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All sourcesNASA NTRSDOE OSTI
DOE OSTI2022

Constraining the Nucleon Size with Relativistic Nuclear Collisions

The notion of the "size" of nucleons and their constituents plays a pivotal role in the current paradigm of the formation and the fluctuations of the quark-gluon plasma produced in high-energy nuclear collision experiments. Here, we report on state-of-the-art hydrodynamic results showing that the correlation between anisotropic flow, $v^{2}_{n}$, and the mean transverse momentum of hadrons, [$p_t$], possesses a unique sensitivity to the nucleon size in off-central heavy-ion collisions. We argue that existing experimental measurements of this observable support a picture where the relevant length scale characterizing the colliding nucleons is of order 0.5 fm or smaller, and we discuss the broad implications of this finding for future global Bayesian analyses aimed at extracting initial state and medium properties from nucleus-nucleus collision data, including $v^{2}_{n}$-[$p_t$] correlations. Determinations of the nucleon size in heavy-ion collisions will provide a solid independent constraint on the initial state of small system collisions, and will establish a deep connection between collective flow data in nucleus-nucleus experiments and data on deep inelastic scattering on protons and nuclei.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2024

Human Factors Considerations in Artificial Intelligence Applications for Nuclear Power Plants

In recent years, there has been a wave of artificial intelligence (AI) technologies that offer to solve problems from shopping habits to mortgage approvals to critical systems operations. The rapidity of the development of these systems has led to both excitement and apprehension about the roles these systems should play in our modern societies. Furthermore, this paper focuses on the critical infrastructure industry, in general, and nuclear power generation, in particular, and seeks to scrutinize how we can leverage these novel technologies in human-centered ways to maintain or enhance the established high levels of reliability and resilience in these industries. First, we discuss the broader aspects of cognitive systems and activities that are critical to understanding the human-AI space. Then we explore different approaches to explainability in AI and the notions of trust. We then move on to discuss several human factors concepts and methods and how they can support the design of human-AI teams. We then explore recent research related to nuclear power that has been undertaken and evaluate the current industry and regulatory landscapes. Finally, we discuss identified research gaps and recommendations for solving these for the critical infrastructure space.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2021

R -mode Stability of GW190814's Secondary Component as a Supermassive and Superfast Pulsar

The nature of GW190814's secondary component m 2 of mass (2.50–2.67)M ⊙ in the mass gap between the currently known maximum mass of neutron stars and the minimum mass of black holes is currently under hot debate. Among the many possibilities proposed in the literature, the m 2 was suggested as a superfast pulsar while its r-mode stability against the run-away gravitational radiation through the Chandrasekhar-Friedman-Schutz mechanism is still unknown. Previously, Fortin et al. constructed a sample of 33 unified equations of state using the same nuclear interactions from the crust to the core consistently; from that sample we use those equations that fulfill all currently known astrophysical and nuclear physics constraints to compare the minimum frequency required for m 2 to rotationally sustain a mass greater than 2.50 M ⊙ with the critical frequency above which the r-mode instability occurs. Here, we use two extreme damping models assuming that the crust is either perfectly rigid or elastic. Using the stability of 19 observed low-mass X-ray binaries as an indication that the rigid crust damping of the r-mode dominates within the models studied, we find that m 2 is r-mode-stable while rotating with a frequency higher than 870.2 Hz (0.744 times its Kepler frequency of 1169.6 Hz) as long as its temperature is lower than about 3.9 × 10 7 K, further supporting the proposal that GW190814's secondary component is a supermassive and superfast pulsar.

79 ASTRONOMY AND ASTROPHYSICS↗
DOE OSTI2025

Mechanistic Fission Gas Release Uncertainty Induced by Microstructure Data

Fission Gas Release (FGR) is an important engineering safety parameter for nuclear fuel. While fuel performance modeling with BISON currently relies on mechanistic models to predict it, comparison with experimental data shows both under or over prediction depending on operation mode (steady or transient).Predicting microstructure data is essential to accurately predicts the engineering scale parameters. An important source of uncertainty in mechanistic models arises from the missing captured physics. Continuous validation and refinement of these models against experimental data are also necessary to ensure their reliability and accuracy in predicting engineering parameters.

36 - MATERIALS SCIENCE↗
DOE OSTI2024

DRiFT current mode, trigger settings and flexible detector specifications applied to scintillator arrays

MCNP radiation transport output is post-processed by DRiFT, a Detector Response Function Toolkit to simulate detailed nuclear instrumentation response. DRiFT can be used to assess the performance and potential limitations of scintillator, gas, and semiconductor detectors under a variety of simulated conditions not easily achievable in a laboratory setting. This work describes new updates in DRiFT for scintillator simulations which focus on the capability to simulate scintillators in current mode, an expansion of trigger options, and the ability to customize individual detector properties in a simulation. These improvements are designed to facilitate the ability to model large arrays of scintillator detectors with higher fidelity than was previously possible and are demonstrated in three examples. The first shows the difference between operating DRiFT in current and pulse mode. In the second example, which is intended to demonstrate deviations in individual detector performance, each detector has properties (PMT gain, optical transport, scintillation yield, etc.) that vary between detectors and are specified in DRiFT. A final example examines how DRiFT could be used to optimize digitizer settings in high rate measurements with split signals using the new common trigger option.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2020

Comparison of simulated and experimentally measured detector impulse responses to pulsed x-rays

The Gamma Array Simulation Toolkit (GAST) is a multi-physics software stack comprised of Geant4, MATLAB, and experimentally measured photomultiplier tube (PMT) impulse response waveforms that is used to model and inform the end-toend performance of radiation detector systems. In order to validate the ability of GAST to predict the impulse response of a modeled detector configuration operated in current mode, experimental measurements using the Nevada National Security Site Transformational Diagnostics and Imaging 2 MeV endpoint energy electron linear accelerator (linac) with a tungsten target were performed. Seven different detector configurations consisting of unique combinations of scintillator material, size, reflectivity conditions, presence or absence of light guide, and PMT were evaluated. The resultant impulse response of each detector configuration was recorded and compared to a GAST simulation that emulated the respective linac experiment conditions. The comparison of the experimental and simulated detector impulse responses indicated that GAST can reliably predict detector impulse responses.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2020

NNSA-IAEC Science Area V Environmental ISR: Waste Management and Subsurface Science (Final Report FY18-FY20)

In August 2017, the U.S. National Nuclear Security Administration (NNSA) and the Israel Atomic Energy Commission (IAEC) launched a joint waste management program to evaluate the feasibility of siting a radioactive waste repository at intermediate depths in Israel. The bilateral collaboration involves three National Laboratories (LLNL, LANL, and Sandia), and the Nuclear Research Center, Negev, Geological Survey of Israel, and Ben Gurion University in Israel. This research is designed to assist with the evaluation of locations for an intermediate nuclear waste repository in Israel. This final report is a high-level summary of the results from the past three years by the team that worked on the WM-2: Radionuclide colloid-facilitated transport in fractured carbonate rock project for Science Area V, Environmental ISR: Subsurface Science and Waste Management. We have completed all of our tasks and met all of our milestones. MW-2 had three integrated tasks. Task 1 was laboratory experiments investigating colloid transport of radionuclides in natural fractured carbonate rocks from the Avdat formation, Israel. Task 2 was field experiments injecting radionuclide analogues pre-sorbed to clay colloids into fractured carbonate rock, and task 3 was numerical modeling of laboratory and field experimental results to assess the importance of colloids in the migration of relevant radionuclides. This task was incorporated into tasks 1 and 2. A 4 th task was initiated in FY20, to investigate the role of organics in facilitating radionuclide transport under the same conditions explored in task 1 due to the high organic content of the rocks being investigated for the immediate borehole locations. Preliminary results will be summarized here and the work will continue in FY21-23. The overall objective of this research is to evaluate the role of colloids (naturally occurring < 1 micron particles) in facilitating the transport of long-lived radionuclides from a nuclear waste repository situated in fractured carbonate rocks. Currently little data exists on radionuclide transport in carbonate rocks, but this is the main rock type available for siting a repository in Israel. Laboratory experiments were carried out in both Israel and the U.S., field experiments have taken place in Israel, and reactive transport modeling involved LLNL, LANL and Israel partner institutions. LLNL hosted a graduate student, Emily Tran, from Ben Gurion University in FY18 and FY19 and much of the work presented here was part of her PhD research.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2023

Ab Initio Evaluation of Plutonium Dioxide $S(α,β)$ and Thermal Neutron Cross Sections

Plutonium Dioxide (PuO 2 ) is an advanced fuel-cycle nuclear fuel material for thermal and fast reactors, with a melting point at 3261.8 K and thermal conductivity above 3 Wm -1 K -1 for temperatures below 1500 K. Plutonium from spent UO 2 is reprocessed to form chemically stable mixed oxide (MOX) fuel. In contrast to metallic fuels, the low PuO 2 conductivity results in high thermal gradients between fuel center and surface, resulting in structural annealing and efficient fission product migration over burnup. Crystal binding affects the doppler broadening of epithermal resonances for nuclear fuels and scattering behavior of thermal neutrons. Currently, there is no ENDF/B Thermal Scattering Law, i.e., $S(α,β)$, evaluation for PuO 2 , only free atom cross sections for Plutonium and Oxygen, respectively, which neglect lattice contributions to total cross section. Herein, ab initio lattice dynamics (AILD) techniques are employed to calculate the phonon density of states (DOS) using spin-orbit-coupling density functional theory (DFT) to predict the PuO2 paramagnetic ground-state structure. These results highlight the evaluation of $S(α,β)$ for Plutonium and Oxygen, respectively in PuO 2 and consequential generation of thermal neutron scattering cross sections for high fidelity criticality safety analysis and reactor calculations.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Comparison of Simulated and Experimentally Measured Detector Impulse Responses to Pulsed X-rays

The Gamma Array Simulation Toolkit (GAST) is a multi-physics software stack comprised of Geant4, MATLAB, and experimentally measured photomultiplier tube (PMT) impulse response waveforms that is used to model and inform the end-to-end performance of radiation detector systems. In order to validate the ability of GAST to predict the impulse response of a modeled detector configuration operated in current mode, experimental measurements using the Nevada National Security Site Transformational Diagnostics and Imaging 2 MeV endpoint energy electron linear accelerator (linac) with a tungsten target were performed. Seven different detector configurations consisting of unique combinations of scintillator material, size, reflectivity conditions, presence or absence of light guide, and PMT were evaluated. The resultant impulse response of each detector configuration was recorded and compared to a GAST simulation that emulated the respective linac experiment conditions. The comparison of the experimental and simulated detector impulse responses indicated that GAST can reliably predict detector impulse responses.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2025

The inorganic chemist's guide to actinide radiation chemistry: a review

This review aims to provide an overview of the current state of radiation chemistry with respect to the actinide elements, thorium through californium. Despite the inherent radioactivity of the actinides, only a few studies explore the effects of ionizing radiation on their redox chemistry and surrounding environment. This fundamental knowledge gap, coupled with the current renaissance in actinide-based technologies such as nuclear power, space exploration, and medicine, underscores the importance of research in this interdisciplinary area. This review will focus on the interactions between reactive species formed by radiolysis with actinides and their complexes, offering an inorganic chemist's perspective on research in radiation chemistry. In addition, a thorough discussion of our current understanding of radiation-induced changes in actinide speciation in both aqueous solution and the solid-state will be provided, focusing on changes in oxidation state distribution, complexation, and secondary coordination effects within inorganic materials. Finally, this review will discuss challenges and opportunities for inorganic chemists to explore this unique intersection of fields.

Actinides↗
DOE OSTI2021

Lightning Burnthrough to Containment Breach of 55-Gallon TRU Waste Drums

We investigated by arc-plasma heating the feasibility of attributing inherent lightning protection to 55-gallon DOT 7A, Type A, open head carbon steel drums made of 1.5 millimeter painted carbon steel, designed to protect Department of Energy transuranic nuclear waste. The Sandia Lightning Simulator transferred continuing current in 300 ampere (A), 400 A, and 500 A tests to achieve a 350 coulomb charge transfer and simulate cloud-to-ground lightning attachment to test coupons and 9 drums. A tungsten electrode was placed 0.75 inch from the drums. High-speed photography was recorded to observe the exterior containment breach, or "first light," seen on camera when burnthrough opened a hole in the containment. Sheet metal burnthrough occurred between 18 and 71 coulombs in lid and rolling hoop tests, but 12-gauge closure ring tests did not result in burnthrough, which suggests this feature may provide an inherent air terminal protective feature.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2021

Preliminary Modeling of Chloride Deposition on Spent Nuclear Fuel Canisters in Dry Storage Relevant to Stress Corrosion Cracking

This study presents a method that can be used to gain information relevant to determining the corrosion risk for spent nuclear fuel (SNF) canisters during extended dry storage. Currently, it is known that stainless steel canisters are susceptible to chloride-induced stress corrosion cracking (CISCC). However, the rate of CISCC degradation and the likelihood that it could lead to a through-wall crack is unknown. This study uses well-developed computational fluid dynamics and particle-tracking tools and applies them to SNF storage to determine the rate of deposition on canisters. The deposition rate is determined for a vertical canister system and a horizontal canister system, at various decay heat rates with a uniform particle size distribution, ranging from 0.25 to 25 µm, used as an input. In all cases, most of the dust entering the overpack passed through without depositing. Most of what was retained in the overpack was deposited on overpack surfaces (e.g., inlet and outlet vents); only a small fraction was deposited on the canister itself. These results are provided for generalized canister systems with a generalized input; as such, this paper is intended to demonstrate the technique. This study is a part of an ongoing effort funded by the U.S. Department of Energy, Nuclear Energy Office of Spent Fuel Waste Science and Technology, which is tasked with doing research relevant to developing a sound technical basis for ensuring the safe extended storage and subsequent transport of SNF. This work is being presented to demonstrate a potentially useful technique for SNF canister vendors, utilities, regulators, and stakeholders to utilize and further develop for their own designs and site-specific studies.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

BISON Capability, Validation and Demonstration for Reactivity-Initiated Accidents

Reactivity Initiated Accidents (RIA) are design basis accidents that could have adverse impacts on the core coolability. In the unlikely event that sufficient reactivity is inserted into the reactor core by the ejected/dropped control rod, prompt energy deposition into the fuel can occur, which when sufficiently high can lead to fuel rod failure or, at large energy deposition levels, expulsion of UO2 fragments or molten UO2 material from the fuel rod. This results in a release of fission product and fuel into the coolant and has the potential to compromise core coolability and threatening the pressure boundary of the primary coolant system. The design basis RIA is one of the industry challenging problems that the CASL aimed to address; the CASL RIA Challenge Problem Charter [1] states, ?The Pressurized Water Reactor (PWR) Rod Ejection Accident (REA) and Boiling Water Reactor (BWR) Control Rod Drop Accident (CRDA) are postulated accidents with consequences that are important to nuclear safety (fuel rod integrity and core coolability). Currently each reload core design must be analyzed to meet regulatory acceptance criteria. The goal of CASLs ModSim capability for RIA is to model the event at a higher fidelity, with validation to existing tests, to better model the transient neutronics and the progression of the fuel and cladding thermal-mechanical behavior. These improved analytical capabilities can be used to better inform reload core design, limits on fuel assembly discharge burnup, restrictions on placement of fuel in the reactor, control rod insertion limits, operating margin, and performance sensitivities.? In support of that charter, BISON, the fuel performance code has been used to demonstrate the simulation of thermal-mechanical behavior of LWR fuels during a reactivity-initiated accident. The combination of mechanical, thermal, and thermal-hydraulic phenomena all present during an RIA makes a multi-physics code such as BISON a valuable tool for modeling these scenarios. This paper will highlight many of the activities associated with BISON relevant to reactivity-initiated accident capability development and validation efforts. These efforts have been performed under both CASL and the Nuclear Energy Advanced Modeling and Simulation (NEAMS) programs.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Guidance on Including Social, Organizational and Technical Influences in Nuclear Utility and Plant Modernization Plans

The objective of this report is to provide the nuclear power industry with guidance on including socio-technical issues in plans developed for current and future modernization efforts. Socio-technical issues of concern include social, organizational and related technical aspects of modernized systems. These can be addressed using a set of systematic analysis and design methods developed to address socio-technical issues of the sort facing the nuclear power industry. The next phase of this work will involve developing and implementing a modernization plan that includes socio-technical tools and methods identified and discussed in a report by Dainoff and colleagues (2020). The plan is expected to be developed and implemented in cooperation with a nuclear utility that has expressed interest in participating. The Dainoff et al report provides the foundation for the present effort and presents a methodological tool set for analyzing socio-technical issues applicable to many modernization efforts. The tools described by Dainoff et al are based on principles of human factors engineering, safety engineering, and systems engineering and are discussed within the specific context of nuclear utility modernization and transformation efforts. It is expected that the plan and the concepts described in the current report will be developed and implemented with a nuclear utility partner. The plan developed with the nuclear utility partner will have a substantially higher level of detail than the guidance provided in this report, particularly with regard to details of plan logistics, etc.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2020

Preliminary Plan for Evaluation of Reactor Pressure Vessel Surveillance Materials from Palisades Nuclear Generating Station

The Palisades Nuclear Generating Station (PNGS), located in Michigan, is owned and operated by Entergy. It is a Combustion Engineering 2-loop pressurized water reactor (PWR) producing 805 MWe (2,565 MWth). The PNGS was built between 1967 and 1970, with approval to operate at full power in 1973; the plant’s original license was due to expire on March 24, 2011. An application for 20-year extension was filed in 2005 with the Nuclear Regulatory Commission and was granted on January 18, 2007. Although the plant was scheduled for decommissioning by 2031, Entergy currently plans to close the PNGS in 2022. The PNGS included in its reactor pressure vessel (RPV) surveillance program a capsule, designated A-60, containing specimens of one of the vessel plates and a weld metal with nickel content of about 1.36 wt% and copper content of about 0.20 wt%. This capsule was irradiated to a fluence of 1.8 x 10 20 n/cm 2 . The capsule was removed from its surveillance position in 1995 and has been resident in the spent fuel pool since that time. The material is also of special interest because of its very high nickel content and because of the potential for development of NiMnSi (nickel-manganese-silicon) precipitates, dubbed “late blooming phases.” The surveillance program also includes a capsule, designated T-150, dedicated specifically for thermal aging, which would provide results for at least 33 effective full power years, which is beyond the current thermal-aging database for such materials. Given that license extensions to 60 years of operation have been approved by the US Nuclear Regulatory Commission for most of the currently operating light water reactors in the United States, and that the first extension to 80 years was recently approved, there exists the probability that some RPVs will reach and possibly exceed a fast neutron fluence (> 1 MeV) of 1 x 10 20 n/cm 2 . This is a fluence regime with no US surveillance data and very little test reactor data, except for the Light Water Reactor Sustainability Program–sponsored University of California Santa Barbara Advanced Test Reactor (ATR) ATR-2 project. Thus, the materials in the A-60 capsule represent a valuable resource for directly exploring the effects of commercial surveillance irradiation on a typical plate and a high-nickel weld with similar materials irradiated in the ATR-2 test reactor project. This report provides background information for the surveillance program, previous results of surveillance materials testing, the preliminary plan for capsule retrieval and disassembly, and the plan for testing and microstructural examination of the mechanical test specimens of these unique materials to assess the features induced by very high irradiation fluence or very long thermal aging time

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Prediction of strain, inter-layer interaction and critical current in CORC ® wires under axial strain by T-A modeling

Superconducting conductors on round core (CORC ® ) cables and wires can meet the needs of large high-field magnets, such as particle accelerators and compact nuclear fusion machines, due to their simple cabling process, high current-carrying capacity and reliable operation under high mechanical stresses. Many high-field magnets require CORC ® cables to carry a current of thousands of amperes in a background magnetic field exceeding 20 T. As a result, the large electromagnetic forces will deform the cable in the axial direction due to hoop stress and in the transverse direction by compressive stress. Therefore, it is essential to determine the irreversible deformation limit of the CORC ® cable under axial tensile load and optimize the cabling parameters to potentially extend this limit. Analytical and numerical methods are developed to assess the performance degradation of CORC ® wires under axial tensile load. The strain level, interlayer contact pressure and friction and their impact on the critical current are calculated by combining the mechanical response and the T-A method. Analyzing the results shows that the winding angle of the tape and the Poisson’s ratio of the inner core are key factors affecting the irreversible tensile strain limit of CORC ® wires. The smaller the winding angle and the higher the Poisson’s ratio of the inner core, the higher the irreversible tensile strain limit. For multi-layer CORC ® wires, the initial contact pressure caused by the cabling process must also be considered. The inter-layer interaction is coupled with the tape strain of each layer. The results of this research can serve as a basis for optimizing and designing CORC ® wires with extended irreversible strain limits.

75 CONDENSED MATTER PHYSICS, SUPERCONDUCTIVITY AND↗
DOE OSTI2020

Fire Risk Investigation in 3D (FRI3D) Software and Process for Integrated Fire Modeling

Modeling and implementing fire safety for nuclear power plants is a costly activity. Because of the complexity of fire phenomena and multiple operational procedures, it is difficult to computationally provide assurance that mitigation methods are adequate for critical areas using current analysis methods in a cost-effective manner. The nuclear industry needs efficient methods to provide more accurate modeling and optimize mitigation methods to improve nuclear power plant understanding for fire scenarios. This report describes the work done over the last two year on the Fire Risk Investigation in 3D (FRI3D) software. This software simplifies the modeling process by importing existing plant models and data, incorporating a 3D modeling environment and coupling with the fire simulation software CFAST. This eliminates the manual labor and errors associated with much of the fire modeling process. The report lays out the status and path of FRID software for the Risk Informed Safety Analysis pathway of the Light Water Reactor Sustainability project at Idaho National Laboratory and industry collaborators to best help reduce costs and improve realism in fire probabilistic risk analysis.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2023

pyEGAF: An open-source Python library for the Evaluated Gamma-ray Activation File

The Evaluated Gamma-ray Activation File (EGAF) is one of the most comprehensive resources for thermal neutron-capture data. This database contains data from prompt gamma activation analysis measurements carried out in a consistent manner using the same experimental configuration at the Budapest Research Reactor for 245 isotopes. Although these valuable datasets have been freely available for many years, one of the drawbacks is the outdated and cryptic Evaluated Nuclear Structure Data File (ENSDF) format that is currently adopted for dissemination, making it difficult for users unfamiliar with the format to access and utilize the data contained therein. Furthermore, the ENSDF format does not readily lend itself to modern computational technologies and a parser is required to interpret the complicated mixed-record format. To help overcome these challenges, we have developed a translator to convert the ENSDF-formatted datasets into an open standard JavaScript Object Notation (JSON) format enabling accessibility to applications using different programming languages running in different environments. To compliment this effort, we have also developed an open-source software package implemented in Python, pyEGAF, that is designed to interact with the JSON data structures for general purpose access, manipulation, and analysis of the neutron-capture $\gamma$-ray data in EGAF. The new format, together with the pyEGAF library, greatly enhances access to the wider applications community where EGAF data may be useful or is required.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗