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DOE OSTI2015

Year End Progress Report on Rattlesnake Improvements

Rattlesnake is a MOOSE-based radiation transport application developed at INL to support modern multi-physics simulations. At the beginning of the last year, Rattlesnake was able to perform steady-state, transient and eigenvalue calculations for the multigroup radiation transport equations. Various discretization schemes, including continuous finite element method (FEM) with discrete ordinates method (SN) and spherical harmonics expansion method (PN) for the self-adjoint angular flux (SAAF) formulation, continuous FEM (CFEM) with SN for the least square (LS) formulation, diffusion approximation with CFEM and discontinuous FEM (DFEM), have been implemented. A separate toolkit, YAKXS, for multigroup cross section management was developed to support Rattlesnake calculations with feedback both from changes in the field variables, such as fuel temperature, coolant density, and etc., and in isotope inventory. The framework for doing nonlinear diffusion acceleration (NDA) within Rattlesnake has been set up, and both NDA calculations with SAAF-SN-CFEM scheme and Monte Carlo with OpenMC have been performed. It was also used for coupling BISON and RELAP-7 for the full-core multiphysics simulations. Within the last fiscal year, significant improvements have been made in Rattlesnake. Rattlesnake development was migrated into our internal GITLAB development environment at the end of year 2014. Since then total 369 merge requests has been accepted into Rattlesnake. It is noted that the MOOSE framework that Rattlesnake is based on is under continuous developments. Improvements made in MOOSE can improve the Rattlesnake. It is acknowledged that MOOSE developers spent efforts on patching Rattlesnake for the improvements made on the framework side. This report will not cover the code restructuring for better readability and modularity and documentation improvements, which we have spent tremendous effort on. It only details some of improvements in the following sections.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2015

Dosimetric characterization of two radium sources for retrospective dosimetry studies

Purpose: During the first part of the 20th century, {sup 226}Ra was the most used radionuclide for brachytherapy. Retrospective accurate dosimetry, coupled with patient follow up, is important for advancing knowledge on long-term radiation effects. The purpose of this work was to dosimetrically characterize two {sup 226}Ra sources, commonly used in Sweden during the first half of the 20th century, for retrospective dose–effect studies. Methods: An 8 mg {sup 226}Ra tube and a 10 mg {sup 226}Ra needle, used at Radiumhemmet (Karolinska University Hospital, Stockholm, Sweden), from 1925 to the 1960s, were modeled in two independent Monte Carlo (MC) radiation transport codes: GEANT4 and MCNP5. Absorbed dose and collision kerma around the two sources were obtained, from which the TG-43 parameters were derived for the secular equilibrium state. Furthermore, results from this dosimetric formalism were compared with results from a MC simulation with a superficial mould constituted by five needles inside a glass casing, placed over a water phantom, trying to mimic a typical clinical setup. Calculated absorbed doses using the TG-43 formalism were also compared with previously reported measurements and calculations based on the Sievert integral. Finally, the dose rate at large distances from a {sup 226}Ra point-like-source placed in the center of 1 m radius water sphere was calculated with GEANT4. Results: TG-43 parameters [including g{sub L}(r), F(r, θ), Λ, and s{sub K}] have been uploaded in spreadsheets as additional material, and the fitting parameters of a mathematical curve that provides the dose rate between 10 and 60 cm from the source have been provided. Results from TG-43 formalism are consistent within the treatment volume with those of a MC simulation of a typical clinical scenario. Comparisons with reported measurements made with thermoluminescent dosimeters show differences up to 13% along the transverse axis of the radium needle. It has been estimated that the uncertainty associated to the absorbed dose within the treatment volume is 10%–15%, whereas uncertainty of absorbed dose to distant organs is roughly 20%–25%. Conclusions: The results provided here facilitate retrospective dosimetry studies of {sup 226}Ra using modern treatment planning systems, which may be used to improve knowledge on long term radiation effects. It is surely important for the epidemiologic studies to be aware of the estimated uncertainty provided here before extracting their conclusions.

07 ISOTOPE AND RADIATION SOURCES↗

DOE OSTI2024

DASSH-F: Subchannel Based Thermal Analysis

The DASSH thermal analysis code is designed to rapidly allow a reactor design engineer to obtain flow rates requirements that satisfy peak temperature constraints in the domain. The advantage of using DASSH over a hand calculation is that it has a more rigorous treatment of the pin power distribution and coolant heat transfer within an assembly and between assemblies. The advantage of using DASSH over a conventional 3D subchannel code or a computational fluid dynamics code (CFD) is that it can obtain the desired solution in a matter of minutes in serial with minor computer memory needs. The DASSH methodology is virtually identical to SUPERENERGY-2 with additional functionalities taken from follow on work to SUPERENERGY-2 done at ANL in the 1980s. DASSH today is an integral component of the Argonne Fast Reactor analysis suite for reactor design work. DASSH obtains the power distribution from a coupled neutron-gamma heating calculation in GAMSOR (including DIF3D) at each time point of a companion fuel cycle analysis calculation with REBUS. The domain in DASSH assumes a hexagonal grid typical for fast reactors with much of the geometry information taken from the DIF3D model. DASSH assumes the assemblies that are loaded into each grid position are ducted to control the coolant flow. Considerable detail is given on the subchannel formulation of DASSH in this document. Much of the formulation and design of the code builds upon research done by previous authors with little new investigation. Thus the decisions made in developing the subchannel model used in DASSH have their origins over 50 years ago. Much of the heat transfer methodology in DASSH is built upon correlations for both the coolant mixing and heat transfer coefficients for pins and ducts. DASSH is thus not a rigorous treatment of a given problem, but a rapid assessment of the temperature field that has known limitations with respect to an experimental measurement or CFD calculation. The DASSH input and output are detailed along with usage of the software. The DASSH output provides tables of evaluated material properties and key coolant and pin temperature results. DASSH can create Python scripts that generate domain summary pictures. DASSH can also generate assembly temperature maps and VTK output files which allow the DASSH solution to be visualized. As the primary purpose of the DASSH software is to compute the coolant and fuel pin temperature distribution for a given model of a reactor, much of the output focus is giving the user quick summary tables needed to assess the performance of a given orifice flow specification. The present version of DASSH has a crude orifice search capability and an efficient orifice flow search capability. The flow search tries to meet user specified constraints for 1) peak 2-sigma clad midwall temperature, 2) peak coolant temperature, and 3) desired bulk outlet temperature. There are many development shortcomings in DASSH detailed in this document, but this version is functional for modern analysis needs. This document serves as the manual for the Fortran based DASSH software that was developed to replace the Python version of DASSH developed as part of the VTR program.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2017

Improved Modelling for the Assembly Discontinuity Factors in the APEC Method

Modern reactor analysis of light water reactor depends largely on a two-step procedure based on the simplified equivalence theory (SET). Consequently, the resulting homogenized two-group cross-section (XS) and associated discontinuity factor (DF) are subject to noticeable, intrinsic uncertainties. In our previous study, we introduced the albedo-corrected parameterized equivalence constants (APEC) method, which is a new leakage correction method applicable to the fuel assembly homogenized two-group XS of the standard two-step procedure. In the APEC method, fuel assembly two-group XS is expressed as a simple polynomial function of assembly-wise current-to-flux ratio (CFR), and it was demonstrated that two-group XSs can be well corrected during a nodal calculation and the nodal accuracy is improved a lot. However, since the conventional assembly discontinuity factor (ADF) is still used in the current APEC method, a small but non-negligible error is unavoidable. Therefore, it is necessary to include improved DF model in the APEC method for a higher accuracy. The functionalized discontinuity factor (FDF) using the definition of CFR was investigated by authors before the development of the APEC method, where it was motivated by the boundary perturbation theory by F. Rahnema. However, in our previous study on FDF, lattice calculations were performed only with one-dimensional two-group diffusion equation and could not be applied to actual core analysis. In this paper, feasibility study of DF functionalization is performed based on a 2-D method of characteristics (MOC) lattice code, DeCART2D.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2020

Refreshing Waste Tracking Software for the 2020's - 20059

The Waste Compliance and Tracking System (WCATS) is a software application designed to manage waste operations from cradle to grave. The system provides the support needed for characterization, generation, processing, and shipment of all waste generated or stored at Los Alamos National Laboratory (LANL). Waste is tracked with task records for containers in real-time via both desktop and mobile device applications. The software allows application administrators to configure sites with appropriate limits for Pu-239 equivalent Curies (PE-Ci), fissile gram equivalents (FGEs), volume limits, as well as custom limits for site-specific technical safety requirements (TSRs). WCATS was developed based on a need at Los Alamos National Laboratory for a modern, customizable waste tracking system that could support many different waste types. The application is built using Oracle{sup R} database and Oracle{sup R} ADF Swing application framework. The technologies and software frameworks selected for the system were finalized in 2006. Although Oracle{sup R} Database has proven to be robust, the user interface and application framework interface with the database have been deprecated and need to be replaced. The technologies selected for modernization follow the open-source model, which allows the WCATS team to more easily recruit developers, improve the speed of development, and ease the deployment and adoption of the system. The tools that support the re-engineering effort are Vue.js{sup R}, a JavaScript{sup R} framework, and Spring Boot{sup R}, an API (application programming interface) development tool written in Java{sup R}. An API allows for simplified communication from the database to the user interface of the application. In addition, legacy Crystal Reports{sup R} will be replaced with reports developed with the open source Pentaho{sup TM}f Report Designer community edition. Because the application uses HTML and JavaScript{sup R}, it builds upon tools that have been used for many years in web development. These tools have large community support and their proven track record will ensure that enhancements to WCATS can continue for years to come. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2018

Licensing Modernization Project for Advanced Reactor Technologies: FY 2018 Project Status Report

The promise of new nuclear technology and the future of commercial power in the United States (U.S.) are linked to the existence of an efficient and safety-focused regulatory review and licensing process. With an appropriate regulatory framework, reactor suppliers and regulators can design, license, build, and export advanced reactor units more effectively and efficiently and thereby help meet the growing need for clean and reliable energy. To help address the challenge of maintaining compatibility between the regulatory environment and new commercial nuclear plant designs, the U.S. Department of Energy (DOE) is supporting an industry-led effort called the Licensing Modernization Project (LMP). This cost-shared initiative, started in 2016 and scheduled to continue through mid-2019, targets amendment to key elements of the U.S. nuclear power reactor regulatory framework to specifically address licensing barriers in advanced-reactor concepts. The project is focused on updating guidance for certain technical licensing requirements that are largely incompatible with non-light-water reactors (non-LWR) and establish a new pathway for design-safety evaluations and license-application development. Upon completion, the project will have created a new pathway, founded on modern probabilistic assessment techniques, that offers developers, suppliers, regulators, and owner/operatorowner-operators greater clarity and assurance in design and licensing decisions. The LMP employs a team of advanced reactor technology and licensing subject matter experts working closely with U.S. regulators, affected industries, and interested members of the public. The project is managed by Southern Company and supported through a cost share arrangement with DOE. Technical contributions are also provided to the project from the Idaho National Laboratory (INL) Regulatory Development area, the Nuclear Energy Institute (NEI), and the U.S. Nuclear Regulatory Commission (NRC). The LMP identified a suite of high priority issues that provided significant uncertainty within the regulated non-LWR community. It was determined that much of this uncertainty could be retired by issuing updated licensing guidance founded on safety performance demonstrations and modern probabilistic methodologies. The success criteria for the project were established as: • Generate sets of technology inclusive (TI), risk-informed, and performance-based (TI-RIPB) technical requirements, recommendations, and proposals that could be reviewed, amended, and adopted for use by both NRC staff and industry. The proposals would create an optional-use regulatory framework without imposing additional (i.e., mandatory) requirements. • Support LMP the recommended updates of LMP such that they can be finalized, endorsed by NRC, and used by applicants within 5 years. This also supports the DOE advanced- reactor development and deployment goal of completing licensing reviews for technically mature non-LWR concepts sufficient to allow construction to proceed by the early 2030’s. • Further overall programmatic objectives established by DOE’s Gateway for Accelerated Innovation in Nuclear (GAIN) initiative by resolving technology-inclusive policy issues that adversely impact regulatory reviews, siting, permitting, and/or licensing of non-LWRs. This would provide significantly more regulatory certainty for advanced reactor developers.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2017

New hardware and software platform for experiments on a HUBER-5042 X-ray diffractometer with a DISPLEX DE-202 helium cryostat in the temperature range of 20–300 K

Huber-5042 diffractometer with a closed-cycle Displex DE-202 helium cryostat is a unique scientific instrument for carrying out X-ray diffraction experiments when studying the single crystal structure in the temperature range of 20–300 K. To make the service life longer and develop new experimental techniques, the diffractometer control is transferred to a new hardware and software platform. To this end, a modern computer; a new detector reader unit; and new control interfaces for stepper motors, temperature controller, and cryostat vacuum pumping system are used. The system for cooling the X-ray tube, the high-voltage generator, and the helium compressor and pump for maintaining the desired vacuum in the cryostat are replaced. The system for controlling the primary beam shutter is upgraded. A biological shielding is installed. The new program tools, which use the Linux Ubuntu operating system and SPEC constructor, include a set of drivers for control units through the aforementioned interfaces. A program for searching reflections from a sample using fast continuous scanning and a priori information about crystal is written. Thus, the software package for carrying out the complete cycle of precise diffraction experiment (from determining the crystal unit cell to calculating the integral reflection intensities) is upgraded. High quality of the experimental data obtained on this equipment is confirmed in a number of studies in the temperature range from 20 to 300 K.

75 CONDENSED MATTER PHYSICS, SUPERCONDUCTIVITY AND↗

DOE OSTI2017

NRC technical basis for evaluation of its position on protection against common cause failure in digital systems used in nuclear power plants - 378

Digital technology has advantages over analog systems, including automated monitoring and alerts for standby safety functions, and predictive algorithms to maintain critical safety systems. Additionally, digital technology generally has higher reliability and can be designed to reduce single point vulnerabilities. For these reasons many nuclear plants have applied digital technology to safety and non-safety related applications, including reactor protection system, feedwater and turbine controls, etc. with a corresponding significant improvement in trip reduction. Nonetheless, digital instrumentation and control (I and C) systems also present potential new vulnerabilities that need to be assessed, including potential failures due to increased complexity of digital systems, the introduction of unique failure modes due to software (including software common cause failure (CCF)), and limited operating history of digital systems in nuclear safety related applications compared to analog systems. The fact that software is intangible means that common methods, such as analysis or testing, used for detecting CCF may not be effective when applied to software. Consequently, digital technology is perceived to pose a potential risk from the introduction of undetected systematic faults that could result in CCF. Despite the I and C system upgrades and modifications performed to date, the U.S. Nuclear Regulatory Commission (NRC) and industry stakeholders have identified the need to modernize the regulatory infrastructure to efficiently address risks associated with the use of digital technology for nuclear safety applications and address regulatory uncertainties. The NRC's current position on CCF is guided by the staff requirements memorandum (SRM) on SECY 93-087. The SRM provides specific acceptance criteria for the evaluation of CCF, which the staff implemented in the Branch Technical Position (BTP) 7-19. However, industry stakeholders have proposed using methods to characterize the likelihood of software CCF and eliminate it from further consideration in a defense-in-depth and diversity analysis. The NRC's current position does not consider these alternatives, and thus corresponding acceptance criteria is not currently available. The work discussed in this paper assesses the underlying technical basis associated with CCF, provides technical support for updating the NRC position and considers proposed methods for addressing potential CCF in digital systems while enhancing efficiency, clarity, and confidence. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2023

The Energy Transition: Advanced Nuclear Needed but Address Climate Vulnerabilities Now

The term “Energy Transition” is an attempt to capture an elaborate set of activities related to the modernization and decarbonization of energy grids. Performed concurrently and often in an ad hoc manner across local, state, regional and national boundaries, it is bringing chaos to what should arguably be one of the most conservatively managed of all critical infrastructure sectors. What’s more, with climate change producing an increasing tempo of extreme events, confidence in the intended resilient and redundant structure of the electric grids is likely to ebb. Even without these climate induced stressors, the nation’s electric grid was built for an earlier century. In addition to a drive towards greater efficiency via digitization and a continuing price decline in distributed energy resources (DERs), one could argue that climate change concerns are the primary driver of the energy transition. Non-CO2 emitting generation sources like wind and solar have become an important part of the overall generation fleet, albeit ones that cannot be counted upon to provide dispatchable power. Current projections indicate deployment of even larger percentages of DERs in coming years. Until far better storage capabilities arrive, the variability of wind and solar, inconsistent performance of traditional thermal generation plants, and energy delivery failures associated with natural gas pipelines will reinforce mounting reliability concerns. This pertains to both electric transmission and distribution. The recent shuttering of nuclear power plants in Germany, Japan, the US and elsewhere are also putting more downward pressure on dispatchable generation. Russia’s attack on Ukraine has roiled energy markets worldwide and forced some countries to return to coal as a primary fuel. In view circumstances such as these, it is essential that significant changes be made to policies and planning criteria, and to the standards and code on which they are based. Given the accelerating pace of extreme weather events, this needs to occur as soon as possible.

24 POWER TRANSMISSION AND DISTRIBUTION↗

DOE OSTI2017

Computational studies on scattering of radio frequency waves by density filaments in fusion plasmas

In modern magnetic fusion devices, such as tokamaks and stellarators, radio frequency (RF) waves are commonly used for plasma heating and current profile control, as well as for certain diagnostics. The frequencies of the RF waves range from ion cyclotron frequency to the electron cyclotron frequency. The RF waves are launched from structures, like waveguides and current straps, placed near the wall in a very low density, tenuous plasma region of a fusion device. The RF electromagnetic fields have to propagate through this scrape-off layer before coupling power to the core of the plasma. The scrape-off layer is characterized by turbulent plasmas fluctuations and by blobs and filaments. The variations in the edge density due to these fluctuations and filaments can affect the propagation characteristics of the RF waves—changes in density leading to regions with differing plasma permittivity. Analytical full-wave theories have shown that scattering by blobs and filaments can alter the RF power flow into the core of the plasma in a variety of ways, such as through reflection, refraction, diffraction, and shadowing [see, for example, Ram and Hizanidis, Phys. Plasmas 23, 022504 (2016), and references therein]. There are changes in the wave vectors and the distribution of power–scattering leading to coupling of the incident RF wave to other plasma waves, side-scattering, surface waves, and fragmentation of the Poynting flux in the direction towards the core. However, these theoretical models are somewhat idealized. In particular, it is assumed that there is step-function discontinuity in the density between the plasma inside the filament and the background plasma. In this paper, results from numerical simulations of RF scattering by filaments using a commercial full-wave code are described. The filaments are taken to be cylindrical with the axis of the cylinder aligned along the direction of the ambient magnetic field. The plasma inside and outside the filament is assumed to be cold. There are three primary objectives of these studies. The first objective is to validate the numerical simulations by comparing with the analytical results for the same plasma description—a step-function discontinuity in density. A detailed comparison of the Poynting flux shows that numerical simulations lead to the same results as those from the theoretical model. The second objective is to extend the simulations to take into account a smooth transition in density from the background plasma to the interior of the filament. The ensuing comparison shows that the deviations from the results of the theoretical model are quite small. The third objective is to consider the scattering process for situations well beyond a reasonable theoretical analysis. This includes scattering off multiple filaments with different densities and sizes. In conclusion, simulations for these complex arrangements of filaments show that, in spite of the obvious limitations, the essential physics of RF scattering is captured by the analytical theory for a single filament

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2016

Disposal of Greater Than Class C Low-Level Radioactive Waste in Andrews County, Texas - 16408

The U.S. Nuclear Regulatory Commission (NRC) staff has recommended approval of an initiative to permit the Texas Commission on Environmental Quality (TCEQ) to license the disposal of commercial and co-mingled Greater Than Class C Low-Level Radioactive Waste (GTCC LLW) in Andrews County, Texas. Such a framework is clearly articulated in the Section 274 of the Atomic Energy Act of 1954, as amended. The NRC staff also recommended approval to proceed with a rulemaking to align the Waste Classification Tables in Title 10 of the Code of Federal Regulations (CFR), Part 61.55 with the paragraph 61.2 definition of 'waste' to provide a disposal pathway for waste containing certain alpha-emitting transuranic elements with concentrations exceeding 100 nanocuries per gram (nCi/g). A pathway for the disposal of commercial and co-mingled GTCC LLW, including waste with certain alpha emitting transuranic elements exceeding 100 nCi/g (TRU waste), might emerge if the initiative described in the Commissions' July 22, 2015 document, SECY-15-0094, Historical and Current Issues Related to Disposal of GTCC LLW, is approved. The NRC staff recommendations envisioned the NRC and TCEQ working together to review a site-specific performance assessment that would be submitted by Waste Control Specialists LLC (WCS) as part of an amendment to its existing Radioactive Materials License No. R04100. The TCEQ's recommendations were prompted by a petition for rulemaking filed by WCS to allow the disposal of GTCC waste in Texas. On September 10, 2014, the TCEQ Commissioners directed their staff to engage stakeholders and request input from the NRC and the U.S. Department of Energy (DOE) regarding actions needed to proceed with any future rulemaking. On January 30, 2015, the TCEQ requested clarification from the NRC regarding their legal authority and jurisdiction to regulate GTCC, GTCC-like, and TRU waste. The deliberations that have transpired over the past year have raised important legal, policy, and technical matters. This initiative could provide a safe disposal pathway for certain waste streams that have been orphaned over the past 30 years. One important realization that has emerged is the conservative assumptions used to establish the upper limits for Class C LLW when Part 61 was promulgated in 1981 would be much too restrictive when applied to a modern, highly engineered disposal facility located in arid west Texas. Waste that was not suitable for near surface disposal in 1981 could certainly be demonstrated to be safely disposed at the WCS Federal Waste Disposal Facility in Andrews County, Texas. This paper will address the legal, policy, and technical matters supporting this important topic currently under consideration both nationally and within the State of Texas. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2016

On the adaptivity and complexity embedded into differential evolution

This research deals with the comparison of the two modern approaches for evolutionary algorithms, which are the adaptivity and complex chaotic dynamics. This paper aims on the investigations on the chaos-driven Differential Evolution (DE) concept. This paper is aimed at the embedding of discrete dissipative chaotic systems in the form of chaotic pseudo random number generators for the DE and comparing the influence to the performance with the state of the art adaptive representative jDE. This research is focused mainly on the possible disadvantages and advantages of both compared approaches. Repeated simulations for Lozi map driving chaotic systems were performed on the simple benchmark functions set, which are more close to the real optimization problems. Obtained results are compared with the canonical not-chaotic and not adaptive DE. Results show that with used simple test functions, the performance of ChaosDE is better in the most cases than jDE and Canonical DE, furthermore due to the unique sequencing in CPRNG given by the hidden chaotic dynamics, thus better and faster selection of unique individuals from population, ChaosDE is faster.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗

DOE OSTI2016

TH-D-BRB-04: Pinnacle Scripting: Improving Efficiency While Maintaining Safety

Scripting capabilities and application programming interfaces (APIs) are becoming commonly available in modern treatment planning systems. These links to the treatment planning system (TPS) allow users to read data from the TPS, and in some cases use TPS functionality and write data back to the TPS. Such tools are powerful extensions, allowing automation of routine clinical tasks and supporting research, particularly research involving repetitive tasks on large patient populations. The data and functionality exposed by scripting/API capabilities is vendor dependent, as are the languages used by script/API engines, such as the Microsoft .NET framework or Python. Scripts deployed in a clinical environment must be commissioned and validated like any other software tool. This session will provide an overview of scripting applications and a discussion of best practices, followed by a practical introduction to the scripting capabilities of three commercial treatment planning systems. Learning Objectives: Understand the scripting capabilities available in several treatment planning systems Learn how to get started using scripting capabilities Understand the best practices for safe script deployment in a clinical environment R. Popple, Varian Medical Systems has provided research support unrelated to the topic of this session.R. Cardan, Varian Medical Systems for grant research, product evaluation, and teaching honorarium.

60 APPLIED LIFE SCIENCES↗

DOE OSTI2022

Power System Waveform Classification Using Time-Frequency and CNN

Many modern reclosers and circuit breakers have microprocessor relays that record waveforms of system events. In some cases, utilities may record a half-a-dozen event captures for every event. This is thousands of events per year. The industry needs faster, more automated, more conclusive, and easy-to-use systems that can process massive amounts of event recordings without extensive input/support from power system engineers. To address the need for a commercially viable solution that can classify waveform data, energies were directed to develop a universal neural network (NN) structure (deep learning algorithm) that works for a wide variety of system event types. The structure that showed the most promise was one that included the use of spectrograms. The technique has shown positive results in audio engineering, particularly with respect to speech recognition. A waveform signature could be treated as a spoken word like audio waveforms for specific things such as “YES” or “UP”. No two people produce the exact same waveform when speaking each of these words, but audio processing algorithms based on spectrograms and convolutional neural networks (CNN) can still distinguish the word regardless of the speaker. No two circuits produce the exact same waveform for a given event, but the NN can be trained to classify the event type regardless of the circuit or location on the circuit. A Power System Neural Network (PSNN) has been developed to use a CNN to classify events within waveform data for power systems. The waveform is converted to an array of values by way of spectrograms and interpreted as an image. This image is passed into the CNN. The test results on independent simulated test and validation datasets show greater than 99% accuracy. While the results thus far are based on simulated data, the performance of the PSNN is very promising and should work for a wide variety of power system conditions of interest. Ultimately, much of the custom code and tools used today and much of the manual effort expended today may be automated using this PSNN.

24 POWER TRANSMISSION AND DISTRIBUTION↗

DOE OSTI2022

kynema-fmb [SWR-23-07]

Kynema-FMB (FKA: Kynema) is an open-source performance portable flexible multibody (FMB) dynamics solver designed for time-domain simulations. While originally tailored for wind turbine structural dynamics, the formulation and implementation are those of a general flexible-multidbody dynamics solver that can readily be applied to a wide range of systems. Kynema was designed with a narrow focus, namely to provide a lightweight, fast, accurate FMD solver for coupling to computational-fluid-dynamics (CFD) codes, especially the CFD codes in the Kynema suite, for fluid-structure-interaction (FSI) simulations. Kynema-FMB is equipped to model systems that can be represented as a collection of beams and rigid bodies that are connected through constraints. Degrees of freedom are defined in the inertial/global frame of reference and include displacements and rotations (formally as rotation matrices, but stored as quaternions). The underlying formulation is built on a Lie-group time integrator designed for index-3 differential-algebraic equations, which is second-order accurate in time (Bruls et al., 2012). Beam models are based on geometrically exact beam theory and are discretized as high-order spectral finite elements similar to those in BeamDyn (Wang et al., 2017). The governing equations for a FMD system like a wind turbine constitute a highly nonlinear system of constrained partial-differential equations. Kynema-FMB uses analytical Jacobians in the nonlinear-system solves in each time step. Linear systems use sparse storage and several third-party sparse-linear-system solvers are enabled. Ill conditioning of linear systems is mitigated with preconditioning described in Bottasso et al, 2008. Kynema-FMB is integrated with a simple open-source controller (ROSCO). There is an application programming interface (API) for coupling to geometry-resolved CFD (like that in Sharma et al., 2023) and actuator-force CFD (like that in Kuhn et al., 2025). In the latter, for actuator-line models, Kynema-FMB includes an internal blade-element solver that depends on user-provided lookup tables for coefficients of lift and drag, i.e., aerodynamic polars. Kynema-FMB is written in C++ and leverages Kokkos and Kokkos-Kernels (KokkosEcosystem) as its performance portability layer enabling simulations on both CPU and GPU systems. The repository is equipped with extensive automated testing at the unit and regression/system levels. The following describes the high-level development objectives conceived for Kynema: *Kynema will follow modern software development best practices, including test-driven development (TDD), version control, hierarchical automated testing, and continuous integration (CI) for a robust development environment. *The core data structures are memory efficient and enable vectorization and parallelization at multiple levels. *Data structures are data-oriented to exploit methods for accelerated computing including high utilization of chip resources (e.g., single instruction multiple data (SIMD) instruction sets) and parallelization using GP-GPUs. *The computational algorithms incorporate robust open-source libraries for mathematical operations, resource allocation, and data management. *The API design considers multiple stakeholder needs and ensure integration with existing and future ecosystems for data science, machine learning, and AI. *Kynema-FMB is written in modern C++ and leverages Kokkos as its performance-portability library with inspiration from the kynema stack.

Sprague, MichaelA.↗

DOE OSTI2015

Minimum critical masses and their uncertainties for several fissile nuclides and reflectors - 14296

A systematic effort was undertaken to reevaluate the minimum critical mass for several fissile nuclides and reflectors. This was done using MCNP6 in combination with several modern nuclear data libraries, viz. ENDF/B-VII.1, JENDL-4.0 and JEFF-3.1.1. This combination was benchmarked extensively against criticality safety benchmarks. The procedure to extract the minimum critical mass from the MCNP6 calculations was carefully implemented, so that it would have a negligible impact on the accuracy of the end results. All parts of the calculation, i.e. nuclear data processing, MCNP6 models, and post-processing, were applied consistently for all cases, involving varying fissile nuclides and reflectors. Results for the minimum critical mass were generated for {sup 233,235}U, {sup 236,239,241}Pu, {sup 242m}Am, {sup 243,245,247}Cm, and {sup 249,251}Cf, for the case of water moderation and water reflection. For {sup 235}U, the minimum critical mass was also calculated for cases with other reflectors, such as beryllium, heavy water, graphite, lead, and concrete. The results for the water reflection cases are in good agreement with most of the literature values, but for the cases with reflectors such as graphite, lead, or concrete, there are sizable differences. The accuracy of the results was assessed in two ways. First, a comparison was made with criticality safety benchmark results, by selecting benchmarks that resemble the minimum mass systems the most. It was clear that for {sup 233,235}U and {sup 239}Pu with water reflection, the differences in minimum critical masses between the libraries are consistent with the differences obtained for benchmark cases. The difference between the benchmark calculations and the benchmark values is then an indication of the accuracy with which one can simulate comparable systems. Secondly, sensitivity runs were performed with the new MCNP6 sensitivity option, based on ENDF/B-VII.1 nuclear data. For each fissile nuclide, the sensitivity of k{sub eff} to cross sections such as capture, fission, elastic, inelastic was calculated for the minimum critical mass case. The same was done for the {sup 1}H and {sup 16}O cross sections in the moderator and reflector. The uncertainty in k{sub eff} was calculated by combining the sensitivity profiles with the appropriate ENDF/B-VII.1 covariance matrix, using the SUSD code. As a final step, the uncertainty in k{sub eff} can be translated to an uncertainty in the minimum critical mass. (author)

42 ENGINEERING↗

DOE OSTI2016

SCALE Code System

The SCALE Code System is a widely-used modeling and simulation suite for nuclear safety analysis and design that is developed, maintained, tested, and managed by the Reactor and Nuclear Systems Division (RNSD) of Oak Ridge National Laboratory (ORNL). SCALE provides a comprehensive, verified and validated, user-friendly tool set for criticality safety, reactor and lattice physics, radiation shielding, spent fuel and radioactive source term characterization, and sensitivity and uncertainty analysis. Since 1980, regulators, licensees, and research institutions around the world have used SCALE for safety analysis and design. SCALE provides an integrated framework with dozens of computational modules including three deterministic and three Monte Carlo radiation transport solvers that are selected based on the desired solution strategy. SCALE includes current nuclear data libraries and problem-dependent processing tools for continuous-energy (CE) and multigroup (MG) neutronics and coupled neutron-gamma calculations, as well as activation, depletion, and decay calculations. SCALE includes unique capabilities for automated variance reduction for shielding calculations, as well as sensitivity and uncertainty analysis. SCALE’s graphical user interfaces assist with accurate system modeling, visualization of nuclear data, and convenient access to desired results.SCALE 6.2 provides many new capabilities and significant improvements of existing features.New capabilities include:• ENDF/B-VII.1 nuclear data libraries CE and MG with enhanced group structures,• Neutron covariance data based on ENDF/B-VII.1 and supplemented with ORNL data,• Covariance data for fission product yields and decay constants,• Stochastic uncertainty and correlation quantification for any SCALE sequence with Sampler,• Parallel calculations with KENO,• Problem-dependent temperature corrections for CE calculations,• CE shielding and criticality accident alarm system analysis with MAVRIC,• CE depletion with TRITON (T5-DEPL/T6-DEPL),• CE sensitivity/uncertainty analysis with TSUNAMI-3D,• Simplified and efficient LWR lattice physics with Polaris,• Large scale detailed spent fuel characterization with ORIGAMI and ORIGAMI Automator,• Advanced fission source convergence acceleration capabilities with Sourcerer,• Nuclear data library generation with AMPX, and• Integrated user interface with Fulcrum.Enhanced capabilities include:• Accurate and efficient CE Monte Carlo methods for eigenvalue and fixed source calculations,• Improved MG resonance self-shielding methodologies and data,• Resonance self-shielding with modernized and efficient XSProc integrated into most sequences,• Accelerated calculations with TRITON/NEWT (generally 4x faster than SCALE 6.1),• Spent fuel characterization with 1470 new reactor-specific libraries for ORIGEN,• Modernization of ORIGEN (Chebyshev Rational Approximation Method [CRAM] solver, API for high-performance depletion, new keyword input format)• Extension of the maximum mixture number to values well beyond the previous limit of 2147 to ~2 billion,• Nuclear data formats enabling the use of more than 999 energy groups,• Updated standard composition library to provide more accurate use of natural abundances, andvi• Numerous other enhancements for improved usability and stability.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗

DOE OSTI2026

Reassessing Double-Ended Guillotine Break Requirements: Evidence-Based Analysis of Regulatory Assumptions After Five Decades of Nuclear Operation

After five decades of nuclear power operation encompassing more than 20,000 reactor-years across 35 countries and 647 reactors, zero double-ended guillotine breaks (DEGBs) have been documented in commercial reactor coolant systems—despite DEGB being the fundamental design-basis assumption driving Emergency Core Cooling System (ECCS) sizing, structural protection requirements, and containment design specifications. This report examines the basis for DEGB requirements in nuclear power plant design. The DEGB postulate assumes the instantaneous, complete circumferential severance of the largest diameter pipes in reactor coolant systems, driving major design requirements under 10 Code of Federal Regulations 50.46, General Design Criterion 4 and containment design specifications. The United States (4,880 reactor-years) and France (2,505 reactor-years) contribute the largest operational datasets. Probabilistic assessments estimate direct DEGB occurrence probabilities with extremely low event frequencies, far below the 10-5/reactor-year thresholds typically used to define non-credible events in nuclear-safety analyses; i.e., events with probability this low fall into beyond-design-basis events. Current material-science knowledge demonstrates that the ductile steel materials used in nuclear piping systems exhibit stable crack-growth behavior fundamentally incompatible with instantaneous severance. International regulatory experience, particularly Germany’s comprehensive break-preclusion implementation, and successful leak-before-break (LBB) applications in almost all of U.S. pressurized water reactor units validate that alternatives can maintain safety performance while reducing economic burden. Current DEGB protection systems impose estimated lifetime costs of hundreds of millions of dollars per unit, over the life of a plant across the nuclear industry (including ongoing costs), representing substantial resource allocation toward scenarios with extremely low probability. Although this report acknowledges uncertainties regarding long-term aging effects, potential synergistic degradation mechanisms, and site-specific seismic considerations that warrant continued evaluation as regulatory policy evolves, there remains no documented evidence that a DEGB has occurred as a consequence of the conditions or mechanisms described in this report. This report acknowledges the Nuclear Regulatory Commission’s (NRC’s) recent efforts—outlined in the draft Interim Staff Guidance (ISG) NRC-DSS-ISG-2025-XX (“Treatment of Certain Loss-of-Coolant Accident Locations as Beyond-Design-Basis Accidents Draft Interim Staff Guidance”)—to reduce overly conservative requirements for large-break loss of coolant accidents through technical justifications and exemptions. However, extensive operating experience and validated methodologies—such as LBB and in-service inspection programs—demonstrate that the probability of a DEGB in reactor coolant-loop piping is extremely low, even under seismic conditions. The authors and reviewers of this report recommend that DEGB be removed as a design-basis event through formal rulemaking, rather than case-by-case exemptions, to better reflect credible failure modes, align with current data, and align with modern, risk-informed safety analysis.

22 - GENERAL STUDIES OF NUCLEAR REACTORS↗