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DOE OSTI2024

Actinides and Neutron Scattering at the Spallation Neutron Source and High Flux Isotope Reactor

The importance of actinides for science, technology, and medicine is widely recognized and does not require a lengthy introduction. However, the fundamental understanding of actinides and their chemical and physical properties lags behind the knowledge of most other elements in the periodic table. Most actinides (89 ≤ Z ≤ 103) are man-made and were produced in particle accelerators or nuclear reactors in the second half of the 20th century. Although trends in bonding and electronic structure are well characterized and documented across the periodic table, fundamental questions remain regarding the chemistry and physics of the actinides. This lack of knowledge is largely because most isotopes of actinides are intensely radioactive, which greatly increases the difficultly in handling them.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2026

Combined Nitrogen and Phosphorous Recovery via Electrochemical Technology Integration into Municipal Wastewater Treatment Plants

This project developed electroN-P, an electrochemical technology for recovering nitrogen and phosphorus as a fertilizer product from anaerobic digester centrate, targeting an advancement from TRL 4 to 6. The reactor was scaled 220× from a 250 mL batch cell to a continuous 4-channel system treating 55 L of wastewater, recovering >80% of phosphorus. Energy consumption under constant-voltage operation was lower than the embedded energy of conventional fertilizers, while constant-current operation produced a cost-competitive product with a smaller reactor footprint. TEA and LCA outcomes were highly sensitive to the magnesium source; alternative magnesium salt configurations projected reduced costs and embedded energy compared to sacrificial rods, but at the expense of significantly longer operation times. The recovered fertilizer performed comparably to diammonium phosphate and triple superphosphate in soil and plant trials. Integrating the technology into a whole-plant model reduced aeration energy by nearly 50%, lowering the levelized cost of water treatment from $0.668/m 3 to $0.648/m 3 . Results support targeting commercialization at smaller (≤1 MGD) facilities with high-strength digester streams. Further investigation into the transport and corrosion kinetics governing sacrificial anode wear is recommended to support continuous, longer-term operation.

32 ENERGY CONSERVATION, CONSUMPTION, AND UTILIZATI↗

DOE OSTI2021

Membrane-Confined Iron Oxychloride Nanocatalysts for Highly Efficient Heterogeneous Fenton Water Treatment

Heterogeneous advanced oxidation processes (AOPs) allow for the destruction of aqueous organic pollutants via oxidation by hydroxyl radicals ( • OH). However, practical treatment scenarios suffer from the low availability of short-lived • OH in aqueous bulk, due to both mass transfer limitations and quenching by water constituents, such as natural organic matter (NOM). In this work, we overcome these challenges by loading iron oxychloride catalysts within the pores of a ceramic ultrafiltration membrane, resulting in an internal heterogeneous Fenton reaction that can degrade organics in complex water matrices with pH up to 6.2. With • OH confined inside the nanopores (~ 20 nm), this membrane reactor completely removed various organic pollutants with water fluxes of up to 100 L m –2 h –1 (equivalent to a retention time of 10 s). This membrane, with a pore size that excludes NOM (>300 kDa), selectively exposed smaller organics to •OH within the pores under confinement and showed excellent resiliency to representative water matrices (simulated surface water and sand filtration effluent samples). Moreover, the membrane exhibited sustained AOPs (>24 h) and could be regenerated for multiple cycles. Our results suggest the feasibility of exploiting ultrafiltration membrane-based AOP platforms for organic pollutant degradation in complex water scenarios.

36 MATERIALS SCIENCE↗

DOE OSTI2022

To dee or not to dee: costs and benefits of altering the triangularity of a steady-state DEMO-like reactor

Shaping a tokamak plasma to have a negative triangularity may allow operation in an edge-localized mode-free L-mode regime and with a larger strike-point radius, ameliorating divertor power-handling requirements. However, the shaping has a potential drawback in the form of a lower no-wall ideal beta limit, found using the MHD codes CHEASE and DCON. Using the new fusion systems code FAROES, we construct a steady-state DEMO2 reactor model. This model is essentially zero-dimensional and neglects variations in physical mechanisms like turbulence, confinement, and radiative power limits, which could have a substantial impact on the conclusions deduced herein. Keeping its shape otherwise constant, we alter the triangularity and compute the effects on the levelized cost of energy (LCOE). If the tokamak is limited to a fixed B field, then unless other means to increase performance (such as reduced turbulence, improved current drive efficiency or higher density operation) can be leveraged, a negative-triangularity reactor is strongly disfavored in the model due to lower β N limits at negative triangularity, which leads to tripling of the LCOE. However, if the reactor is constrained by divertor heat fluxes and not by magnet engineering, then a negative-triangularity reactor with higher B 0 could be favorable: we find a class of solutions at negative triangularity with lower peak heat flux and lower LCOE than those of the equivalent positive triangularity reactors.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2015

Anaerobic digestion of paunch in a CSTR for renewable energy production and nutrient mineralization

Highlights: • Anaerobic digestion and nutrient mineralization of paunch in a CSTR. • Low CH{sub 4} yield and high CH{sub 4} productivity was obtained at an OLR of 2.8 g VS L{sup −1} day{sup −1.} • Post-digestion of the digestate resulted in a CH{sub 4} yield of 0.067 L g{sup −1} VS. • Post-digestion is recommended for further digestate stabilization. - Abstract: A laboratory study investigated the anaerobic digestion of paunch in a continuous stirred tank reactor (CSTR) for the recovery of biogas and mineralization of nutrients. At an organic loading rate (OLR) of 2.8 g VS L{sup −1} day{sup −1} with a 30-day hydraulic retention time (HRT), a CH{sub 4} yield of 0.213 L g{sup −1} VS and CH{sub 4} production rate of 0.600 L L{sup −1} day{sup −1} were obtained. Post-anaerobic digestion of the effluent from the CSTR for 30 days at 40 °C recovered 0.067 L g{sup −1} VS as CH{sub 4}, which was 21% of the batch CH{sub 4} potential. Post-digestion of the effluent from the digestate obtained at this OLR is needed to meet the stable effluent criteria. Furthermore, low levels of soluble ions such as K{sup +}, Ca{sup 2+} and Mg{sup 2+} were found in the liquid fraction of the digestate and the remainder could have been retained in the solid digestate fraction. This study demonstrates the potential of biogas production from paunch in providing renewable energy. In addition, recovery of plant nutrients in the digestate is important for a sustainable agricultural system.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

DOE OSTI2013

Reactor physics verification of the MCNP6 unstructured mesh capability

The Monte Carlo software package MCNP6 has the ability to transport particles on unstructured meshes generated from the Computed-Aided Engineering software Abaqus. Verification is performed using benchmarks with features relevant to reactor physics - Big Ten and the C5G7 computational benchmark. Various meshing strategies are tested and results are compared to reference solutions. Computational performance results are also given. The conclusions show MCNP6 is capable of producing accurate calculations for reactor physics geometries and the computational requirements for small lattice benchmarks are reasonable on modern computing platforms. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2013

Neutron production using a pyroelectric driven target coupled with a gated field ionization source

A palm sized, portable neutron source would be useful for widespread implementation of detection systems for shielded, special nuclear material. We present progress towards the development of the components for an ultracompact neutron generator using a pulsed, meso-scale field ionization source, a deuterated (or tritiated) titanium target driven by a negative high voltage lithium tantalate crystal. Neutron production from integrated tests using an ion source with a single, biased tungsten tip and a 3 Multiplication-Sign 1 cm, vacuum insulated crystal with a plastic deuterated target are presented. Component testing of the ion source with a single tip produces up to 3 nA of current. Dielectric insulation of the lithium tantalate crystals appears to reduce flashover, which should improve the robustness. The field emission losses from a 3 cm diameter crystal with a plastic target and 6 cm diameter crystal with a metal target are compared.

43 PARTICLE ACCELERATORS↗

DOE OSTI2016

Optimized design of reactor cores for continuous operation of Belgian nuclear power plants

Two Belgian nuclear power plants, owned and operated by Electrabel, were initially planned to be phased out in 2015, reaching their maximum life time of operation by the initial design. The fuel management of those two units has been subsequently designed by Tractebel to make the best use of the energy in the fuel assemblies either in the core or in the storage pool and to minimize the size of the fresh reloads in the terminating cycles. However, for ensuring energy security in Belgium, the Government decided in the same year to extend the life time of operation of the two units for 10 years, which raised the need to resume their continuous operation, in the opposite direction of the core contents and with a significant break in the usual sequence of cycle reloads. All this required adaptation of the tools and process for optimized reactor core design, in order to ensure the continuous and flexible operation of those units. The Tractebel LWRWIMS-PANTHER (LWP) core physics package is based on the 3-dimensional (3D) neutronic code PANTHER and the lattice code LWRWIMS for the generation of the assembly nuclear data. Both have been qualified by Tractebel for the Belgian Units to perform a complete range of PWR reactor calculations for fuel management design safety parameters assessment, transient analysis and operational support. In-core fuel management involves calculating core reactivity, power distribution and isotopic inventory for subsequent fuel cycles of a nuclear power plant in order to maintain adequate safety margins and operating lifetime for each core loading. This paper presents Tractebel's improved core design tools and process, the approach to optimize the reactor core design, and the applications to the 2 successive but opposite design requirements.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2015

Delayed neutron fraction and prompt decay constant measurement in the MINERVE reactor using the PSI instrumentation

The critical decay constant (B/A), delayed neutron fraction (B) and generation time (A) of the Minerve reactor were measured by the Paul Scherrer Institut (PSI) and the Commissariat a l'Energie Atomique (CEA) in September 2014 using the Feynman-alpha and Power Spectral Density neutron noise measurement techniques. Three slightly subcritical configuration were measured using two 1-g {sup 235}U fission chambers. This paper reports on the results obtained by PSI in the near critical configuration (-2g). The most reliable and precise results were obtained with the Cross-Power Spectral Density technique: B = 708.4±9.2 pcm, B/A = 79.0±0.6 s{sup -1} and A 89.7±1.4 micros. Predictions of the same kinetic parameters were obtained with MCNP5-v1.6 and the JEFF-3.1 and ENDF/B-VII.1 nuclear data libraries. On average the predictions for B and B/A overestimate the experimental results by 5% and 11%, respectively. The discrepancy is suspected to come from either a corruption of the data or from the inadequacy of the point kinetic equations to interpret the measurements in the Minerve driven system. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2021

Thermonuclear neutron emission from a sheared-flow stabilized Z-pinch

We report the fusion Z-pinch experiment (FuZE) is a sheared-flow stabilized Z-pinch designed to study the effects of flow stabilization on deuterium plasmas with densities and temperatures high enough to drive nuclear fusion reactions. Results from FuZE show high pinch currents and neutron emission durations thousands of times longer than instability growth times. While these results are consistent with thermonuclear neutron emission, energetically resolved neutron measurements are a stronger constraint on the origin of the fusion production. This stems from the strong anisotropy in energy created in beam-target fusion, compared to the relatively isotropic emission in thermonuclear fusion. In dense Z-pinch plasmas, a potential and undesirable cause of beam-target fusion reactions is the presence of fast-growing, “sausage” instabilities. This work introduces a new method for characterizing beam instabilities by recording individual neutron interactions in plastic scintillator detectors positioned at two different angles around the device chamber. Histograms of the pulse-integral spectra from the two locations are compared using detailed Monte Carlo simulations. These models infer the deuteron beam energy based on differences in the measured neutron spectra at the two angles, thereby discriminating beam-target from thermonuclear production. An analysis of neutron emission profiles from FuZE precludes the presence of deuteron beams with energies greater than 4.65 keV with a statistical uncertainty of 4.15 keV and a systematic uncertainty of 0.53 keV. This analysis demonstrates that axial, beam-target fusion reactions are not the dominant source of neutron emission from FuZE. These data are promising for scaling FuZE up to fusion reactor conditions.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2022

A High Temperature Gas Reactor Integrated Energy System Using Energy Storage as a Buffer for Hydrogen Production

Analysis of an integrated energy system producing hydrogen in a high temperature steam electrolysis system via steam provided by a sensible heat thermal energy storage system charged by a high temperature gas reactor. Preliminary results over a continuous discharge cycle and single charging cycle shows minimal reactor impact despite 25% electrical power load change. Hydrogen production is held constant with steam provided by the thermal energy storage system.

08 HYDROGEN↗

DOE OSTI2017

Core Conversion of VVR-K Research Reactors

The conversion of the VVR-K research reactor to low-enrichment uranium fuel includes switching to a compact core with low-enrichment uranium and gradually transitioning from a light-water side reflector to a beryllium reflector with restoration of the working reactivity margin without makeup with fresh fuel. In 2015, the reactor was stopped in order to modernize the main elements of the core, the control and protection system for the channels of the working rods, and the emergency electric power source. First criticality of VVR-K with low-enrichment fuel followed by first start of the reactor occurred in spring 2016. The VVR-KN type FA specially developed for VVR-K make it possible to not only preserve but also improve the neutronphysical characteristics of the core. The design-basis reactivity margin of the working core load ~7%Δk/k is adequate for reactor operation for three 20-day cycles with ~10% average burnup reached in the central FA. The beryllium reflector is built up in stages starting at the fourth cycle. The beryllium reflector will make it possible to preserve the radial dimensions of the core and increase the neutron flux density in the central and peripheral irradiation channels.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2018

MP1 Intermediate Characterization Report Summary

Great effort worldwide has been invested in replacing high enriched uranium (HEU) fuel with low enriched uranium (LEU) fuel in nuclear plants. This effort aims in safeguarding nuclear material and promoting the proliferation resistance of current reactors. The United States Department of Energy (DOE), National Nuclear Security Administration's (NNSA) Materials Management and Minimization Reactor Conversion program is actively working to convert civilian research and test reactors from the use of HEU to LEU fuel. To date, DOE NNSA has converted or verified the shutdown of various domestic reactors and also reactors worldwide. The task to develop and test a new LEU fuel while maintaining current performances has been assigned to the High-Performance Research Reactor (USHPRR) program. U alloyed with Mo achieves the goal of high uranium density necessary for the conversion of the high performance research reactors, and has also demonstrated good performance in pile [1]. The alloying process of U with Mo stabilizes the body-centered cubic (bcc) gamma phase (?-U) at room temperature. Without this alloying property enhancement, this phase is only stable at temperatures only above 770 °C. Retaining this isotropic gamma phase within the U-Mo alloy provides swelling and oxidation resistance and desired mechanical properties [2]. Generally 10% w. has been considered sufficient to achieve stabilization. However, local phenomena within the alloy can affect the local Mo content and lead to gamma phase decomposition. Because of this, it has been suggested that controlling gamma phase decomposition is most critical during processing [2-3]. The work detailed in this document is part of the HPPR program and focuses on the characterization of fresh fuel (a.k.a. as fabricated, before reactor testing) in collaboration with other laboratories including PNNL (Pacific Northwest National Laboratory), LANL (Los Alamos National Laboratory) and BWXT (Babcock and Wilcox Technologies). This testing is part of a 4 way validation for which similar samples from master plates were taken and distributed to each laboratory to independently analyze utilizing a standardized set of measurement and analysis procedures. When complete this work will supply a “toolbox” of information for alloying and fabrication parameters with its direct effects on microstructure quantified. Various parameters have been analyzed in this characterization effort (including grain size, quality of the coating layer, composition of the samples, carbide content etc.), based on the direction of the working group in ref.[4]. The final aim of this work is to increase the understanding of the impact of processing conditions on the final fuel microstructure, using fabrication processes that meet “commercial viability” requirements [4]. The strength of this campaign rests on the extent of the samples studied and the variation of the manufacture parameters which will permit to achieve the required understanding. Thus permitting to use these experiments to down-select test samples used to evaluate irradiation performance [4] to be tested in various locations of the Idaho National Laboratory (INL) Advanced Test Reactor (ATR).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2017

Experimental and Computational Investigation of the Emission of Radioactive Products from Irradiated Fuel of Fast Sodium-Cooled Reactor

The key experimental and computational studies of the emission of long-lived radioactive fission products and actinides from uranium dioxide irradiated in BN-600 into sodium and the gas phase in the range 150–800°C, performed on the experimental bench at the Leipunskii Institute of Physics and Power Engineering, are presented. The results of measurements of the properties of aerosols of certain radioactive fission products formed upon heating of irradiated nuclear fuel in sodium are presented. The experimental results are of interest for evaluating the radiation impact of reactor facilities on workers and the public during beyond design basis accidents.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI1988

Nuclear Safety [Vol. 29, No. 3, July-September 1988]

Nuclear Safety is a review journal that covers significant developments in the field of nuclear safety. Its scope includes the analysis and control of hazards associated with nuclear energy, operations involving fissionable materials, and the products of nuclear fission and their effects on the environment. Primary emphasis is on safety in reactor design, construction, and operation; however, the safety aspects of the entire fuel cycle, including fuel fabrication, spent-fuel processing, nuclear waste disposal, handling of radioisotopes, and environmental effects of these operations, are also treated. Table of Contents for this issue follows. GENERAL SAFETY CONSIDERATIONS: 259 Fifteenth Water Reactor Safety Information Meeting by E. G. Silver; CONTROL AND INSTRUMENTATION: 284 Reliability Technology to Improve and/or Maintain Emergency Diesel Generator Performance by S. Karimian and J. H. Taylor, 293 A Noise Diagnostics System for Operator Advice by G. Hessel, P. Liewers, P. Schumann, W. Schmitt, and F.-P. Weiss; PLANT SAFETY FEATURES: 307 A Passive Containment System for Advanced Light-Water Reactors by O. B. Falls, Jr., and F. W. Kleimola; ENVIRONMENTAL EFFECTS: 318 Data Base Construction for a Computerized Radiological Risk Investigation System by L. M. Hively, J. E. Nyquist, J. L. Bledsoe, and A. L. Sjoreen, 326 Erratum to "Radiation Hormesis and Nuclear Safety," Vol. 29, No. 1; OPERATING EXPERIENCES: 327 Operational Safety Experience and Passive Safety Testing at the Fast Flux Text Facility by Q. L. Baird, J. L. Rathbun, D. D. Stepnewski, R. L. Stover, and A. E. Waltar, 344 Backfilling of Independent Residual Heat Removal Systems in West Germany and Switzerland by G. Eckert and Y. Salomon, 353 Reactor Shutdown Experience Compiled by J. W. Cletcher, 356 Selected Safety-Related Events Compiled by G. A. Murphy, 363 Operating U.S. Power Reactors Compiled by E. G. Silver; RECENT DEVELOPMENTS: 384 General Administrative Activities Compiled by E. G. Silver, 390 Reports, Standards, and Safety Guides by D. S. Queener, 395 Status of Power-Reactor Projects Undergoing Licensing Review Compiled by E. G. Silver, 400 Proposed Rule Changes as of Mar. 31,1988; ANNOUNCEMENTS: 283 International ENS/ANS Conference on Thermal Reactor Safety "NUCSAFE 88", 317 Northwestern University Short Course on Radiation Safety, 407 Second International Seminar on Small- and Medium-Sized Nuclear Reactors, 407 International Workshop on New Developments in Occupational Dose Control and ALARA Implementation at Nuclear Power Plants and Similar Facilities, 408 Fourth International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-4), 404 The Authors.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2022

Experiment Assembly and Testing in Hot Cell in Support of Advanced Fuels Irradiation Testing

Follow-on experimentation of irradiated nuclear fuel is critical to understanding fuel behavior at different stages of the fuel lifecycle. Transient experiments using fuel with high burnup can enlarge understanding of fuel fragmentation, relocation and dispersal under various reactor accident situations. Assembling experiments with such fuel requires remote operations.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI1988

Nuclear Safety [Vol. 29, No. 3, July-September 1988]

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2015

Joint tests at INL and CEA of a transient hot wire needle probe for in-pile thermal conductivity measurement

Thermal conductivity is a key property that must be known for proper design, testing, and deployment of new fuels and structural materials in nuclear reactors. Thermal conductivity is highly dependent on the physical structure, chemical composition, and the state of the material. Typically, thermal conductivity changes that occur during irradiation are currently measured out-of-pile using a 'cook and look' approach. But repeatedly removing samples from a test reactor to make measurements is expensive, has the potential to disturb phenomena of interest, and only provides understanding of the sample's end state when each measurement is made. There are also limited thermo-physical property data available for advanced fuels; and such data are needed for simulation codes, the development of next generation reactors, and advanced fuels for existing nuclear plants. Being able to quickly characterize fuel thermal conductivity during irradiation can improve the fidelity of data, reduce costs of post-irradiation examinations, increase understanding of how fuels behave under irradiation, and confirm or improve existing thermal conductivity measurement techniques. This paper discusses efforts to develop and evaluate an innovative in-pile thermal conductivity sensor based on the transient hot wire thermal conductivity method (THWM), using a single needle probe (NP) containing a line heat source and thermocouple embedded in the fuel. The sensor that has been designed and manufactured by the Idaho National Laboratory (INL) includes a unique combination of materials, geometry, and fabrication techniques that make the hot wire method suitable for in-pile applications. In particular, efforts were made to minimize the influence of the sensor and maximize fuel hot-wire heating. The probe has a thermocouple-like construction with high temperature resistant materials that remain ductile while resisting transmutation and materials interactions. THWM-NP prototypes were fabricated for both room temperature proof-of-concept evaluations and high temperature testing. Evaluations have been performed jointly by the INL and the French Alternative Energies and Atomic Energy Commission (CEA), both in Idaho Falls (USA) and in Cadarache (France), in the framework of a collaborative program for instrumentation of Material Testing Reactors. Initial tests were conducted on samples with a large range of thermal conductivities and temperatures ranging from 20 deg. C to 600 deg. C. Particularly, tests were recently performed on a sample having thermal conductivity and dimensions similar to UO{sub 2} and MOX nuclear fuels, in order to validate the ability of this sensor to operate for in-pile characterization of Light Water Reactors fuels. The results of the tests already completed at INL and CEA indicate that the Transient Hot Wire Needle Probe offers an enhanced method for in-pile detection of thermal conductivity. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗