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At least 271 records · Page 15
All sourcesNASA NTRSDOE OSTI
DOE OSTI2017

Review of the TREAT Conversion Conceptual Design and Fuel Qualification Plan

The U.S. Department of Energy (DOE) is preparing to re establish the capability to conduct transient testing of nuclear fuels at the Idaho National Laboratory (INL) Transient Reactor Test (TREAT) facility. The original TREAT core went critical in February 1959 and operated for more than 6,000 reactor startups before plant operations were suspended in 1994. DOE is now planning to restart the reactor using the plant's original high-enriched uranium (HEU) fuel. At the same time, the National Nuclear Security Administration (NNSA) Office of Material Management and Minimization Reactor Conversion Program is supporting analyses and fuel fabrication studies that will allow for reactor conversion to low-enriched uranium (LEU) fuel (i.e., fuel with less than 20% by weight 235U content) after plant restart. The TREAT Conversion Program's objectives are to perform the design work necessary to generate an LEU replacement core, to restore the capability to fabricate TREAT fuel element assemblies, and to implement the physical and operational changes required to convert the TREAT facility to use LEU fuel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Irradiation Experiment Design Parameters for NBSR Fuel Conversion

It has been proposed to convert the National Institute of Standards and Technology research reactor, known as the NBSR, from high-enriched uranium (HEU) fuel to low-enriched uranium (LEU) fuel. The motivation to convert the NBSR to LEU fuel is to reduce the risk of proliferation of special nuclear material. This report is a compilation of relevant information from recent studies related to the proposed conversion using a metal alloy of LEU with 10 w/o molybdenum. The objective is to inform the design of irradiation experiments that are being planned. This report provides relevant dimensions of the fuel elements, and average and maximum fission rate densities and fission densities. The latter profiles are given for different plates to show the variation throughout the core. A summary of the methodology to obtain these results is presented. Fuel element tolerance assumptions and hot channel factors used in the safety analysis are also given.

42 ENGINEERING↗
DOE OSTI2019

Fuel Plate Surface Temperature in a Fully-Uncovered Fuel Assembly after a Loss-of-Coolant Accident in an MTR-Type Research Reactor

A methodology and model that were developed in the 1960s for calculating the peak fuel temperature in an MTR fuel assembly in a reactor core that has been fully-uncovered in a hypothetical loss-of-coolant accident (LOCA) are revived and updated in this report. The semi-empirical one-node model is based on experiments on fuel elements for the Oak Ridge Research Reactor (ORR) and the Low Intensity Testing Reactor (LITR) that were done in the 1950s and 1960s at the Oak Ridge National Laboratory. The peak plate temperature in 18 LOCA tests in the ORR and two of 24 tests in the LITR, calculated using the model, are compared with experiments. The original model was used in LOCA analyses for the Omega West Reactor (OWR) at Los Alamos National Laboratory with HEU fuel and for three cores of the University of Virginia Reactor (UVAR) with both HEU and LEU fuels. The LOCA analyses for UVAR using this model were accepted for licensing purposes by the US Nuclear Regulatory Commission for an assembly containing 12 HEU fuel plates in 1970, an assembly containing 18 HEU fuel plates in 1984, and an assembly containing 22 LEU silicide fuel plates in 1994. The analyses done for the OWR and UVAR are repeated here and compared with data in the original reports in order to establish the validity of the model revived in this report. The model is then used in a subsequent LOCA analysis for the HEU and LEU cores of the RPI MTR-type reactor in Portugal. It can also be applied to other MTR-type reactors with designs similar to the LITR.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

Design Optimization of A generic Fissile Solution for Mo99 production using Electron Beam-based Neutron Generator using MCNP+CFD

Molybdenum-99 ( 99 Mo) is a medical isotope used in 80% of all medical imaging procedures today. However, 99 Mo is not directly administered to patients for imaging; its decay product, technetium-99m ( 99 mTc), is a metastable isotope with a half-life of ~6 hours. Such a short half-life does not allow for production, separation, and application to a patient before decaying further. Therefore, the solution is to produce the precursor to 99 mTc, 99 Mo. 99 Mo has a half-life of 66 hours, long enough to produce the material and send it to the desired location before it fully decays. Under National Nuclear Security Administration’s (NNSA) highly enriched uranium (HEU) minimization mission, the Material Management and Minimization (M3) office leads the Molybdenum-99 (Mo-99) program. A major goal of the Mo-99 program is to develop methods of producing 99 Mo using low enriched uranium (LEU) opposed to the traditionally used HEU domestically in the U.S. Internationally, LEU is the new standard for use in research reactors and isotope production facility. In this study, we introduce a generic fissile solution system for 99 Mo production facility meeting the design requirement of 1) LEU usage as fissile material and 2) subcritical fissile system coupled with a horizontal electron beam (E-beam) accelerator. The focus of this research is to optimize a fissile solution configuration for maximum 99 Mo production yield using MCNP model, and ultimately implement the optimized design with computational fluid dynamics (CFD) model to understand thermal hydraulic behavior and solution convection characteristic. The report consists of two parts: 1) isotope yield calculation and criticality analysis with a wide range of parametric test matrix, and 2) thermal hydraulic analysis for solution mixing and cooling assessment.

60 APPLIED LIFE SCIENCES↗
DOE OSTI2021

Nuclear Data and Cross Section Testing Using ENDF/B-VIII.0

With the release of the Evaluated Nuclear Data File (ENDF)/B-VIII.0 library, nuclear criticality safety practitioners and engineers have access to the latest cross section sets available for their analyses. However, these cross sections must be rigorously tested and validated to ensure that the nuclear data are responsive to the needs of the individuals responsible for developing, implementing, and maintaining computational tools for criticality safety applications. Thus, the ENDF/B-VIII.0 library is tested and validated with a large collection of experiments that were vetted by the International Criticality Safety Benchmark Evaluation Project and made available in the International Handbook of Evaluated Criticality Safety Benchmark Experiments. A selection of benchmark experiments for use within the criticality safety community were prepared and reviewed within the Verified, Archived Library of Inputs and Data (VALID), which is maintained by the Nuclear Energy and Fuel Cycle Division at Oak Ridge National Laboratory. The performance of the ENDF/B-VIII.0 library is assessed by using VALID models of benchmark experiments with the beta 12 version of SCALE 6.3 KENO V.a and KENO-VI Monte Carlo codes. The performance is compared with the results obtained from with the ENDF/B-VII.1 library. This report considers multigroup (MG) and continuous energy (CE) formats of the ENDF/B-VIII.0 and -VII.1 libraries. The benchmark experiments within VALID that validate the ENDF/B-VIII.0 library cover 15 broad system categories by using a range of fissile materials, uranium enrichments, plutonium isotopic vectors, and mixed uranium/plutonium systems. These forms are represented as metals, solutions, or various arrays of rods or plates that cover a variety of neutron energy spectra: thermal, fast, mixed, and intermediate. Over 600 cases were considered for use with the KENO V.a and KENO-VI codes with the ENDF/B-VIII.0 library. The results of the Monte Carlo comparison of ENDF/B-VIII.0 to ENDF/B-VII.1 with both KENO V.a and KENO-VI indicate that there is a less than 0.53% Δk difference between the bias of calculated k eff from the expected values. The CE ENDF/B-VIII.0 library results in smaller magnitude biases than the ENDF/B-VII.1 data for HEU-MET-FAST, HEU-SOL-THERM, IEU-MET-FAST, LEU-SOL-THERM, PU-SOL-THERM, and U233-MET-FAST systems, while the MG results yielded smaller magnitude biases for HEU-MET-FAST, HEU-SOL-THERM, IEU-MET-FAST, LEU-COMP-THERM, LEU-SOL THERM, MIX-COMP-FAST, and U233-MET-FAST systems. Most notable are the adjustments to the plutonium and 233 U cross section data, which has resulted in noticeably lower biases in the ENDF/B VIII.0 results for the mixed, plutonium, and 233 U systems. Results of the sensitivity data file comparison generated from TSUNAMI-3D for selected VALID cases for the ENDF/B-VIII.0 library indicate a very high level of agreement with correlation coefficients of the effect of nuclear data uncertainty on k eff (the c k integral parameter) all above 0.99. This indicates that cases with the ENDF/B-VIII.0 library would see very similar responses to any nuclear data errors or change as those with the ENDF/B-VII.1 library.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Irradiation Thermo-Mechanical Modeling and Analysis of University of Missouri Research Reactor HEU Fuel Plates

The University of Missouri Research Reactor (MURR) located in Columbia, Missouri is one of six U.S. High Performance Research Reactors (USHPRR), including one critical facility, that is actively collaborating with U. S. Department of Energy (DOE) National Nuclear Security Administration (NNSA) Material Management and Minimization (M3) Office of Reactor Conversion and Uranium Supply to convert from the use of highly enriched uranium (HEU; $\geqslant$ 20 wt% U-235) to low-enriched uranium (LEU; <20 wt% U-235) fuel. A new type of very high-density LEU fuel based on an alloy of uranium and 10 wt% molybdenum (U-10Mo) is expected to allow the conversion to LEU of MURR, as well as four other USHPRR.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

HFIR High Power HEU Neutronics Analyses

Department of Energy National Nuclear Security Administration Office of Material Management and Minimization’s mission includes the conversion of civilian research reactors from highly enriched uranium (HEU) to low-enriched uranium (LEU) fuel. Analyses have shown that the Oak Ridge National Laboratory High Flux Isotope Reactor (HFIR) will need to operate at 95 MW for the LEU silicide dispersion fuel designs to match key performance metrics obtained with HEU fuel at 85 MW. To prove safe operation of HFIR after installation of plant modifications to increase power, a high power HEU test cycle was proposed. Neutronics model updates and reactor physics analyses are performed to support the development of safety design reports for the high power (HP) HEU test cycle. Reactor physics metrics evaluated herein include fuel depletion, actinide production, cycle length, fission rate density distributions, axial power peaking factors, and reactor kinetic parameters. These reactor physics analyses support the development of future LEU safety design reports by providing key input for future HP HEU HFIR thermal hydraulics and reactor transient safety analyses.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
NASA NTRS1971

Mechanism of pulmonary conversion of angiotensin I to angiotensin II in the dog.

The conversion mechanism was studied in vivo in the pulmonary circulation of the intact anesthetized dog and in vitro in plasma by using L-Leu-angiotensin I, D-Leu-angiotensin I, and des-Leu-angiotensin I which had been synthesized by the solid-phase technique. The results obtained indicate that pulmonary conversion in vivo and plasma conversion in vitro occur via a dipeptidylcarboxypeptidase and that a D-amino acid at the C-terminus prevents conversion.

Oparil, S.↗
NASA NTRS2018

Space Technology Mission Directorate: Game Changing Development Program: Nuclear Thermal Propulsion

Nuclear Thermal Propulsion (NTP) can help enable detailed exploration of the solar system, extensive development and utilization of cis-lunar space, and robust human Mars architectures. To improve affordability and viability, NTP systems that utilize low-enriched uranium (LEU) instead of highly enriched uranium (HEU) are being devised. Advanced fuel manufacturing techniques will help enable the use of LEU in certain fission systems previously thought to require HEU. The current LEU NTP baseline engine relies on a fission reactor containing fuel elements made of uranium nitride and refractory metals.

testing and performance↗
DOE OSTI2015

Steady-State Thermal-Hydraulics Analyses for the Conversion of BR2 to Low Enriched Uranium Fuel

The code PLTEMP/ANL version 4.2 was used to perform the steady-state thermal-hydraulic analyses of the BR2 research reactor for conversion from Highly-Enriched to Low Enriched Uranium fuel (HEU and LEU, respectively). Calculations were performed to evaluate different fuel assemblies with respect to the onset of nucleate boiling (ONB), flow instability (FI), critical heat flux (CHF) and fuel temperature at beginning of cycle conditions. The fuel assemblies were characteristic of fresh fuel (0% burnup), highest heat flux (16% burnup), highest power (32% burnup) and highest burnup (46% burnup). Results show that the high heat flux fuel element is limiting for ONB, FI, and CHF, for both HEU and LEU fuel, but that the high power fuel element produces similar margin in a few cases. The maximum fuel temperature similarly occurs in both the high heat flux and high power fuel assemblies for both HEU and LEU fuel. A sensitivity study was also performed to evaluate the variation in fuel temperature due to uncertainties in the thermal conductivity degradation associated with burnup.

Licht, J.↗
DOE OSTI2015

Prospective Activities outlined for Regulatory Approval in Ghana Overview

The Ghana Research Reactor-1 (GHARR-1) is one of Chinese’s Miniature Neutron Source Reactor (MNSR) which was purchased under a tripartite agreement between Ghana, China and the IAEA. The reactor was installed in 1994 and has since been in operation without any incident. It has been used chiefly for Neutron Activation Analysis (NAA) and Training of students in the field of Nuclear Engineering. The GHARR-1 has been earmarked for the Conversion of Core from HEU to LEU which is in accordance with the GTRI program and other related and/or associated programs. Over the past few years the National Nuclear Research Institute (NNRI), the Operating Organization of the Research Reactor for the Ghana Atomic Energy Commission (GAEC), has undertaken various tasks in order to implement the replacement of the reactor core. After completion, of the neutronic calculations, results showed that that an LEU fuel of 12.5% enrichment was desirable. However, recent developments have shown that an LEU fuel with 13% enrichment will be fabricated by the manufacturers, which is captured in a fuel specification document sent to NNRI by the CIAE. It is therefore imperative that all neutronic and thermal hydraulic calculation be done again to help acquire regulatory approval. Furthermore, the radiation exposure to personnel involved in the conversion must be estimated to help convince our regulators. This paper outlines the processes and activities that will enable us meet regulatory requirements.

Abrefah, R. G.↗
DOE OSTI2018

Neutronic and Thermal-Hydraulic Feasibility Studies for High Flux Isotope Reactor Conversion to Low-Enriched Uranium U3Si2-Al Fuel

Oak Ridge National Laboratory (ORNL) operates the High Flux Isotope Reactor (HFIR) for the US Department of Energy Office of Science to perform neutron scattering experiments, produce isotopes, and conduct materials irradiation research. ORNL is funded by the National Nuclear Security Administration (NNSA) Office of Materials Management and Minimization to evaluate the conversion of HFIR’s fuel from high-enriched uranium (HEU) to low-enriched (LEU). Due to potential fabrication issues with the complex UMo fuel design required to convert HFIR, NNSA requested ORNL to evaluate an alternate LEU fuel form – U3Si2-Al dispersion fuel. Neutronic and thermal-hydraulic feasibility analyses were performed with Shift and HSSHTC, respectively, to predict reactor performance and thermal safety margins. A number of designs were proposed and evaluated using an iterative approach in an effort to show that reactor performance could match that obtained using HEU fuel and that thermal safety margins were adequate. This study concludes that conversion of HFIR with U3Si2-Al LEU (19.75 wt.%) fuel with 4.8 gU/cm3 is feasible if, among other requirements, the fuel meat region is centered and symmetric about the fuel plate thickness centerline, the active fuel zone length is increased 5.08 cm to 55.88 cm, the proposed fabrication tolerances can be met, and the fuel can be qualified for HFIR conditions.

Chandler, David↗
DOE OSTI2019

U.S. HIGH PERFORMANCE RESEARCH REACTOR PRELIMINARY DESIGN MILESTONE FOR CONVERSION TO LOW ENRICHED URANIUM FUEL

The U.S. reactors that regularly refuel with highly enriched uranium (HEU) have undertaken conversion to low-enriched uranium (LEU) using a high-density alloy of uranium-10 wt% molybdenum (U-10Mo) that is under development for the U.S. High Performance Research Reactors (USHPRR). The U.S. National Nuclear Security Administration (NNSA) Material Management and Minimization (M3) Reactor Conversion Program is pursuing USHPRR conversion. The USHPRR consists of nine national laboratories and three universities working with a uranium processing plant and commercial fuel fabricator to convert five reactors and an associated critical assembly. Feasibility designs were completed in 2009 for all USHPRR. Since then, the USHPRR have completed key milestones leading to preliminary designs. Between 2007 and 2011, the Fuel Qualification (FQ) Pillar conducted seven irradiation campaigns consisting of 14 large size plates and over 60 mini plates with U-10Mo fuel with a zirconium interlayer. The Preliminary Report on U-Mo Monolithic Fuel for Research Reactors submitted for review to the U.S. Nuclear Regulatory Commission (NRC) in 2017 summarizes these irradiation tests. The Fuel Fabrication (FF) Pillar has installed and is operating a fabrication pilot line at BWX Technologies, Inc. In tandem, the Reactor Conversion (RC) Pillar, has been working with each USHPRR to mature conversion designs and to include recent data. As of 2018, all five USHPRR have completed LEU designs. This includes three Preliminary Safety Analysis Reports (PSARs) submitted between 2014 and 2017 to the NRC by the National Institute of Standards and Technology, University of Missouri, and Massachusetts Institute of Technology for their reactors NBSR, MURR, and MITR, respectively. In 2018, the two USHPRR reactors regulated by the U.S. Department of Energy (DOE) at Idaho National Laboratory (ATR) and Oak Ridge National Laboratory (HFIR) completed optimized conceptual or preliminary designs, with ATR confirming a ‘base fuel’ design of the same fabrication complexity as required for MURR, MITR and NBSR. Instead of available U-10Mo designs, HFIR is pursuing a longer silicide (U3Si2) fuel design with a fuel meat density of 4.8 gU/cm3. However, HFIR conversion will require additional fuel qualification due to the complex fuel design containing boron and since HFIR exceeds existing silicide qualification limits. These preliminary designs allow for a commercial fabrication demonstration, and subsequent irradiation qualification of these fuel element designs in order to complete final SARs that allow for USHPRR conversion to LEU fuel.

Wilson, E. H.↗
DOE OSTI2017

Spark Plasma Sintering (SPS) Update for the Fuel Conversion Effort at the Transient Reactor Test Facility

Over the lifetime of the Transient Reactor Test Facility (TREAT) reactor from 1959 to 1994, a series of historical tests disclosed certain material specific requisites for irradiation testing variables resulted in negligible structural damage after exposure to the reactor and multiple experiments. Comparable fuel design, material testing, and qualifications needs to occur for the high-enriched uranium (HEU) to low enriched uranium (LEU) fuel conversion (<20% U235) of TREAT. This includes, retaining a uniform precipitate dispersion of fueled UO2 micron-size particles throughout a graphite-moderating matrix, where high-density graphite is in direct contact with the fuel and acts as a moderator. On that note, the design of future LEU reactor TREAT fuel core will consider the relative sizing and spacing of fuel as well as the retention of graphite in accordance with the expected thermal, neutron, and energy portfolio for future conversion. In light of the above, LEU conversion fuel blocks currently being manufactured must undergo critical materials testing, irradiation, and examination to determine effects on the manufactured replacement fuel blocks at the millimeter to sub-micron size scale. Herein, we will report and update the community on value-added studies focusing on spark plasma sintering as an alternative technique to fabricate high density graphite dispersion fuel.

36 MATERIALS SCIENCE↗
DOE OSTI2022

ANS Winter 2022: ATF-2C Physics Safety and Scoping Analysis

ATF-2C comprises four tiers with six rodlets each. The experiment test train (TT) is situated in Loop 2A in the center flux trap (CFT) of the ATR, in which typical pressurized water reactor (PWR) conditions shall be approximated. Tier 1 (bottom) of the TT tests Framatome silicon carbide (SiC) cladding, containing molybdenum heater rods. Low-enriched uranium (LEU) Framatome rodlets with chrome-coated M5 cladding have been modeled in their place for the physics safety analysis to allow the substitution of this tier with fuel if needed. Tier 2 tests LEU Framatome rodlets with M5 cladding, which had irradiated in ATR in prior ATF experiments. Tier 3 tests General Atomics SiC composite cladding, also containing molybdenum heater rods. Tier 4/5/6 (top) tests the Japan Atomic Energy Agency (JAEA)/Mitsubishi Heavy Industries (MHI) fueled rodlets. This upper tier tests LEU rodlets with chrome-coated MDA cladding—several with temperature or pressure instrumentation. To flatten the axial power profile, lower portion of Tier 4/5/6 is surrounded by a hafnium (Hf) shroud, and the bottom two UO2 pellets are of natural enrichment. The ATF-2C TT is to be irradiated in the CFT of ATR (Fig. 1) for three or four 60-day cycles.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Feasibility Studies on a Hexagonal-Lattice Core for a World-Class Cold Neutron Source

Cold neutrons have kinetic energies less than 5 meV and wavelengths greater than 4 Α. They can be transported over tens of meters through super-reflecting neutron guides with minimal losses. Cold neutrons are used to investigate the structure of larger molecular systems such as biological materials, polymers and aggregates, due to their preferential interaction with light elements and general ignorance towards heavy ones. Intense beams of cold neutrons are obtained from slowing down a source of thermal energy neutrons - which may be from either a research reactor or spallation neutron source - through some cryogenic moderator such as liquid hydrogen or deuterium. The present reactor at the NIST Center for Neutron Research (NCNR) - NBSR - is such a cold neutron source (CNS) facility which provides cold neutron beams for over 2,000 scientists performing various experiments annually. Feasibility studies on a replacement reactor for the NBSR are underway at NCNR with the primary purpose of optimized cold neutron production. Low enriched uranium (LEU) fuel is required for the new reactor to conform to nuclear non-proliferation agreements. The low thermal power rate of 20 MW is taken as a priority with a fuel cycle length designated as 30 days. A horizontal split compact core with a large D{sub 2}O reflector tank is currently proposed and studied, with the expectation of achieving much better cold neutron performance than the present NIST reactor (NBSR). A preliminary core design has been completed using MCNP6 modeling and simulation. The physics performance characteristics of the proposed new core indicate that the new design is competitive with the most notable existing advanced cold neutron sources. In previous studies, only rectangular shaped fuel elements were considered because they have similar external dimensions as the NBSR fuel elements. However, from a compact core standpoint, a square lattice is not the most effective geometry for a tight configuration. Moreover, the previous studies also showed that the square lattice of the fuel elements left little room to accommodate vertical insertion blade-type control elements. As a result, all control blades had to reside in the D{sub 2}O reflector, which made the design cumbersome (in that case the control blades must penetrate the reflector tank) and neutronically uneconomical. In this summary, a preliminary feasibility study on a hexagonal core lattice is performed to render an alternative geometry for the new reactor (HEX core) to provide sufficient space for the control elements. The identical LEU fuel (U{sub 3}Si{sub 2}/Al) with U-235 enrichment 19.75% is used in the new design. The maximum thermal power and fuel management scheme remain the same as the previous investigations (three 30-day cycled batches for each batch of three fuel elements). The horizontal split compact core concept is also applied in the hexagonal lattice core design, with the objective of creating maximum thermal flux traps between the split core halves. A detailed description of the model for the hexagonal lattice core is presented in the following section. The HEX core design achieves a number of successes over the previous split-core design. Given that much of the motivation of this work was based on implementation of control blades into a concept which lacked them, it is notable that neutron economy was well-preserved in the process of achieving conservatively effective reactivity control; start-up k{sub eff} was in fact increased by 0.23% and peak thermal flux increased by ∼25%. A useful metric in determining the efficacy of resources towards creating an environment for the production of cold neutrons - dubbed the 'quality factor' (QF) - is the ratio of peak thermal neutron flux in a system to the total thermal power produced. Because HEX core's objectives are very focused in comparison to existing research reactors, it is fitting to pursue a design which out-performs all others in terms of some specific performance metric, namely this QF term. While HFIR remains the greatest cold neutron source in the world, it does so through use of HEU fuel (which will be replaced with LEU in the near future), and suffers a greater level of fast neutron contamination that is expected from the HEX design. The new core design achieves these successes due to the hexagonal lattice fuel arrangement, a 13.5% reduction in cladding mass, and a 13.1% reduction in fuel element volume. In spite of a 5.5% reduction in fuel mass, no compromises in reactor performance were noted. Provided the favorable characteristics shown of the control blade design, refueling intervals longer than the prescribed 30 days may be considered. It can be speculated that with these quantitative benefits, the overall capital and continuous costs of the HEX core design will be less than the previously proposed split-core concept.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2019

Support of UJV Rez, a. s. for Shipments of Spent Nuclear Fuel Within the M3 Program - 19035

In 2005, UJV Rez, a. s. (UJV, the former Nuclear Research Institute Rez), Czech Republic joined the Russian Research Reactor Fuel Return (RRRFR) program under the US-Russian Global Threat Reduction Initiative (GTRI) (now Material Management and Minimization program, M{sup 3}). The primary goal of the program is to advance nuclear nonproliferation objectives encouraging eligible countries to convert their research reactors from highly enriched uranium (HEU) to low enriched uranium (LEU) fuel and eliminating stockpiles of HEU. UJV's daughter company, Research Centre Rez, operates the LVR-15 research reactor with output of 10 MWth, which has been in operation since 1957. After more than 50 years of operation of the reactor, a large amount of spent nuclear fuel (SNF) of Russian origin had been accumulated. The SKODA VPVR/M high capacity casks were used for SNF shipment from the LVR-15 research reactor back to the Russian Federation (RF). The cask has a capacity of 36 fuel assemblies, and 16 casks are available. This means that 576 FA can be transported in one shipment. Two SNF shipments of HEU and LEU SNF from UJV were realized in 2007 and 2013. After the shipments were completed, only LEU nuclear fuel remained on the territory of the Czech Republic. Proven to be an excellent and very sophisticated system for SNF transportation, the SKODA VPVR/M casks were further employed for SNF shipments from Bulgaria, Hungary, Poland, Ukraine, Belarus, Serbia, Vietnam, Uzbekistan and Georgia to Russia. Some shipments were carried out as air shipments with use of the TUK-145/C cask system developed in Russia that included the SKODA cask. In 2016, the new SKODA MNSR casks were developed on the base of the SKODA VPVR/M casks in the Czech Republic and currently are being employed in the shipments of HEU cores of Chinese Miniature Neutron Source Reactors (MNSR) to China. The TUK-145/C-MNSR cask system developed in Russia is used for air transportation. The shipment from Ghana was carried out in 2016; the shipment from Nigeria is under realization. Shipments from Syria, Iraq and Pakistan are planned. Sixteen shipments from eleven countries using a total of 109 SKODA casks have already been completed without any incident or accident and more than 3500 fuel assemblies or cores have been shipped to Russia and China. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Bond Strength of Plasma Sprayed Zr Diffusion Barrier on U-Mo Fuels

The United States government is continually committed to strengthening nuclear security and nonproliferation for reducing the threat of radioactive materials acquisition by terrorists. To meet this important mission, the US Department of Energy and National Nuclear Security Agency's (NNSA) Office of Material Management and Minimization (M{sup 3}) Reactor Conversion Program (CONVERT) seeks to minimize the use of highly enriched uranium (HEU) in civilian applications through work with governments and facilities around the world to convert research reactors to the use of non-weapons-usable low enriched uranium (LEU) fuels. Each reactor converted not only eliminates the need for HEU at civilian sites, but also reduces the amount of HEU being manufactured, stored, and transported, where the nuclear material is most vulnerable. In instances where suitable LEU fuels do not exist for the conversion of particular reactors, CONVERT contributes to development of new LEU fuels. Currently, the CONVERT Program is committed to development of a high-density U-Mo fuel that will be able to convert high performance research reactors, while establishing an efficient and economic fabrication capability to manufacture the fuel. To that end, Los Alamos National Laboratory (LANL) has been tasked with development of a method to affix zirconium (Zr) metal fuel diffusion barrier via plasma spraying, which will prevent a reaction between the U-Mo fuel and the 6061 aluminum cladding. A successful diffusion barrier requires adequate Zr layer thickness, uniform coverage, sufficient bonding at the Zr/fuel and Zr/aluminum cladding interface, and maintenance of the original nuclear fuel properties i.e. physical dimensions, alloy composition and dispersion, crystallographic phase concentration, neutronics, etc. This work outlines development of the Zr plasma spray processing techniques, parameters affecting the diffusion barrier/nuclear fuel bond strength, and characterization of the Zr diffusion barrier. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗