracy with a proper p-refinement. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2022

Energy deposition model in STREAM with photon transport

Recent efforts have been taken to implement a photon transport module for the neutron transport analysis code STREAM, considering the explicit photon energy distribution in the problem instead of using the locally deposition assumption as employed in previous STREAM versions. The current energy release model in STREAM based on the On-the-fly energy released per fission model is modified accordingly to account for the photon energy deposition computed explicitly by this module. The impacts of the modified energy release model with photon transport in STREAM on the power distribution, fuel temperature and other core parameters during depletion with feedback calculations are studied on several problems from the VERA benchmark suit. Overall, the photon transport provides a better power map for depletion and feedback calculations, especially when strong gamma emitters are present. The fuel temperature, in general, is decreased when the photon transport is employed because fission gammas have been transported out of their birthplace. The new energy deposition model in STREAM with photon transport indicates that gamma energy accounts for about 9.7% of the total energy released in a fresh VERA5 2D core and rises to 10.5% at the End of Cycle. Around 2.5% of the total energy released will be deposited in the coolant in this 2D core. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2024

Advances in Metallic Fuel Database Development and Data Qualification

The Fuels Irradiation and Physics Database (FIPD [1]) is a comprehensive repository of data and documents related to Uranium-Zirconium based metallic fuel test pins. This database stores operational conditions of these pins, calculated using a suite of Argonne National Laboratory analysis codes developed during the Integral Fast Reactor (IFR) program. Key calculated data include axial distributions of power, temperature, fluence, burnup, and isotopic densities. Additionally, the FIPD holds post-irradiation examination (PIE) data such as fission gas release, gas chemistry measurements, and axial distributions derived from profilometry, gamma scanning, and neutron radiography. Complementing these data is an extensive archive of documents related to various pins and experiments. These include raw PIE records, design details, safety analyses, and operational reports. More detail about FIPD can be found in ref. [2]. The database development is an ongoing effort covering metallic fuel experiments from the Experimental Breeder Reactor II (EBR-II) and the Fast Flux Test Facility (FFTF). The recent improvements to the database and the data QA status are summarized in this paper.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2025

Radiological Safety Analysis Computer (RSAC 7.2) Presentation

This presentation will be for the graduate seminar class for the University of New Mexico. We will discuss the Radiological Safety Analysis Computer (RSAC) version 7.2, a critical tool in radiological consequence analysis within the nuclear safety domain. We will begin with an introduction to radiological consequence analysis and the importance of dose modeling in nuclear safety, referencing key regulatory frameworks such as DOE-STD-3009 and 10 CFR 20. The presentation will provide an overview of RSAC, including its history, primary functions, and the inputs and outputs involved in its calculations. A detailed examination of RSAC's atmospheric dispersion model, based on the Gaussian plume methodology, will be presented, highlighting how meteorological factors influence the results. We will discuss the regulatory alignment of RSAC with DOE and NRC guidelines, emphasizing the assumptions required for compliance and the balance between conservative and realistic inputs. The presentation will also address the limitations and assumptions inherent in RSAC, comparing it with other consequence analysis codes like MACCS, HotSpot, and RASCAL. An example RSAC run will be demonstrated, showcasing a hypothetical accident scenario and interpreting the resulting dose calculations. Finally, we will introduce new ideas to enhance the current RSAC framework. This future work aims to provide multi-industry adaptability and integration with new dispersion models, addressing the evolving needs of DOE and NRC. The presentation will conclude with a discussion on RSAC's ongoing role in nuclear safety and the opportunities for future innovation.

computer program↗

DOE OSTI2015

A new code for spectrometric analysis for environmental radiological surveillance on monitors focused on gamma radioactivity on aerosols

pGamma is a code developed by the NERG group of the Technical University of Catalonia - Barcelona Tech for the analysis of gamma spectra generated by the Equipment for the Continuous Measurement and Identification of Gamma Radioactivity on Aerosols with Paper Filter developed for our group and Raditel Servies company. Nowadays the code is in the process of adaptation for the monitors of the Environmental Radiological Surveillance Network of the Local Government of Catalonia (Generalitat of Catalonia), Spain. The code is a Spectrum Analysis System, it identifies the gamma emitters on the spectrum, determines its Concentration of Activity, generates alarms depending on the Activity of the emitters and generates a report. The Spectrum Analysis System includes a library with emitters of interest, NORM and artificial. The code is being used on the three stations with the aerosol monitor of the Network (Asco and Vandellos, near both Nuclear Power Plants and Barcelona). (authors)

54 ENVIRONMENTAL SCIENCES↗

DOE OSTI2013

MC 2 -3: Multigroup Cross Section Generation Code for Fast Reactor Analysis

The MC 2 -3 code is a Multigroup Cross section generation Code for fast reactor analysis, developed by improving the resonance self-shielding and spectrum calculation methods of MC 2 -2 and integrating the one-dimensional cell calculation capabilities of SDX. The code solves the consistent P1 multigroup transport equation using basic neutron data from ENDF/B data files to determine the fundamental mode spectra for use in generating multigroup neutron cross sections. A homogeneous medium or a heterogeneous slab or cylindrical unit cell problem is solved in ultrafine (~2000) or hyperfine (~400,000) group levels. In the resolved resonance range, pointwise cross sections are reconstructed with Doppler broadening at specified isotopic temperatures. The pointwise cross sections are directly used in the hyperfine group calculation whereas for the ultrafine group calculation, self-shielded cross sections are prepared by numerical integration of the pointwise cross sections based upon the narrow resonance approximation. For both the hyperfine and ultrafine group calculations, unresolved resonances are self-shielded using the analytic resonance integral method. The ultrafine group calculation can also be performed for two-dimensional whole-core problems to generate region-dependent broad-group cross sections. Multigroup cross sections are written in the ISOTXS format for a user-specified group structure. The code is executable on UNIX, Linux, and PC Windows systems, and its library includes all isotopes of the ENDF/BVII. 0 data.

20 FOSSIL-FUELED POWER PLANTS↗

DOE OSTI2021

Data Analysis of Energy Code Compliance in Commercial Buildings

What percent of newly constructed commercial buildings comply with the energy code? How much energy and cost could be saved if the compliance increased? Which code requirements have both low compliance rates and high savings potential? These are the questions U.S. Department of Energy (DOE) is trying to answer through its Commercial Energy Code Field Study. Previous commercial studies have been very limited and did not result in a widely accepted and tested methodology. DOE’s goal is to create a standardized methodology that can produce actionable results at a reasonable study cost, that can be used by state and local governments and utilities and provides valuable information to policy makers. The field study team implemented the pilot methodology, compiling a data set of 230 office and retail buildings in two climate zones. The approach is based on identifying lost savings on a total energy cost basis rather than simply counting the quantity of measures that do not meet code. This report is focused on the analysis of the collected data. The primary goal was to analyze the data collected during the field study and determine the actual energy cost impact of each measure in a non-compliance situation. The energy impact results allowed for ranking the measures to identify which have the highest potential for lost savings. These results combined with the time required to verify each measure will allow future compliance verification to focus on measures that had a large impact on energy use over the life of the building and those that have the greatest savings recovery potential per verification hour.

32 ENERGY CONSERVATION, CONSUMPTION, AND UTILIZATI↗

DOE OSTI2023

Data Analysis of Energy Code Compliance in Commercial Buildings (Rev. 1)

29 ENERGY PLANNING, POLICY, AND ECONOMY↗

DOE OSTI2025

ThinCurr: An open-source 3D thin-wall eddy current modeling code for the analysis of large-scale systems of conducting structures

In this paper we present a new thin-wall eddy current modeling code, ThinCurr, for studying inductively-coupled currents in 3D conducting structures -- with primary application focused on the interaction between currents flowing in coils, plasma, and conducting structures of magnetically-confined plasma devices. The code utilizes a boundary finite element method on an unstructured, triangular grid to accurately capture device structures. The new code, part of the broader Open FUSION Toolkit, is open-source and designed for ease of use without sacrificing capability and speed through a combination of Python, Fortran, and C/C++ components. Scalability to large models is enabled through use of hierarchical off-diagonal low-rank compression of the inductance matrix, which is otherwise dense. Ease of handling large models of complicated geometry is further supported by automatic determination of supplemental elements through a greedy homology approach. Here, a detailed description of the numerical methods of the code and verification of the implementation of those methods using cross-code comparisons against the VALEN code and Ansys commercial analysis software is shown.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2021

Verification and Validation of the PLTEMP/ANL Code for Thermal-Hydraulic Analysis of Experimental and Test Reactors (Code Version 4.3)

The document compiles in a single volume the verification and validation works done for the PLTEMP/ANL code during the years of its development and improvement. The verification of sixteen capabilities of the PLTEMP/ANL Version 4.3 code that were identified by research reactor analysts as frequently used in their thermal-hydraulic analysis are described and documented. Each chapter of the document deals with the verification or validation of a specific part of the software that calculates a specific phenomenon, e.g., channel flow calculation, coolant property calculation, heat transfer calculation, and flow instability calculation. The software verification work is usually done by comparing the code with hand calculation, Microsoft spreadsheet calculation, or Mathematica calculation. The software validation is done by comparing the code with experimental data or a widely tested code like the RELAP5 code. In addition, some PLTEMP/ANL verification works that are available in the open literature are simply referenced and not included in the document. PLTEMP/ANL has been used in conversion safety analysis reports of several US and foreign research reactors that have been licensed and converted. A list of such reactors is given.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2018

MC 2 -3: Multigroup Cross Section Generation Code for Fast Reactor Analysis Nuclear

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2014

Interface design of VSOP'94 computer code for safety analysis

Today, most software applications, also in the nuclear field, come with a graphical user interface. VSOP'94 (Very Superior Old Program), was designed to simplify the process of performing reactor simulation. VSOP is a integrated code system to simulate the life history of a nuclear reactor that is devoted in education and research. One advantage of VSOP program is its ability to calculate the neutron spectrum estimation, fuel cycle, 2-D diffusion, resonance integral, estimation of reactors fuel costs, and integrated thermal hydraulics. VSOP also can be used to comparative studies and simulation of reactor safety. However, existing VSOP is a conventional program, which was developed using Fortran 65 and have several problems in using it, for example, it is only operated on Dec Alpha mainframe platforms and provide text-based output, difficult to use, especially in data preparation and interpretation of results. We develop a GUI-VSOP, which is an interface program to facilitate the preparation of data, run the VSOP code and read the results in a more user friendly way and useable on the Personal 'Computer (PC). Modifications include the development of interfaces on preprocessing, processing and postprocessing. GUI-based interface for preprocessing aims to provide a convenience way in preparing data. Processing interface is intended to provide convenience in configuring input files and libraries and do compiling VSOP code. Postprocessing interface designed to visualized the VSOP output in table and graphic forms. GUI-VSOP expected to be useful to simplify and speed up the process and analysis of safety aspects.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2016

Development of random geometry capability in RMC code for explicit analysis of polytype particle/pebble - 18493

Monte Carlo method plays an important role in accurate simulation of High Temperature Gas-Cooled Reactors (HTR) and molten salt reactors, owing to its advantages of the flexible geometry modeling and the use of continuous-energy nuclear cross sections. For the innovative design of pebble bed, prism HTRs and molten salt reactors, there may be more than one kind of particles or pebbles in the system, such as the burnable poison particles, graphite pebbles or fuel pebbles with different enrichments. Different treatments have been used to deal with the double-heterogeneous geometry of these reactors. In some researches, the particle distributions in stochastic media are assumed the regular lattice, while in the others, the explicit packing methods such as Random Sequential Addition (RSA) and Discrete Element Method (DEM) are used, by generating the random coordinates of the particles/pebbles. However, traditional geometry with repeated structure has difficulties in explicit modeling polytype particle/pebble, due to the arbitrary packing fractions of different kinds of particles or pebbles. Although, packing methods such as Random Sequential Addition (RSA) and Discrete Element Method (DEM) could solve the polytype particle/pebble problems to some extent, but these methods are troublesome in geometry modeling and time consuming. In this paper, a new Random Universe Geometry method was developed and implemented in Monte Carlo code RMC, for explicit analysis of polytype particle/pebble in double heterogeneous geometry systems, such as HTRs and molten salt reactors. This method was applied to constructed mixed particles/pebbles problems and also simplified HTR-10 full core problem to test the feasibility, and the results show that the random geometry modeling can provide explicit packing fraction of different kinds of particles or pebbles. The results are also compared with that calculated by RSA method and regular structure, and agree quite well. This work would provide an effective tool for the innovative design of pebble bed, prism HTRs and molten salt reactors with polytype particles or pebbles using Monte Carlo method. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2014

RELAP-7 Software Verification and Validation Plan

This INL plan comprehensively describes the software for RELAP-7 and documents the software, interface, and software design requirements for the application. The plan also describes the testing-based software verification and validation (SV&V) process—a set of specially designed software models used to test RELAP-7. The RELAP-7 (Reactor Excursion and Leak Analysis Program) code is a nuclear reactor system safety analysis code being developed at Idaho National Laboratory (INL). The code is based on the INL’s modern scientific software development framework – MOOSE (Multi-Physics Object-Oriented Simulation Environment). The overall design goal of RELAP-7 is to take advantage of the previous thirty years of advancements in computer architecture, software design, numerical integration methods, and physical models. The end result will be a reactor systems analysis capability that retains and improves upon RELAP5’s capability and extends the analysis capability for all reactor system simulation scenarios.

97 MATHEMATICS AND COMPUTING↗

DOE OSTI2023

MAS: A versatile Landau-fluid eigenvalue code for plasma stability analysis in general geometry

We have developed a new global eigenvalue code, Multiscale Analysis for plasma Stabilities (MAS), for studying plasma problems with wave toroidal mode number (n) and frequency (ω) in a broad range of interest in general tokamak geometry, based on a five-field Landau-fluid description of thermal plasmas. Beyond keeping the necessary plasma fluid response, we further retain the important kinetic effects including diamagnetic drift, ion finite Larmor radius, finite parallel electric field (E||), ion and electron Landau resonances in a self-consistent and non-perturbative manner without sacrificing the attractive efficiency in computation. The physical capabilities of the code are evaluated and examined in the aspects of both theory and simulation. In theory, the comprehensive Landau-fluid model implemented in MAS can be reduced to the well-known ideal MHD model, electrostatic ion-fluid model, and drift-kinetic model in various limits, which clearly delineates the physics validity regime. In simulation, MAS has been well benchmarked with theory and other gyrokinetic and kinetic-MHD hybrid codes in a manner of adopting the unified physical and numerical framework, which covers the kinetic Alfv\'en wave (KAW), ion sound wave (ISW), low-n kink, high-n ion temperature gradient mode (ITG) and kinetic ballooning mode (KBM). Moreover, MAS is successfully applied to model the Alfv\'en eigenmode (AE) activities in DIII-D discharge #159243, which faithfully captures the frequency sweeping of reversed shear Alfv\'en eigenmode (RSAE), the tunneling damping of toroidal Alfv\'en eigenmode (TAE), as well as the polarization characteristics of kinetic beta-induced Alfv\'en eigenmode (KBAE) and beta-induced Alfv\'en-acoustic eigenmode (BAAE) being consistent with former gyrokinetic theory and simulation. With respect to the key progress contributed to the community, MAS has the advantage of combining rich physics ingredients, realistic global geometry and high computation efficiency together for plasma stability analysis in linear regime.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2016

Clean Energy in City Codes: A Baseline Analysis of Municipal Codification across the United States

Municipal governments in the United States are well positioned to influence clean energy (energy efficiency and alternative energy) and transportation technology and strategy implementation within their jurisdictions through planning, programs, and codification. Municipal governments are leveraging planning processes and programs to shape their energy futures. There is limited understanding in the literature related to codification, the primary way that municipal governments enact enforceable policies. The authors fill the gap in the literature by documenting the status of municipal codification of clean energy and transportation across the United States. More directly, we leverage online databases of municipal codes to develop national and state-specific representative samples of municipal governments by population size. Our analysis finds that municipal governments with the authority to set residential building energy codes within their jurisdictions frequently do so. In some cases, communities set codes higher than their respective state governments. Examination of codes across the nation indicates that municipal governments are employing their code as a policy mechanism to address clean energy and transportation.

29 ENERGY PLANNING, POLICY, AND ECONOMY↗