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DOE OSTI2023

Atomistic‐Level Effects of Noncovalent Interactions and Crystalline Packing for Organic Material Structural Integrity upon Exposure to Gamma Radiation

Developing an atomistic understanding of ionizing radiation induced changes to organic materials is necessary for intentional design of greener and more sustainable materials for radiation shielding and detection. Cocrystals are promising for these purposes, but a detailed understanding of how the specific intermolecular interactions within the lattice upon exposure to radiation affect the structural stability of the organic crystalline material is unknown. This study evaluates atomistic-level effects of γ radiation on both single- and multicomponent organic crystalline materials and how specific noncovalent interactions and packing within the crystalline lattice enhance structural stability. Dose studies were performed on all crystalline systems and evaluated via experimental and computational methods. Changes in crystallinity were evaluated by p-XRD and free radical formation was analyzed via EPR spectroscopy. Type of intermolecular interactions and packing within the crystal lattice was delineated and related to the specific free radical species formed and the structural integrity of each material. Periodic DFT and HOMO-LUMO surface mapping calculations provided atomistic-level identifications of the most probable sites for the radicals formed upon exposure to γ radiation and relate intermolecular interactions and molecular packing within the crystalline lattice to experimental results.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

DOE OSTI2022

A 100-Mrad (Si) JFET-based Sensing and Communications System for Extreme Nuclear Instrumentation Environments

Dry cask storage is one of two storage methods approved by the U.S. Nuclear Regulatory Commission for spent fuel after removal from reactor cores. Dry casks consist of a stainless steel canister enclosed in a concrete overpack to contain the hazardous radioactive spent fuel rods and provide radiation shielding. Monitoring spent fuel storage casks is desired to ensure the safe containment of the enclosed spent fuel, but is very difficult due to the related harsh temperature and radiation environment. The sensors and associated electronics to monitor temperature, pressure, and/or radiation need to survive high temperatures and radiation doses for extended time periods. For this reason, there is a severe need for radiation-hardened electrical systems that survive well beyond the existing capabilities of commercially available radiation-rated electronic components, which have primarily been developed for space applications. Junction-gate field-effect transistor (JFET) devices are inherently radiation hardened [exceeding 100 Mrad (Si)]. Additionally, when JFETs are used as building blocks for sensing and communication electronics (i.e., oscillators, amplifiers, filters, and mixers), inherently radiation-hardened circuits can be achieved. To this end, JFET-based radiation-hardened electronics interfacing with cask-embedded sensors capable of driving modulated sensor signals through a stainless steel barrier were designed and tested at a dose rate of approximately 500 krad/h (Si) to beyond a 200-Mrad (Si) total ionizing dose. After 200 Mrad (Si), the sensor and communication circuit signals were correctly decoded at the receiver despite oscillator drift. The results from this experiment demonstrate the potential for creating more complex radiation-hardened JFET-based electrical systems for nuclear environments.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

DOE OSTI2023

SCALE 6.3.1 User Manual

SCALE is a comprehensive modeling and simulation suite for nuclear safety analysis and design developed and maintained by Oak Ridge National Laboratory under contract with the U.S. Nuclear Regulatory Commission, U.S. Department of Energy, and the National Nuclear Security Administration to perform reactor physics, criticality safety, radiation shielding, and spent fuel characterization for nuclear facilities and transportation/storage package designs.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2024

Isotopic gamma lines for identification of shielding materials

Identifying the constituting materials of concealed objects is crucial in a wide range of sectors, such as medical imaging, geophysics, nonproliferation, national security investigations, and so on. Existing methods face limitations, particularly when multiple materials are involved or when there are challenges posed by scattered radiation and large areal mass. Here we introduce a novel brute-force statistical approach for material identification using high spectral resolution detectors, such as HPGe. The method relies upon updated semianalytic formulae for computing uncollided flux from source of gamma radiation, shielded by a sequence of nested spherical or cylindrical materials. These semianalytical formulae make possible rapid flux estimation for material characterization via combinatorial search through all possible combinations of materials, using a high-resolution HPGe counting detector. An important prerequisite for the method is that the geometry of the objects is known (for example, from X-ray radiography). We demonstrate the viability of this material characterization technique in several use cases with both simulated and experimental data in spherical geometry.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗

DOE OSTI2021

SCALE 6.3.0 User Manual

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2021

Monte Carlo Simulations in Support of NorthStar Irradiation Facility Commissioning Procedure

NorthStar Medical Radioisotopes LLC is constructing an irradiation facility at which electron accelerators will induce photo-transmutation (i.e., neutron knockout) of molybdenum-100 (Mo-100) to produce medically useful Mo-99. The facility will house two high-intensity electron RhodotronTM accelerators developed by IBA Industrial. The nominal power of each Rhodotron is 125kW at 40 MeV beam energy. The production target is made of metallic molybdenum enriched with Mo-100 isotope. It will be irradiated by electron beams and will itself play a role in electron conversion to intense X-rays. A full description of the NorthStar target as well as the concept of radiation shielding of the facility can be found in our previous report. The NorthStar irradiation facility is under commissioning now. The goal of the present work is to support the commissioning process by: (1) Studying residual radiation and activation to minimize personnel exposure and inform radiation waste management; (2) Using computational fluid dynamics (CFD) of the irradiated target to ensure the facility design is safe and reliable; and (3) Helping to determine radiation protection requirements during operation. Monte Carlo simulations are used for the analyses. Most of the results are presented as 3D arrays representing spatial distribution of values across simulated geometries: radiation energy depositions, ambient doses, residual isotopes accumulated, and residual doses.

07 ISOTOPE AND RADIATION SOURCES↗

DOE OSTI2023

SPC-70646 Specification for Reactor Supplemental Shielding for Use in DOME

NRIC is developing the Demonstration of Microreactor Experiments (DOME) test bed at the Idaho National Laboratory (INL) Materials and Fuels Complex (MFC): to allow for testing of advanced reactors in support of future licensing and commercial operations. One such need is a neutron and photon radiation shielding system to protect personnel and equipment from harsh neutron and photon fluxes during reactor operations and post shut down operations, disassembly, and decommissioning.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2020

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.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2026

Optics and Systems Design of the Ring-to-Second Target Transport Beam-Line for the SNS Second Target Station

The Second Target Station (STS) project at the Spallation Neutron Source (SNS) is being developed to provide world-leading cold neutron brightness for next-generation neutron scattering experiments. The STS Accelerator Systems (AS) scope includes the design and implementation of the Ring-to-Second Target (RTST) proton beam transport line, which extracts 1.3 GeV proton beam pulses from the existing Ring-to-Beam Transport (RTBT) system and delivers them to the STS target. The RTST design emphasizes operational reliability [high reliability], low activation [minimum activation of components and the tunnel], maintainability, and compatibility with existing SNS infrastructure through extensive reuse of proven RTBT systems and components. The beamline includes a new extraction region, a transport lattice consisting of dipole, quadrupole, and corrector magnets, beam instrumentation systems, vacuum systems, personnel protection systems, and radiation shielding systems. Beam optics and particle tracking studies were performed using PyORBIT to validate extraction trajectories, beam transport, and target beam spot requirements [60–90 cm² beam spot area]. This paper presents the optics design philosophy, extraction system architecture, transport lattice design, instrumentation strategy, vacuum system approach, and radiation protection integration for the RTST beamline. Particle tracking simulations indicate successful beam transport without beam loss under nominal operating conditions. The RTST is designed to transport 1.3 GeV proton beam pulses at repetition rates up to 15 Hz, delivering nominal beam power of 700 kW to the Second Target Station.

Baron, Alex [ORNL]↗

DOE OSTI2026

Development of a broadband hard x-ray radiography platform for pulsed-power experiments

In this article, we develop and demonstrate a broadband hard x-ray radiography platform at the Zebra Pulsed Power Laboratory that integrates point-projection radiography, bremsstrahlung measurements, and hard x-ray pinhole imaging, designed to diagnose current-driven, cylindrically compressed matter. Initial laser-pulsed-power coupled experiments revealed that intense background radiation generated during 1 MA Zebra current shots overwhelmed laser-produced hard x-rays, obscuring radiographic images. Using combined spectral and spatial diagnostics, we identify energetic electrons accelerated by return currents as the dominant source of background hard x-rays, with electron energies inferred to be 3–4 MeV based on Monte Carlo simulations, and demonstrate mitigation through modifications to the radiation shielding and return-current configuration. The diagnostic platform was validated using a wire-pinch hard x-ray source, allowing radiographs of static 1-mm-diameter aluminum wires to be obtained while simultaneously measuring x-ray source spectra and spatial emission distributions within a single shot. Measured wire transmission profiles were quantitatively reconstructed using radiation transport simulations that incorporate an experimentally inferred two-temperature exponential x-ray spectrum from bremsstrahlung signal analysis and spatially distributed emission sources identified by pinhole imaging. Agreement between measured and simulated transmission profiles demonstrates the validity of the radiographic and x-ray source characterization approach, establishing this diagnostic platform as a promising tool for diagnosing magnetically driven, high-density plasmas relevant to warm dense matter and inertial fusion energy research.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2024

SCALE 6.3.2 User Manual

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2024

SCALE Demonstration for Sodium-Cooled Fast Reactor Fuel Cycle Analysis

In support of the US Nuclear Regulatory Commission non-light-water reactor fuel cycle demonstration project, SCALE 6.3.1 capabilities for radionuclide characterization, criticality, and shielding were demonstrated for scenarios in the sodium-cooled fast reactor (SFR) nuclear fuel cycle. Three postulated accident scenarios were selected for analysis in this work. As a basis for all scenarios, irradiated fuel inventories were generated using SCALE/ORIGAMI. To cover multiple SFR design choices, two different types of SFR fuel, uranium/transuranic-loaded and U-based fuels, were considered. For the first scenario, SCALE/MAVRIC was used to calculate the radiation shielding and dose rates inside and outside of the containment building due to a drop of a spent fuel assembly from the fuel-handling system during unloading inside the containment building. For the second scenario, potential critical configurations in an electrofiner were investigated through criticality calculations with SCALE/CSAS. For the third scenario, the activity of the waste salt from an electrorefiner was evaluated using SCALE/ORIGEN. The dose rate produced by the analyzed SFR assemblies is similar to that produced by a typical pressurized water reactor (PWR) fuel assembly with a discharge burnup of 50 GWd/MTU, with the same cooling time of 10 days. The criticality analyses suggested that the different electrorefiner configurations have a large margin to criticality. The activity analysis of the electrofiner waste revealed that shielding and cooling may be required for the waste salt that contains transuranics and fission products produced by the electrorefiner because of the high activity of the waste salt. In general, the application of various capabilities in the SCALE code system for SFR fuel inventory generation, criticality, and shielding was successfully demonstrated for the selected scenarios in the SFR nuclear fuel cycle. Additional analyses can be performed to provide more accurate results when more details of the SFR nuclear fuel cycles are available, for example, the dimension of the electrorefiner and the salt compositions during reprocessing.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2025

SCALE 6.3.3 User Manual

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2022

Update of CEA DES Criticality-Safety Activities and Perspectives [Slides]

Neutronics staff at CEA DES are in charge of a whole set of modeling and simulation tools for neutronics: nuclear data evaluation and processing, transport codes development, calculation sequences edition, verification, validation and uncertainty analysis. Advanced user groups are in charge of specific nuclear analysis in reactor physics, fuel cycle, criticality-safety, radiation shielding and nuclear instrumentation for CEA and its partners, at all stages of nuclear facilities life cycle. CEA is willing to participate/collaborate on the following issues: detector design, experiments, analysis and evaluation, DH validation by combining TAGS measurements with Fission Yield variance-covariance matrices.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2020

Fused filament fabrication of polymer composites for extreme environments

Vast improvements have been made to the capabilities of advanced manufacturing (AM), yet there are still limitations on which materials can effectively be used in the technology. To this end, parts created using AM would benefit from the ability to be developed from feedstock materials incorporating additional functionality. A common three-dimensional (3D) printing polymer, acrylonitrile butadiene styrene, was combined with bismuth and polyvinylidene fluoride via a solvent treatment to fabricate multifunctional composite materials for AM. Composites of varying weight percent loadings were extruded into filaments, which were subsequently 3D printed into blocks via fused filament fabrication. Investigating the material properties demonstrated that in addition to the printed blocks successfully performing as radiation shields, the chemical, thermal, and mechanical properties are suitable for AM. Thus, this work demonstrates that it is possible to enhance AM components with augmented capabilities while not significantly altering the material properties which make AM possible.

36 MATERIALS SCIENCE↗

DOE OSTI2020

Development of the Mobile Systems for Conditioning of Disused Sealed Radioactive Sources in Serbia - 20105

Sealed radioactive sources (SRS) are being used worldwide in the field of medicine, agriculture, industry and research. They can be found in mobile as well as stationary devices. SRS contains radioactive material that is (a) permanently sealed in a capsule or (b) closely bounded and in a solid form. The capsule or material of an SRS should be strong enough to maintain leak tightness under the conditions of use and purpose for which the source was designed, also in case of accidents. In this case only emitted radiation is utilized. Firstly, the hazard from external radiation has to be considered, but the possibility of contamination due to fracture of the capsule should not be disregarded. The radioactive sources are composed of the radiating isotope contained in the filling medium, the single or double isotope holder that partially or totally surrounds the filling medium, the outer cover that contains the parts mentioned above and the capsule closed airtightly by welding or using some other method. The capsule must be tested for leakage periodically. If the SRS is no longer needed (e.g. replaced by a different technique) or it becomes useless for the intended application (e.g. the activity becomes too weak, the equipment containing the source works poorly or becomes obsolete, the source is damaged or leaking) it is considered disused. Disused sealed radioactive sources (DSRS) are typically conditioned and disposed if a facility is available. If the disposal option is not available, conditioned DSRS should be stored under proper conditions. In some cases, the radionuclide(s) in DSRS can be recovered/recycled or the DSRS can be repurposed for other applications. Conditioning of DSRS ensures containment of the radioactive material, provides confinement for leaking sources, provides sufficient radiation shielding, reduces storage/disposal volume by allowing consolidation of multiple sources into a single storage/disposal container, facilitates transport operations and contributes to safety and security as well. Typically, conditioning technologies are deployed either as permanently installed stationary systems in centralized or mobile on-site waste processing facilities, or in a mobile configuration. Centralized stationary facilities provide a single processing location for multiple users that requires transport of the waste to the facility. On the other hand, mobile systems may be provided for the selection and application of the optimum technology for a specific waste stream (such as DSRS) by bringing the process to the point where the waste is generated. In addition, mobile systems could offer additional flexibility by sharing equipment among multiple waste generating sites for processing campaigns that vary in duration, from very short periods to several years. The term 'mobile processing system' refers to any radioactive waste processing system or component which is designed to be transportable and which is not considered permanently installed. Two mobile system for conditioning of disused sealed radioactive sources are developed in the Public Company Nuclear Facilities of Serbia. Development of these mobile systems was supported by SRB9005 national project via Technical Cooperation of the IAEA. The first mobile system, built inside the 20 feet ISO container, will be used for conditioning of DSRS category 3 to 5. The second mobile system, built inside the 7 m long vehicle (Iveco Daily Van), will be used for dismantling of ionizing smoke detectors mostly with Am-241 sources. Designs of the mobile systems were defined in cooperation with two companies from Belgium (Belgoprocess and Leniko) and a Croatian company Ekoteh as well as with the support of the IAEA experts. The generic safety assessment and operational procedures for the mobile systems are developed. Based on safety assessment the acceptance criteria and operational limits and conditions are established. Operational procedures include: (a) equipment and material requirements, (b) assembling procedure of the mobile unit, (c) procedure for acceptance of devices for dismantling and conditioning, (d) dismantling procedure for devices to recover the DSRS, (e) characterization of DSRS, (f) encapsulation procedure of DSRS, (g) disassembling procedure of the mobile unit, and (h) keeping records, identification and traceability. In addition, radiation safety, health safety, security and emergency preparedness plans are prepared. The generic safety assessment and operational procedures could be updated with site specific requirements, DSRS inventory, and different needs for future customers. Developed mobile systems could be used in all situations when it is feasible to perform conditioning of DSRS on the spot in the county and worldwide. Development of these mobile units was just the first step to create the Reference Center for Radioactive Waste Treatment and Disused Radioactive Sources Conditioning for Small Facilities which can become a regional training center in the future, and/or as a tool for comprehensive national search and secure programmes. In the next phases development of e-learning platforms and blended learning packages as well as application for the IAEA Qualified Technical Centre (QTC) for the management of DSRS is foreseen. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2025

COG Software Architecture Design Description Document

This COG Software Architecture Design Description Document describes the organization and functionality of the COG Multiparticle Monte Carlo Transport Code for radiation shielding and criticality calculations, at a level of detail suitable for guiding a new code developer in the maintenance and enhancement of COG. The intended audience also includes managers and scientists and engineers who wish to have a general knowledge of how the code works. This Document is not intended for end-users.

61 RADIATION PROTECTION AND DOSIMETRY↗

DOE OSTI2026

Modeling Radiolysis and Chemical Reactions during Dry Storage of Aluminum-clad Spent Nuclear Fuel

After aluminum-clad spent nuclear fuel (ASNF) is removed from the reactor, it is initially stored in spent fuel pools, which are specially designed water-filled basins that provide temporary cooling to reduce the temperature of the fuel assemblies and provide radiation shielding. ASNF continues to generate heat due to the radioactive decay of elements within the fuel, which persists for many years post-shutdown as the residual radioactive products decay into more stable elements. During the wet storage period, an oxyhydroxide layer composed of boehmite/bayerite forms on the surfaces of the aluminum cladding from exposure to water in the pools. Road-ready packaging for long-term disposition of the ASNF involves dry storage in helium backfilled DOE standard canisters (DSCs). When the ASNF is removed from water storage and dried, most of the water is removed, but some physisorbed and chemisorbed water remains in the oxyhydroxide layers. This residual water can produce hydrogen when exposed to radiation from the ASNF during dry storage. Predicting hydrogen accumulation over time in the DSCs is critical for long-term storage considerations. Previous modeling efforts have developed coupled computational fluid dynamics (CFD)-chemical models to simulate temperature, pressure, and gas phase concentrations within the DSCs. These models use the thermal field predicted by CFD as input to a radiolysis model for the gas phase and the surface oxyhydroxide layer chemistry. Given the long storage period of the DSCs and the impracticality of long-term experiments, a simulation-based approach is necessary to assess chemical evolution within the canisters. This study advances the development of a modeling framework designed to simulate the chemical evolution of spent fuel canisters. Both thermal and radiation-driven reactions are considered, with radiation kinetics quantified using G-values. Sensitivity analysis identifies key parameters influencing species composition. Reaction pathway diagrams offer insight into dominant species formation routes, enabling more effective comparisons between model predictions and experimental observations, particularly regarding the production of hydrogen. Results show that the model predicts significant hydrogen gas production with minimal oxygen generation, primarily due to hydrogen formation via boehmite pathways. These findings underscore the importance of accurately characterizing surface-bound species and radiolysis kinetics. A deeper understanding of these mechanisms is critical for evaluating the long-term safety of nuclear waste storage.

12 - MGMT OF RADIOACTIVE AND NON-RADIOACTIVE WASTE↗