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At least 271 records · Page 15

Structural Characterization of Uranium Tetrafluoride Hydrate (UF 4 ·2.5H 2 O)

Uranium tetrafluoride is an important intermediate in the nuclear fuel cycle. Facile synthesis of its hydrate, uranium tetrafluoride hydrate (UF 4 ·2.5H 2 O), has recently been reported. The hydrate forms by contacting anhydrous UF 4 with neat H 2 O at room temperature for 24 h or by exposing anhydrous UF 4 to high relative humidity (>90%) conditions for several weeks. These pathways are of clear environmental relevance. Further understanding of the structure and optical spectra of UF 4 ·2.5H 2 O, especially of the water molecules, is therefore necessary. Herein, the structure of UF 4 ·2.5H 2 O was probed using time-of-flight neutron powder diffraction to improve understanding of the crystalline water environments in the structure. The complete structure was elucidated and compared to a previously reported partial structure for UF 4 ·2.5H 2 O and a predicted complete structure from density functional theory. The crystalline structure exhibits three distinct water environments: two of the three water sites are bound to uranium, and the third water is unbound or “free”. Furthermore, the completed structure reveals an extensive hydrogen bonding network involving water–fluorine and water–water interactions. One bound water site participates in hydrogen bonding with nearby fluoride ligands (O–H···F–U), and the second bound water site participates in hydrogen bonding with the unbound water (O–H···O) and a nearby fluoride ligand (O–H···F–U); the unbound water participates in hydrogen bonding with bound water (O–H···O–U). Low-temperature experiments and thermal analysis indicate UF 4 ·2.5H 2 O is thermally stable from 10 to 358 K, undergoes dehydration at higher temperatures, and is nearly dehydrated at 473 K. Structural measurements provide foundational understanding and will inform future investigations of the thermal and environmental stability of UF 4 ·2.5H 2 O.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

Multi-Probe 23A Pre-Shot report

The Multi-Probe Radiography campaign will field a full day of experiments at the Omega EP laser facility. The experiments will consist of shots that alternate between Omega EP’s two short-pulse laser beams to generate proton and deuteron beams from a variety of film and foam targets. The proton beam is intended to be used as a radiographic source, while the deuteron beam will be used as the pitcher in a pitcher-catcher neutron-generation concept. For the proton beam, our goal is to characterize the angular distribution and energy spectra of unwanted x-rays that are generated during proton acceleration and were found to be problematic for simultaneous radiography in MP-22A. As part of this primary goal, we will test whether an x-ray shield, XBLK, can be used to prevent these x-rays from impacting other x-ray diagnostics. For the deuteron beam, our primary goal is to characterize the energy spectra and beam profile of deuterons accelerated from CD foils and foams.

72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS↗

Optimization of the Post-Operational Phase on Two Belgian Multi-Unit Nuclear Power Plants: the Case of the Non-Fissile Irradiated Core Items - 20156

The current legal framework in Belgium foresees the progressive phase out of nuclear power between October 2022 (Doel 3) and December 2025 (Doel 2). Upon its definitive shutdown, each unit of the Tihange and Doel sites will enter a Post-Operational Phase (POP) and be prepared for its Decontamination and Decommissioning (D and D). Prior to obtaining the D and D license, the Operator Electrabel is legally required to remove any non-fissile irradiated core items stored in the deactivation pools. The non-fissile irradiated core items consist essentially of control rods, poison rods and source thimbles as well as thimble plugs and foreign materials irradiated during operation: - Their significant content in highly radiant radionuclides (up to 6 TBq of Co-60 per kg of irradiated material) renders all existing operational waste management processes inadequate due to insufficient biological shielding; - Their high concentrations in long-lived radionuclides call for their disposal in a geological repository for which no final design nor waste acceptance criteria are expected prior to 2050. Uncertainties in the Belgian energy supply and security, however, require the Operator to be prepared for a partial nuclear phase out, where one or more units would benefit from lifetime extension while the remaining units would undergo decommissioning. The present paper aims at presenting how Electrabel, in partnership with Tractebel, addressed this challenge by maximizing the use of synergies within the respective sites as well as between both sites themselves, all the while accounting for site specificities. The most recent results and state of progress of the project will be detailed and the first lessons learned will be shared. The project has been split in multiple tasks and phased as follows: - An inventory phase aimed at mapping the contents, origin, composition and history of the non-fissile irradiated core items; - A pre-characterization phase based on neutron activation models; - A waste sorting phase aimed at separating waste forms for which an evacuation route exists from those for which such route does not exist; - A feasibility phase aimed at exploring all possible scenarios for the management of non-fissile irradiated core items and identifying the optimal feasible solution for each site; - A preparation phase (currently ongoing), developing further the optimal solution and ensuring that back-up solutions are available for any foreseeable change of context (licensing issue, modification in the nuclear phase-out program, etc.) and initiating early contacts with potential subcontractors for segmentation works and cask manufacturers, as well as the Belgian regulatory body and waste management agency. This phase also foresees the investigation of destructive and non-destructive radiological measurements to support the detailed characterization of the waste forms; - A realization phase (future work). (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Multi-Probe Campaign FY24 Omega EP Shot Request [Slides]

The ultimate goal of the Multi-Probe experimental campaign is to assess the feasibility of using multiple probe species (x-ray, proton, neutron, and electron) in future LANL radiographic facilities. To use Omega EP most efficiently, two different experiments are proposed for each of the two requested days in FY24. In this way, shots may be interleaved on the two short pulse beams and the shot rate doubled. In FY24, we propose to continue to develop the neutron source by first optimizing the “pitcher” source of deuterons/protons and then adding the “catcher” neutron converter on day 2. The second series of experiments continues examining the origin and nature of the x-ray background found in our last two experiments. This background is not well understood or characterized by the community. The first day will measure the electrons emanating from the proton and x-ray targets while also optimizing the background shielding first used in FY23. The second day will test a geometry that enables simultaneous radiographs using both short-pulse beams.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

An influence of manufacturing tolerances on Pin-Cell k-infinity of MOX fuel using data from the FUBILA experiment program

An influence of manufacturing tolerances (MTs) was evaluated on the pin-cell k-infinity of MOX fuel through the random sampling of CASMO5 calculations. The data from the FUBILA experiment program was used as the manufacturing parameters. For the uncertainties of element/isotope mass fractions, their covariance matrices were calculated by the generalized least square method. The total k-infinity uncertainty was 120-250 pcm (percent mille). From the breakdown of k-infinity uncertainty per materials, the MTs of fuel pellet had a dominant influence. The individual influences were also evaluated for element/isotope mass fractions, an inner/outer diameter, and a density. Those of element/isotope mass fractions were less than a few dozen pcm. Since the perturbations of fuel pellet diameter and density caused large variations to the total amount of heavy metals and effect of spatial self-shielding, they had a negative correlation with the k-infinity perturbation. On the AG3 (Al-Mg alloy) over-cladding, the inner/outer diameter perturbations had both negative and positive correlations. The negative one was due to decrease in amount of light water and the positive one was due to increase in amount of AG3. Since the effect of neutron slowing-down by light water had a dominant influence, the influence of the former negative correlation was larger than that of the latter positive correlation. It is, therefore, concluded that the consideration of the MTs that had a large influence on the effect of neutron slowing-down is important for precise quantification of the pin-cell k-infinity uncertainty of MOX fuel. (author)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Analyses of Dose Rates for Second Inner Reflector Plug Replacement at Spallation Neutron Source

The Inner Reflector Plug (IRP) is a central component of the Spallation Neutron Source (SNS) target monolith, which houses the liquid mercury target and four liter-sized neutron moderator units. It is exposed to high-level radiation fields during routine operation and builds up significant activity. The IRP needs to be replaced due to moderator neutron poison and decoupler burn-out, which is used for shaping neutron pulses. The first IRP exchange took place in March 2018 and next one, which is IRP2, is planned to be replaced during the facility maintenance period in the end of December 2025, beginning of January 2026 calendar year. The replacement of IRP is a complex task due to its location in the area receiving high irradiation, being under significant amount of shielding, and size, which requires removal in segments. For planning the replacement workflow to help to reduce radiation exposure to workers dose rates for each stage of extraction/replacement operation are predicted.

42 ENGINEERING↗

Comparison of the Baseline USL Calculation Methods for Loosely-Coupled and Novel Neutronic Systems [Slides]

Current work includes reconstructing 187-group ENDF B/VII.1 covariance matrix for comparison study to 44-group ENDF B/VII.1 matrix to determine how covariance matrix structures affect the USL calculations and investigating the exact domination mechanism of region-wise sensitivities in a loosely-coupled system. Future work will involve seeking to understand how bias distributions change with reactivity and how to calculate propagate distribution error into USL calculations, as well as how cross section perturbation studies can be extended to other types of calculations, such as shielding calculations.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

ECAR-7448 Rev 0 MARVEL Post-Shutdown Dose Analysis in the Upper Confinement

This report documents a bounding post-shutdown dose rates over time in the upper confinement area and above the top of the TREXc Pit Shield at the end of two years of operation of the MARVEL microreactor. The dose rates take into consideration the doses from the core fission products, the sodium-potassium coolant (NaK), the steel components, the Stirling engines, and the gallium eutectic (EGaInSn) secondary coolant. The MCNP model in this analysis is based on the 90% design of the MARVEL reactor from ECAR- 7447. Neutron fluence used as inputs to this analysis were calculated in ECAR-7447.

21 - SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLAN↗

Description of FY25 Theory and Simulation Performance Target: Development of an integrated modeling framework for fusion reactor design and assessment

The urgency to deliver fusion power is growing now more than ever, with increasing pressure for both public programs and private companies to meet milestones timelines and overcome significant remaining technical challenges to ensure growth of a nascent fusion industry in time to meet rapidly growing clean energy demands. With incredible advancements in computation and years of investment in fusion model development and validation, integrated modeling is poised to fill a key role in accelerating the timeline to a fusion pilot plant (FPP). Future fusion pilot plants will operate in regimes far beyond current experience, and device design will rely on physics-based prediction and extrapolation. Many concepts will also rely on simulation to assess safety (shielding, tritium management, materials activation and lifetimes), economics and scalability before the decision to build. Importantly, integrated simulation can be used to reveal and solve the complexities of system integration that may otherwise not be apparent in physical components or models developed in isolation. New experimental test facilities that produce relevant conditions to validate and resolve key technical challenges for various subsystems (materials, blankets, fuel cycle, etc.) have been repeatedly called for by the fusion community but are not yet realized. Integrated modeling has an important role in identifying realistic load conditions (thermal, electromagnetic, plasma, neutron and photon loads, etc.) and defining the components and experiments for these test facilities in order to ensure meaningful validation that sufficiently reduces modeling uncertainties and technical risk for the full integrated reactor. The Fusion REactor Design and Assessment (FREDA) SciDAC project is building a component-based integrated modeling framework & data structure to enable self-consistent, multi-fidelity, iterative optimization workflows for the fusion reactor design process. FREDA aims to shorten the time to viable designs by providing a set of flexible workflows to support the various stages of the design process using an integrated model hierarchy, ranging from the simple analytic descriptions to the highest fidelity, theory-based plasma and engineering modeling developed by the fusion and fission communities. These tools are expected to be needed for timely support of FPP design in the milestone program and in the FIRE collaboratives. The plasma simulation backbone of FREDA is IPS-FASTRAN with newly developed coupled Core-Edge Pedestal-SOL (CESOL) workflows, which is being extended to the far-SOL region up to the plasma facing components. FREDA incorporates the FERMI engineering modeling suite and will enable self-consistent evaluation of the thermal shields, limiters, blanket, magnets, and other surrounding structures with predictions of temperatures, erosion, dpa, activation, tritium generation and transport, creep, corrosion, material degradation, etc. Parametric generation of 3D CAD enables rapid iteration of component geometry in response to plasma and loading specifications.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

VERA-Grizzly Ex-Core Calculations: Watts Bar Unit 1 Cycles 1-2

The critical structures that comprise light-water reactor (LWR) nuclear power plants are subjected to operating environments that can challenge their integrity. Structures in close proximity to the reactor core, such as the reactor pressure vessel (RPV) and the biological shield wall, are subjected to high levels of radiation emanating from the core, as well as elevated temperatures. As the US fleet of operating LWRs ages, the effects of these operating environments on the integrity of these structures must be considered to ensure their continued safe operation. Extending the lifetime of commercial reactors and maintaining the aging reactor fleet require accurate prediction of the exposure of ex-core components to neutron and photon radiation. In particular, concrete degradation studies must be performed to evaluate the safety and long-term operation of reactors with lifetime extensions. The concrete reactor bioshield is important for providing radiological protection during operation and must last for the entire lifetime of the reactor. Recent interest in lifetime extensions furthers the need to accurately simulate concrete material degradation in the reactor bioshield. As a result of this need, the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program has funded this study to couple its tools, Virtual Environment for Reactor Applications (VERA) and Grizzly. VERA allows users to set up models to calculate time-dependent and fully coupled solutions (with thermal feedback) for ex-core quantities of interest such as vessel and coupon fluence and detector responses for multiple statepoints and cycles. Grizzly is a finite-element application based on the Multiphysics Object Oriented Simulation Environment (MOOSE) framework that is used to enable aging materials calculations. This report highlights the work performed to calculate the fluence in the vessel and concrete for Watts Bar Nuclear Plant Unit 1 (WBN1) Cycles 1 and 2. The fluences obtained from VERA were successfully transferred to Grizzly using a Python script. Four simulations were run with Grizzly: (1) the Mazars model with the initial Young’s modulus being the instantaneous modulus, (2) the Mazars model with the initial Young’s modulus being the delayed modulus, (3) the Mazars model with the initial Young’s modulus being the delayed modulus with the addition of the effects of micro-damage caused by irradiation, and (4) the Mazars model with the initial Young’s modulus being the instantaneous modulus, and with the addition of micro-damage and creep. Details regarding the methods used to obtain the fluence and the statistical errors associated with the VERA Monte Carlo Shift calculations are discussed in greater detail in this report. The results obtained from the four Grizzly models are also presented in this report.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Advanced Test Reactor Safety Basis Update for Gas-Cooled Experiments

The Advanced Test Reactor (ATR) supports neutron irradiation of several types of experiments. One such experiment type is referred to as a gas leadout. Gas leadout experiments actively flow gas through the experiment which allows for active temperature control. It also allows for in-situ data of the experiment. For example, fission gas migration through a fuel sample can be monitored via activity of the sweep gas. Historically, ex-pile equipment and fission product monitors were housed in shielded ATR cubicles. Due to other facility updates, cubicle space is no longer available for gas leadout experiment equipment. To support continued operation of gas leadout experiments, ATR completed a safety basis update that supports a new housing for leadout equipment that may process potentially contaminated gas. In addition to the structure and associated equipment, technical safety requirements regarding handling and storage of experiments needed to be revised to support fueled gas leadout experiments and associated outage configurations. The safety basis update addressed the full lifecycle of these experiments, including experiment movement and interim storage, and credible abnormal events such as failures or leaks in contaminated gas tubing in occupied areas. This paper discusses the completed analyses performed to support the safety basis update associated with gas leadout experiments, including thermal-hydraulic evaluation, probabilistic analysis, and dose consequence analyses.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Validation of realistic Monte Carlo plasma gamma-ray source on JET discharges

A novel modelling methodology has been developed for the creation of a realistic plasma gamma-ray source for Monte Carlo transport simulations in the tokamak JET. The methodology couples the TRANSP code for plasma transport calculations with the MCNP Monte Carlo particle transport code, thus connecting plasma physics with gamma-ray transport. Here, we present the validation of the developed source methodology by comparing calculated gamma-ray spectra with measurements performed at JET. The validation focuses on gamma-ray spectra measured by the tangential gamma-ray spectrometer during two JET three ion RF scenario discharges, performed in the JET 2019 deuterium experimental campaign. For validation the calculated plasma gamma-ray spectrum was combined with the neutron induced prompt gamma-ray background, originating in the vacuum vessel, and scaled to absolute values calculating the total number of plasma gamma-ray and neutron emitting reactions. The comparison between calculated and measured gamma-ray spectra shows good agreement with the shape of the calculated gamma-ray spectra matching that of measurements for both studied discharges. Moreover, the calculated absolute values of the gamma-ray spectra were of the same order of magnitude at the position of the gamma-ray detector located at the end of a long line-of-sight in a biological shield. Finally, the comparison has validated the developed plasma gamma-ray source methodology for MCNP photon transport calculations at JET. The validation provides a basis for the developed plasma gamma-ray source to be used as a support for the development of future tokamaks such as DEMO.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

Production of Cf-252 and other transplutonium isotopes at Oak Ridge National Laboratory

In 1957 Glenn T. Seaborg conceived and advocated for the construction of the High Flux Isotope Reactor (HFIR) and the Transuranium Processing Plant (since then renamed the Radiochemical Engineering Development Center, or REDC) at Oak Ridge National Laboratory. There, heavily shielded hot cells, glove boxes, and laboratories allow recovery of transuranium elements produced in substantial quantities. Seaborg’s vision of HFIR and REDC producing milligram quantities of berkelium, californium, and einsteinium has been fulfilled beginning in 1966 through May 2019 with 78 production campaigns yielding a cumulative totals of 1.2 g of 249 Bk, 10.2 g of 252 Cf, 39 mg of 253 Es, and 15 pg of 257 Fm. Notably, 252 Cf is a neutron source used in many industrial applications including oil exploration; process control systems for the cement industry, coal analysis, and power production; sources to start nuclear reactors and perform nondestructive materials analyses; homeland security and national defense detection devices; and medical research. Isotopes made available through transplutonium production at HFIR/REDC have enabled scientists to study the nuclear properties and reactions, chemical properties, optical properties, and solid-state properties of transplutonium elements. Long-lived isotopes have served as targets in heavy ion accelerators to produce heavier elements leading to the discovery of 104 Rf, 105 Db, 106 Sg, 113 Nh, 114 Fl, 115 Mc, 116 Lv, 117 Ts, and 118 Og. This paper reviews the evolution of the processing flowsheets to produce, separate, and purify transplutonium isotopes, which have evolved over 50 years of operation at HFIR and REDC, and summarizes directions of future work to improve the efficiency of the production operations.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗

Reduced-order modeling for efficient cross section library development in high-temperature gas reactor pebble-bed depletion analysis

Accurate modeling of running-in and equilibrium conditions in pebble-bed reactors (PBRs) requires precise microscopic multigroup neutron cross sections. In Griffin, deterministic neutronics calculations rely on multivariate interpolation over large cross section libraries, resulting in significant memory usage and performance bottlenecks. This work, together with a companion paper on Griffin integration, explores reduced-order models (ROMs) to replace interpolation with lightweight surrogates. Several ROM techniques are benchmarked, with deep neural networks (DNNs) demonstrating superior memory efficiency, scalability, and predictive accuracy. A total of 295 DNNs were trained to build a comprehensive isotope library, integrated into Griffin through a custom LibTorch interface for depletion analysis. Initial results demonstrate that DNN-based ROMs drastically reduce memory demands while preserving accuracy, enabling finer tabulations and additional state variables without overhead. In conclusion, the framework also supports online cross section generation and real-time DNN updates through transfer learning, improving fidelity by capturing self-shielding and evolving nuclide compositions during burnup.

22 - GENERAL STUDIES OF NUCLEAR REACTORS↗

On the moisture migration of concrete subject to high temperature with different heating rates

A better understanding of moisture migration in concrete at high temperature can play an important role to improve the fire-resistance and radiation-shielding capability of concrete structures. In this regard, within the numerical framework of Multi-physics Lattice Discrete Particle Model, the moisture clog in concrete during heating is defined as the moisture-saturated region between the so-called “water front” and “water back” (isosurfaces of saturation = 1). The moisture migration in concrete subject to slow and fast heating has been investigated by simulating experimental tests in which moisture has been monitored via different techniques, namely 3D Neutron Tomography and Ground-Penetrating Radar. An overall consistency between experimental and numerical results has been observed, indirectly proving the effectiveness of numerical modelling and experimental monitoring of moisture migration.

36 MATERIALS SCIENCE↗

A new neutron time-of-flight detector for yield and ion-temperature measurements in DT implosions on OMEGA

A new neutron time-of-flight (nTOF) detector for yield and ion-temperature measurements in deuterium–tritium (DT) implosions was designed, installed, and calibrated on the OMEGA Laser System. This detector was installed at θ = 46.8° from the vertical and φ = 84.9°, previously a region without other detectors, in order to decrease the uncertainty of low-mode asymmetry measurements by six other nTOF detectors. The new nTOF detector uses the “PMT only” concept [Glebov et al., Rev. Sci. Instrum. 92, 013509 (2021)] and consists of a Photek photomultiplier tube (PMT)-110 in a thin aluminum housing. There is no lead shielding in front of the PMT to allow measurements of the x-ray instrument response function of the detector. The detector is installed in the OMEGA H2-port line of sight at 9.3 m from the target chamber center in the OMEGA Target Bay and is called the H2 nTOF detector. The H2 nTOF measures DT yields in the 1 × 10 13 to 3 × 10 14 range and ion temperatures from 2 to 12 keV. H2 nTOF reduced the hotspot flow uncertainty in the z-direction from ±10 to ±8 km/s, and the uncertainty in the anisotropic flow axis from σ θ = ±26° to ±13° and σ φ = ±25° to ±21°. Here, the design details, performance, and calibration results of the H2 nTOF detector in DT implosions on OMEGA will be presented.

Laser fusion↗

Phonon density of states in uranium Laves phases UNi 2 and UCo 2

Recent computational work evaluated the phase stability of UCo 2 and UNi 2 in the C14, C15, and C36 Laves phases and reported their computationally determined phonon density of states. To corroborate these computational findings, we report the experimentally determined phonon density of states for UCo 2 and UNi 2 using inelastic neutron scattering at the wide angular-range chopper spectrometer at the Spallation Neutron Source at Oak Ridge National Laboratory. We found excellent agreement between experimental and computational results for UNi 2 and noted that some differences were observed for UCo 2 . Furthermore, by comparing the phonon density of states of UCo 2 with that of other U-bearing Laves phases, we found that UCo 2 showed distinct vibrational properties. From the one-phonon density of states, we calculated the lattice contribution to the specific heat of UNi 2 and UCo 2 . Finally, we bound the electronic contribution to the specific heat in UNi 2 .

Actinides↗