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DOE OSTI2020

Gamma ray Imaging Based Methods to Improve the Accuracy of Uranium Holdup Quantification

Deposits in commercial uranium processing facilities of interest to nuclear safeguards are predominantly low enriched uranium (LEU). The shapes and sizes of holdup deposits can vary significantly, thus making it difficult to model the deposits accurately for calibration purposes. This presents a challenge to currently employed methods such as the Generalized Geometry Holdup (GGH) that rely on quantifying 235U mass by simplifying deposit shapes as a point, a line, or an area. In this work, gamma ray imaging using high energy resolution Germanium Gamma Imagers (GeGI™) is employed to determine the distribution of uranium inside the source containment. Two types of imaging methods are employed; coded-aperture imaging and Compton imaging. Gamma-ray emissions seen in different regions of the image are quantified using an inverse gamma-ray transport solver being developed at ORNL. Using measured data, the inverse method will be used to solve for unknown source parameters such as source matrix thickness, density, and attenuation due to container wall and shielding. The intrinsic efficiency of the detector is determined based on a library of built-in response functions and key dimensions of the detector (e.g., thickness and radius of the detector crystal). Measurements are performed in both imaging modalities using uranium sources of well-known masses and enrichments configured inside mocked-up holdup fixtures. Results from the two imaging modalities are intercompared. The paper presents preliminary results of the imaging measurements and the progress made in the computational methods to quantify uranium mass.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗

DOE OSTI2020

Fuel Conversion Efforts at the High Flux Isotope Reactor – a 2020 Status Update

The High Flux Isotope Reactor (HFIR) provides one of the world’s highest steady-state neutron fluxes in the world for neutron scattering experiments focused on impactful scientific discovery, as well as materials irradiation studies and production of medical, industrial, and research isotopes. Efforts are ongoing to convert HFIR from high-enriched uranium (HEU) to low-enriched uranium (LEU) fuel while maintaining or enhancing current performance and safety margin, thus sustaining HFIR’s mission portfolio and reactor-based neutron science leadership. This paper presents a status update on the HFIR fuel conversion efforts.

Chandler, David↗

DOE OSTI2021

Evaluating Safeguards Statistical Assumptions via Stochastic Simulation

Herein, the authors built and tested a stochastic simulation to estimate achieved detection probabilities (DPs) on a stratum basis, over a tailorable range of diverted amounts from 0 to 2 SQ, using typical IAEA inspection data: i.e., SQ in stratum, number of items, number of gross/partial/bias defect measurements conducted, and realistic relative standard deviation (RSD) values for typical IAEA verification measurements. For bulk strata, the model calculates achieved DP at 0.01 SQ diversion increments; for item strata, the model calculates DP using the smallest realistic diversion increment (e.g., a plate, pin, or coupon). After successfully benchmarking against IAEA deterministic models, the simulation was used to test the sensitivity of DP to certain standard assumptions and selected input parameters. First, the equal defect assumption was tested; the results suggest significant complexity in the effectiveness of partial defect measurements. Next, the authors explored the sensitivity of DP to the assumed RSD of attribute tests. Then, the authors compared non-normal models for instrument performance (e.g., logistic, step, or arbitrary functions) to the typical results from a normal distribution (characterized by RSD). This last comparison was supplemented with experimentally derived performance data for an HM-5. The HM-5 was used to make enrichment measurements on both LEU and HEU MTR fuel elements as plates were removed, and the results fit with logistic and step curves and applied in the simulation. These stochastic DP results were compared to DP estimates from a deterministic model assuming a normal curve and typical RSD, yielding insights that could improve effectiveness in the field. These early results illustrate the potential of stochastic models to better understand achieved DP and to improve safeguards effectiveness.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗

DOE OSTI2021

Modelling Nuclear Thermal Propulsion Reactor Startup Transients

The National Aeronautics and Space Administration (NASA) has set the goal of a manned mission to Mars by the year 2030 [1] and charged the national academy of sci- ences "to identify primary technical and programmatic chal- lenges, merits, and risks for maturing space nuclear propulsion technologies of interest to a future human Mars exploration mission" [2]. One relevant technology, nuclear thermal propul- sion (NTP), has notable advantages over traditional chemical rockets; most important among them is the ability to produce larger specific impulse on the order of 900s. The reduction of mission time is crucial for a manned mission to Mars to reduce the risk for the crew. Due to its higher specific impulse, NTP satisfies this need and is pursued as one technology to get humans to Mars [3, 4]. The construction of an NTP sys- tem has to negotiate several challenges laid out in Ref. [2]; one of these challenges is the need to startup the NTP sys- tem from essentially cold conditions to full power within one minute. This paper focuses on studying the neutronics and thermal-hydraulics behavior of a simplified NTP model dur- ing prescribed reactivity insertions and mass flow rate (MFR) ramps. It is the goal of this paper to investigate startup, peak material temperatures, and average specific impulse for a low enriched Uranium (LEU), ceramic and metal material (CER- MET) NTP system when varying reactivity insertion and MFR ramps.

33 ADVANCED PROPULSION SYSTEMS↗

DOE OSTI2022

Engineering RuBisCo for food safety

A method of improving a ribulose-1,5-bisphosphate carboxylase/oxygenase (RuBisCo) to have a higher protein score is disclosed. The method includes the steps of: making a modified RbcL of the RuBisCo, by, on an RbcL unit of the RuBisCo, either substituting Met for Leu, Phe, Val, or Ile or combinations thereof; substituting Lys for Arg, Thr, or His or combinations thereof; or both of these substitutions. The modified RbcL consequently modifies the RuBisCo and is added to a biomass host where it is stable for homologous recombination. Plastid and nucleus integration was observed. Example RbcL sequences are disclosed with the desirable substitutions. The improved RuBisCo can be used as an improved proteinaceous food source for humans and animals.

Davis, Ryan Wesley↗

DOE OSTI2023

DOE Advanced Gas Reactor Fuel Development and Qualification Program Overview

Slides discussing the DOE Advanced Gas Reactor Fuel Development and Qualification Program Overview to include focus on LEU UCO TRISO fuel in cylindrical compacts, with data for fuel qualification in support of reactor licensing and a path to establish a domestic commercial TRISO fuel fabrication capability. Timelines from 2004 through 2029. Major program highlights include: Developed fuel performance models based on past observed behavior, Study to understand differences between US and German (“good”) fuel, Developed and improved fuel fabrication and characterization methods, Fabricated lab-scale fuel (AGR-1) with very high quality; excellent in-pile and post-irradiation high-temperature fuel performance, Scale-up coating fabrication (AGR-2) with excellent in-pile performance, AGR-3/4 fuel fabrication and irradiation experiment to assess FP transport behavior, Scale-up matrix fabrication (AGR-5/6/7); & fuel performance evaluation in progress.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2023

Unraveling the Mechanics: Computational Modeling of Residual Stresses in U10Mo Fuel with Aluminum Cladding

LEU (Low Enriched Uranium) plate-type fuel elements consist of a high-density, low-enrichment U–Mo alloy-based fuel foil encapsulated in an aluminum cladding and is fabricated through the Hot Isostatic Pressing (HIP) technique. Understanding the mechanism of possible failure modes of nuclear fuel is critical to mitigate potential consequences. One of the major contributing factors in various failure modes is stress and it greatly affects the integrity under temperature, pressure, and irradiation. These stresses could originate from various sources, including manufacturing, and operation, and can lead to deformation, cracking, or even complete failure of fuel elements. This study focuses on the development of a computational model that accurately predicts the residual stresses generated in the U10Mo fuel during fabrication. It has been observed that cladding creep has a substantial impact on the residual stresses in the U10Mo fuel post the HIP fabrication process. Furthermore, during the HIP bonding process the fuel plate system is heated to a temperature of 560 oC and as a result, the aluminum cladding transitions from (Al 6061-T6 to -O). This presents a challenge in capturing the change in material properties accurately. Therefore, a temperature dependent creep model such as hyperbolic sine creep model is considered to estimate the creep properties of the Aluminum cladding. The proposed calibrated creep model accurately predicts the residual stresses in the U10Mo fuel foil and agrees well with the experiments.

42 ENGINEERING↗

DOE OSTI2023

Molybdenum Sleeves Experiments in the Sandia Critical Experiments Facility

Sandia National Laboratories and the Institut de Radioprotection et de Sûreté Nucléaire have collaborated on the design and execution of a set of critical experiments that explore the effects of molybdenum in water-moderated fuel-rod arrays. The molybdenum was included as sleeves on some of the fuel rods in the critical experiment fuel arrays. Approach-to-critical experiments were performed on five configurations of fuel and molybdenum sleeves using the 7uPCX fuel in core hardware that set the triangular fuel rod pitch at 15.494 mm. The experiments are evaluated as benchmark critical experiments for the 2023 edition of the International Criticality Safety Benchmark Evaluation Project (ICSBEP) Handbook as LEU-COMP-THERM-111.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2023

Fuel Conversion Efforts at the High Flux Isotope Reactor – a 2023 Status Update

Sizemore, Carol↗

DOE OSTI2015

Recent upgrades and new scientific infrastructure of MARIA research reactor, Otwock-Swierk, Poland

The MARIA reactor is open-pool type, water and beryllium moderated. It has two independent primary cooling systems: fuel and pool cooling system. Each fuel assembly is cooled down separately in pressurized channels with individual performances characterization. The fuel assemblies consist of five layers of bent plates or six concentric tubes. Currently it is one of the most powerful research reactors in Europe with operation availability at least up to 2030. Its nominal thermal power is 30 MW. It is characterized by high neutron flux density: up to 3x10{sup 14} n cm{sup -2} s{sup -1} in case of thermal neutrons, and up to 2x10{sup 13} n cm{sup -2} s{sup -1} in case of fast neutrons. The reactor is operated for ca. 4000 h per year. The reactor facility is equipped with fully equipped three hot cells with shielding up to 10{sup 15} Bq. Adjacent to the reactor facility, the radio-pharmaceutics plant (POLATOM) and Material Research Laboratory are located. They are equipped with a number of hot cells with instrumentation. The transport system of radioactive materials from reactor facility to Material Research Laboratory is available. During 2014 the MARIA reactor has been operated with three different types of fuel the same time: previous 36% enriched fuel, and two types of new LEU fuels. In the meantime, molybdenum irradiation programme has been developed. Maria is a multifunctional research tool, with a notable application in production of radioisotopes, radio-pharmaceutics manufacturing (ca. 600 TBq/y), {sup 99}Mo for medical scintigraphy (ca. 6000 TBq/y), neutron transmutation doping of silicon single crystals, wide scientific research based on neutron beams utilization. From the beginning MARIA reactor was intended for loop and fuel testing research activities. Currently it is used mostly as material testing and irradiation facility and for that reason it has wide experimental capabilities. There are eight horizontal irradiation channels from among whom six of them are equipped with instrumentation for condensed matter physics research: - H3 - spectrometer and diffractometer with double monochromator; - H4 - small angle scattering spectrometer; - H5 - polarized neutrons spectrometer; - H6, H7 - two 3-axial crystal neutron spectrometers; - H8 - neutron radiography stand. For two horizontal channels are ongoing exploitation programs: - H2 - station with epithermal neutron beam produced in uranium converter is being developed. Intelligent converter will be installed on the periphery of reactor core. The intensity of the beam will be at the level 2x10{sup 9} n cm{sup -2}s{sup -1} what makes the beam unique in the Europe. - H1 - special pneumatic horizontal mail is being developed for irradiation material samples in the vicinity of the core i.e. in the distal part of the H1 channel. The number of neutron irradiation facilities in MARIA reactor is increasing every year. Numerous of thermal neutron irradiation channels including fast hydraulic rabbit system and large size channels for fast neutron irradiation are used routinely. Recently new in-pile facility with ITER-like neutron energy spectrum for 14 MeV neutron irradiation has been constructed. Taking into account its performance and ability of almost incessant operation the facility appears as one of the most powerful 14 MeV neutron sources. The facility shall be used for material research connected with thermonuclear devices (ITER) and 4. generation nuclear reactors. The system of independent fuels channels used in MARIA reactor appear to be very flexible and very convenient to be used as irradiation channels for uranium targets for {sup 99}Mo production. Currently, MARIA reactor supplies ca. 18% world production of {sup 99}Mo. The MARIA reactor research activities are still extended. The current scientific projects are connected e.g. with silicon neutron transmutation doping, in-pile gamma heating measurements, French calculation codes implementation (TRIPOLI4, APOLLO2). The horizontal neutron beams utilization is also developed. The MARIA reactor, due to its primary application connected with loop and fuel testing, is very convenient for testing the nuclear instrumentation, control and measurement systems.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2015

Evaluation of a measurement system for Uranium electrodeposition control to radiopharmaceuticals production

For 2016, studies by international bodies forecast a crisis in the supply of Molybdenum ({sup 99}Mo), which is the generator of {sup 99m}Tc, widely used for medical diagnoses and treatments. As a result, many countries are making efforts to prevent this crisis. Brazil is developing the Brazilian Multipurpose Reactor (RMB) project, under the responsibility of the National Nuclear Energy Commission (CNEN). The RMB is a nuclear reactor for research and production of radioisotopes used in the production of radiopharmaceuticals and radioactive sources, broadly used in industrial and research areas in Brazil. Electrodeposition of uranium is a common practice to create samples for alpha spectrometry and this methodology may be an alternative way to produce targets of low enriched uranium (LEU) to fabricate radiopharmaceuticals, as {sup 99}Mo, used for cancer diagnosis. To study the electrodeposition, a solution of 10 mM uranyl nitrate, in 2-propanol, containing uranium enriched to 2.4% in {sup 235}U, with pH = 1, was prepared and measurements with an alpha spectrometer were performed. These studies are justified by the need to produce {sup 99}Mo since, despite using molybdenum in bulk, Brazil is totally dependent on its import. In this project, we intend to obtain a process that may be technologically feasible to control the radiation targets for {sup 99}Mo production. (authors)

07 ISOTOPE AND RADIATION SOURCES↗

DOE OSTI2016

Research reactor plate fuel multi-physics hotspot model for the HFIR

Ongoing development of a multi-physics hotspot model for nuclear research reactor plate fuel is presented with a focus on the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL). The background of plate fueled research reactors in general and the HFIR in particular are discussed along with the need to convert from high enriched uranium (HEU) to low enriched uranium (LEU) fuel, which drives the need to improve the analysis capability of fuel thermal-fluid performance. HFIR HEU fuel plate fabrication is described to explain the origin of fuel segregations and cladding non-bonds that currently contribute to the hotspot flux peaking factors. Previous efforts at modeling hotspots caused by these defects are reviewed and areas for improvement associated with high resolution multi-physics simulation are identified. Modes of heat transfer through non-bonds, including thermal contact conductance, are examined. Conjugate heat transfer over hotspots is also explored to high resolution up to and through the wall boundary. The challenges associated with multi-physics verification and validation (VV), particularly the requirements beyond VV 20, are presented. An approach to Validation within the limits of existing data for conduction heat transfer through the fuel plate, and convection into the cooling fluid is set forth. Preliminary simulations are presented assessing hot spot modeling approaches for contact conductance, and fluid turbulence modeling within COMSOL. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2016

Analysis of Synergistic Fuel Cycle Options with Thorium and Heavy Water Reactors

Fuel cycle options (FCOs) with thorium-based fuels have characteristics that may have benefits for meeting certain objectives, such as transuranic actinide inventory management and overall natural resource utilization. Neutron irradiation of thorium and uranium-233 results in negligible production of transuranic actinides. This advantage is compromised, to a degree, when thorium fuels are used in conjunction with uranium-238; however, strategic reprocessing and fuel use (e.g., recycle of plutonium and/or minor actinides) can be used to manage this problem. Furthermore, thorium FCOs with reprocessing can extend the availability of natural nuclear resources by breeding new fissile material, recovering and recycling unused material, and achieving high burnups. In instances where natural thorium is more readily available than uranium, FCOs which use thorium may be preferable from sustainability and energy security perspectives. However, many hybrid uranium-thorium FCOs struggle to perform well in this regard, since large amounts of external fissile material (and therefore natural uranium) are required to breed the required uranium-233. DOE's evaluation and screening effort reports that for every metric ton (MT) of U-233 bred in a light water reactor for reuse, about 1000 MT of natural uranium are required to support such breeding, much of which is used inefficiently due to enrichment. The combination of HWRs and thorium-based fuels has the potential to reduce this consumption of natural uranium resources. Because heavy water moderates neutrons without significant absorption, fissile material requirements are lower compared to those of LWRs. This has enabled conventional 'once-through' HWRs to operate with only slightly-enriched or even natural uranium. Also, HWRs can be combined with LWRs in multistage FCOs where used LWR fuel can be processed and the remaining fissile content is sufficient to support use as fuel in HWRs in spite of the presence of neutron poisons (e.g., uranium-236); this concept has been demonstrated at a research scale in Korea. Thus, additional energy is extracted from the same initial natural uranium feed and uranium enrichment effort. Another consideration with using thorium-based fuels in HWRs is the current limited operational experience of the reactors and associated fuel cycle facilities using these fuels. However, work continues internationally towards gaining further experience. For example, at equilibrium the third stage of India's planned Three-Stage Nuclear Power Program uses thorium/uranium-233 or thorium/plutonium fuels with a fissile supply from second-stage fast reactors. Further, China plans to experiment with low-enriched uranium (LEU)-Th CANFLEX fuel bundles in a once-through configuration in its HWRs and a long-term objective is to transition to thorium use in a closed HWR FCO. Vanderbilt University (VU) and Oak Ridge National Laboratory (ORNL) are characterizing multi-stage, thermal-reactor FCOs that incorporate thorium and examining their potential performance. This paper presents insights from the examination of FCOs which consider the synergy of HWRs and thorium. Two multi-stage FCOs which incorporate thorium-based fuels in HWRs have been examined for their potential, one which uses recycled PWR uranium oxide fuel in conjunction with thorium in an HWR, and another that uses recycled HWR fuel in an MSR which also includes thorium. Both of the systems studied offer resource utilization improvements as compared to a once-through LWR fueled with low-enrichment uranium and highlight the potential synergy of thorium and HWRs. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2016

Validation of KENO Thermal Moderator Doppler Broadening Method in SCALE 6.2 Beta5 Using Continuous-Energy B-VII.1 Library

KENO, the Monte Carlo criticality safety program within the SCALE code package, has until recently used continuous-energy (CE) cross sections corresponding to the set of evaluated nuclear data closest to the temperature of interest. This method, which is used for both 1-D cross sections and thermal scattering probability data, does not fully represent the phenomenon of Doppler broadening (DB) and can result in significant differences in eigenvalues when the material temperature lies between the evaluated data sets. A method for creating temperature-dependent nuclear data has been implemented in KENO for CE calculations. This work summarizes the results of incorporating this method into two critical experiments using low-enriched uranium (LEU) in light water moderated pin lattices. The LEUCOMP-THERM-046 (LCT-046) experiment is a benchmark documented in the International Handbook of Evaluated Criticality Safety Benchmark Experiments (ICSBEP Handbook); it consists of 22 critical configurations or cases. The KRITZ-2:1 experiment, a part of the International Reactor Physics Experiment Evaluation Project (IRPhEP), consists of two cases which span a significant range of moderator temperatures. The results of this work give a good indication of the value of the newly incorporated cross section processing. However, more diverse evaluations should be performed to provide a broader validation of this method. The results suggest that when dealing with systems similar to LCT-046, there is a significant improvement made when the moderator scattering data are properly broadened and sampled. The DB of the hydrogen (in water) 2-D cross sections resulted in a consistent C/E value across the temperature range, whereas the DB of the 1-D only cross sections did not exhibit this improvement. With respect to criticality safety, this consistency adds precision to the calculational method, which could potentially impact subcriticality limits. After comparing the KRITZ and LCT-046 results, the most plausible conclusion to make as to why the added thermal moderator treatment improved the results for LCT-046 and not for the KRITZ experiment is that the two systems were dissimilar with respect to their temperatures. That is, the relative temperature differences among the cases in each experiment are more significant for LCT-046 than for KRITZ. The cross section changes as a function of temperature are also smaller at higher temperatures. This difference resulted in a lesser impact of broadening in general for the KRITZ hot case.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2016

PGSFR Core Design and Performance Characteristics

Korea Atomic Energy Research Institute (KAERI) has developed a Prototype Generation-IV Sodium-cooled Fast Reactor (PGSFR) as a representative of Gen-IV concepts. The final PGSFR core will be composed of transuranic (TRU) fuels to transmute LWR spent fuels and self-recycled its fuels, in combination with a pyro-processing facility. However, the initial POSFR core consists of low enriched uranium (LEU) fuels in order to: 1) provide an irradiation test for TRU fuels and structural materials, 2) keep pace with the development of pyro-processing technology, and 3) obtain operational experience on the reactor systems and components. During the first phase of PGSFR development plan (2012-2015), the power level was decided to be 150 MWe from many trade-off studies using different power level. The PGSFR uranium core performance characteristics at the initial, transition and equilibrium states have been carried out. In this paper, the PGSFR core design and related performance characteristics are introduced. The PGSFR uranium core design was performed at the initial/transition/equilibrium core based on the candidate fuel management scheme. It was confirmed that this design can satisfy surely all design limit and requirement. The chapter 4 of PGSFR Preliminary Safety Information Document (PSID) published in December 2015 contains the detailed information and characteristics of PGSFR core. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2016

Advanced Test Reactor Mini-Plate Hydraulic Testing

Currently, flow testing is being performed at Oregon State University (OSU) on the hydromechanical Fuel Test Facility (HMFTF) and the Endurance Flow Loop (EFL) to aid in the qualification of low enriched plate-type Uranium fuels for use in high performance research reactors in the US, under the Global Threat Reduction Initiative (GTRI). Under this project, varying geometries of small capsules proposed for irradiation testing in the Advanced Test Reactor (ATR) are being hydraulically testing prior to being placed in the ATR. The premise of this study was to conduct a number of flow tests on these capsules to characterize the pressure loss through each as a function of flow rate and compare the results against predicted theoretical values. Two categories of flow tests were performed in the Endurance Flow Loop to supplement flow testing of proposed LEU plate-type fuel for conversion of High Performance Research Reactors: Capsule Comparison Tests and Orifice Characterization Tests. The purpose of these tests were to determine the pressure drop across each experiment as a function of flow rate and to compare the results against a simple analytical model of the pressure loss through a pipe, to support the safety basis for the in-pile experiments. The capsule comparison tests compared the pressure loss through three types of capsules: RERTR, MP-1, and EMPIRE. In this category, nine flow tests were performed to determine the effect of having one, two, and three of each type of capsule placed in series in a test section. The analytical model accurately predicted, for each case, an increase in hydraulic resistance for each capsule that was added, though the values were conservative. The orifice characterization tests compared the pressure loss through various configurations of small orifice spacers in conjunction with blank spacers and the MP-1 capsules. The analytical model predicted that adding blank spacers would slightly increase the pressure drop due to added frictional effects, but indicated no change in the pressure drop based on the orientation of the capsules. The measured results indicated that the addition of the blank spacers had no measurable effect on the pressure drop, while changing the orientation of the capsules did. In all 15 flow tests, the theoretical model provides a conservative estimate of the pressure drop through each configuration for an equivalent flow rate, over that of the measured results.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2016

First-principles study of inert gas incorporation and migration in zirconium nitride

Uranium-Molybdenum alloy based fuels are the most promising fuel for the future high performance research reactors using low enriched uranium (LEU). However, when used in dispersion with aluminum, significant amount of interaction products form at the interface between the UMo fuel particle and the surrounding Al during in-pile tests which leads to non-acceptable swelling. To reduce the swelling in U-Mo/Al dispersion fuel at high burnups, different diffusion barrier coatings, e.g., ZrN, were deposited on the U-Mo particles to avoid the interaction between fuel particles and the Al matrix. These diffusion barrier coatings are very effective in reducing the formation of interaction layer and therefore the fuel swelling. So far, ZrN is selected as the candidate coating material for the European Dispersion Fuel program due to the superior thermophysical properties of ZrN. The thermodynamic and kinetic stability of ZrN coating with respect to Al and U-Mo has been studied by the current authors using density functional theory calculations. So far there are very few studies of the diffusion behaviors of inert gases in ZrN. A recent publication on irradiated (Pu,Zr)N reveals that the majority of fission-induced Xe is retained in the material, while He is release to the fuel pin to a large extent. This raises the questions on the underlying differences of inert gas behavior in nitride fuel. Therefore, the purpose of this study is to understand the incorporation and migration behavior of inert gases He, Kr and Xe in ZrN using density functional theory (DFT) based first-principles calculations. Using density-functional theory calculations we provided a systematic study of the incorporation and migration of fission gases in ZrN. We investigated the formation energies of several defects (vacancies, interstitials, divacancies, Frenkel pairs, and Schottky defects) and the incorporation energies of inert gases in these defects. Results show that the formation energies of point defects are strongly affected by the chemical potentials of N and Zr species. The energetically most favorable incorporation site for fission gases is found to be the first nearest neighbor Schottky defect, in which the fission gas atom sits at somewhere between the Zr and N vacancies. The predicted solution energies of rare gases, i.e, He, Kr and Xe, in ZrN, is almost linearly dependent on the atom size of inert gases. We compared the solution energy Xe in ZrN with several other nuclear materials with the same B1 structure (MX), e.g., TiN, UN and UC. The solution energies of Xe in MX are closely related to the size of vacancy sites. The potential migration mechanisms of Xe in ZrN include interstitial and vacancy-assisted diffusion. The calculations show that vacancies essentially act as Xe trapping sites. The migration behavior of Xe in N and Zr divacancy and Schottky defects were compared with that through tetrahedral interstitials. It is found that the migration barrier is much lower with the interstitial mechanism than those with the vacancy mechanism. Therefore, at low temperatures Xe diffusion in ZrN is mostly through the interstitial mechanism or existing grain boundaries while at high temperature volume diffusion could be preferred, i.e., diffusion with Zr and N vacancies. (authors)

36 MATERIALS SCIENCE↗

DOE OSTI2016

Mechanical analysis to study pore formation mechanism in U-Mo/Al dispersion fuel

U-Mo/Al dispersion fuel is under development for the conversion of HEU-fuel using high-power research reactors to LEU fuel. One of challenges this fuel faces is large-size pore formation at the interfaces between the interaction layer (IL) and Al matrix. It is hypothesized that these pores are formed where the area is under a tensile stress that facilitates pore nucleation and growth. For a quantitative analysis with a simultaneous consideration of meat swelling and creep, ABAQUES finite element analysis (FEA) package was utilized to simulate five dispersion mini-plates with known creep rate constants for the U-Mo, IL and Al matrix. The FEA results are compared with measurement data for meat swelling and creep. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗