Global Threat Reduction Initiative Research Articles

The involvement of ÚJV Řež and other Czech companies in shipments of spent nuclear fuel from nuclear research reactors

ÚJV Řež is extensively involved in a program concerning the shipment of spent nuclear fuel from nuclear research reactors to the respective countries of origin. One of the objectives of the Global Threat Reduction Initiative (GTRI) is to reduce the global threat potentially posed by the use of highly-enriched uranium (HEU) fuel in nuclear research reactors. The program is underway in cooperation with the USA, the International Atomic Energy Agency, the Russian Federation and China, and is funded primarily by the USA. A significant factor concerns the use of the Czech ŠKODA VPVR/M cask, which was developed and produced by ŠKODA JS. As part of the Russian Research Reactor Fuel Return (RRRFR) program, spent nuclear fuel from the LVR-15 nuclear research reactor at Řež was shipped from Czechia to the Russian Federation in 2007 and 2013. Non-irradiated nuclear fuel and other nuclear materials were also shipped from Czechia to the Russian Federation as part of the RRRFR program.ÚJV Řež is also involved in the shipment of spent nuclear fuel from other countries. ÚJV Řež has participated in seventeen shipments from twelve countries involving a total of 3,328 kg of uranium, of which 713 kg of HEU. This highlights Czechia’s commitment and positive contribution to global efforts to secure nuclear materials against misuse.

The Nuclear Non-Proliferation Regime: An Historical Perspective

ABSTRACTAs the twenty-first century appears to be entering a more intensified globalised nuclear age, nations and academics alike consistently criticise the frameworks and foundational agencies created to ensure non-proliferation, security, and multi-lateral negotiations. Notwithstanding the significant reductions in nuclear arsenals made over the last several decades, thanks to the Nuclear Non-Proliferation Treaty and the numerous nuclear agencies and watchdogs, it is evident that the risks emanating from nuclear weapons and their associated materials have continually expanded. By ensuring that the global community has protections and safeguards in place to promote international dialogue and nuclear non-proliferation—including the Cooperative Nuclear Threat Reduction Program, International Atomic Energy Agency, Global Threat Reduction Initiative, as well as nuclear free zones and review processes—the nuclear non-proliferation regime can arguably ensure that the devastating threat to humankind is reduced. However, as countries scramble to once again refine and modernise their nuclear stockpiles, the necessity to review, re-shape, and re-think these foundational safeguards is more vital than ever.

Predicting Large Deflections of Multiplate Fuel Elements Using a Monolithic FSI Approach

As part of the Global Threat Reduction Initiative, the Oak Ridge National Laboratory is evaluating conversion of fuel for the High Flux Isotope Reactor (HFIR) from high-enriched uranium to low-enriched uranium. Currently, multiphysics simulations that model fluid-structure interaction phenomena are being performed to ensure the safety of the reactor with the new fuel type. A monolithic solver that fully couples fluid and structural dynamics is used to model deflections in the new design. A classical experiment is chosen to validate the capabilities of the current solver and the method. A single-plate simulation with various boundary conditions as well as a five-plate simulation are presented. Use of the monolithic solver provides stable solutions for the large deflections and the tight coupling of the fluid and structure and the maximum deflections are captured accurately.

Enhanced Radioactive Material Source Security.

Requirements for additional security measures for sealed radioactive sources have evolved since they were first implemented after the terrorist events of 11 September 2001. This paper will describe the sequence of those measures, commencing with the early orders issued by the U.S. Nuclear Regulatory Commission to the May 2013 adoption of 10 CFR Part 37, Physical Protections of Category 1 and Category 2 Quantities of Radioactive Material. Part 37 requirements will be discussed in detail, as the 37 NRC Agreement States, which regulate approximately 88% of the radioactive material licensees, will be required to enact by 19 March 2016. In addition to the Part 37 requirements, the paper will also highlight some of the other ongoing efforts of the U.S. Department of Energy's National Nuclear Security Administration's Global Threat Reduction Initiative and the Conference of Radiation Control Program Directors.

Aluminum cladding oxidation of prefilmed in-pile fueled experiments

A series of fueled irradiation experiments were recently completed within the Advanced Test Reactor Full size plate In center flux trap Position (AFIP) and Gas Test Loop (GTL) campaigns. The conduct of the AFIP experiments supports ongoing efforts within the global threat reduction initiative (GTRI) to qualify a new ultra-high loading density low enriched uranium-molybdenum fuel. This study details the characterization of oxide growth on the fueled AFIP experiments and cross-correlates the empirically measured oxide thickness values to existing oxide growth correlations and convective heat transfer correlations that have traditionally been utilized for such an application. This study adds new and valuable empirical data to the scientific community with respect to oxide growth measurements of highly irradiated experiments, of which there is presently very limited data. Additionally, the predicted oxide thickness values are reconstructed to produce an oxide thickness distribution across the length of each fueled experiment (a new application and presentation of information that has not previously been obtainable in open literature); the predicted distributions are compared against experimental data and in general agree well with the exception of select outliers.

System of Systems Approaches for Mobile Source Transit Security

AbstractThe petroleum industry uses radioactive isotopes in well‐logging operations to support oil exploration activities. Well‐logging service providers use radiological sources to obtain the most accurate estimate of porosity in geologic formations. The National Nuclear Security Administration (NNSA), through its Global Threat Reduction Initiative (GTRI), has a mission to enhance the security of high‐priority nuclear and radiological materials in civilian use. As part of this mission, GTRI intends to facilitate tracking and security of well‐logging radiological sources both in the United States and around the world. GTRI is sponsoring the Pacific Northwest National Laboratory to develop a Mobile Source Transit Security System that can be used by the well‐logging service providers to track these sources in near real time on a worldwide basis. This paper describes the system capabilities, constraints, and conceptual design. System of systems approaches and challenges that were considered in the development of the system are highlighted.

Reactor Physics Scoping and Characterization Study on Implementation of TRIGA Fuel in the Advanced Test Reactor

The Advanced Test Reactor (ATR) is conducting scoping studies for the conversion of its fuel from a highly enriched uranium (HEU) composition to a low-enriched uranium (LEU) composition, through the Reduced Enrichment for Research and Test Reactors Program, within the Global Threat Reduction Initiative. These studies have considered a wide variety of LEU plate-type fuels to replace the current HEU fuel. Continuing to investigate potential alternatives to the present HEU fuel form, this study presents a preliminary reactor physics scoping and feasibility analysis of TRIGA fuel within the current ATR fuel element envelope and compares it to the functional requirements delineated by the Naval Reactors Program, which includes >4.8×1014 fissions/s·g−1 of 235U in test positions, a fast–to–thermal neutron flux ratio that has a <5% deviation from its current value, a desired steady cycle power within the corner lobes, and an operational cycle length of 56 days at 120 MW. Other design parameters outside those put forth by the Naval Reactors Program that are investigated herein include axial and radial power profiles, effective delayed neutron fraction, and mean neutron generation time. The result of this study demonstrates potential promise for implementation of TRIGA fuel in the ATR from a reactor physics perspective; discussion of observations and limitations are provided herein.

Modeling the influence of interaction layer formation on thermal conductivity of U–Mo dispersion fuel

The Global Threat Reduction Initiative Program continues to develop existing and new test reactor fuels to achieve the maximum attainable uranium loadings to support the conversion of a number of the world’s remaining high-enriched uranium fueled reactors to low-enriched uranium fuel. Currently, the program is focused on assisting with the development and qualification of a fuel design that consists of a uranium–molybdenum (U–Mo) alloy dispersed in an aluminum matrix. Thermal conductivity is an important consideration in determining the operational temperature of the fuel and can be influenced by interaction layer formation between the dispersed phase and matrix, porosity that forms during fabrication of the fuel plates or rods, and upon the concentration of the dispersed phase within the matrix. This paper develops and validates a simple model to study the influence of interaction layer formation, dispersed particle size, and volume fraction of dispersed phase in the matrix on the effective conductivity of the composite. The model shows excellent agreement with results previously presented in the literature. In particular, the thermal conductivity of the interaction layer does not appear to be as important in determining the effective conductivity of the composite, while formation of the interaction layer and subsequent consumption of the matrix reveals a rather significant effect. The effective thermal conductivity of the composite can be influenced by the dispersed particle distribution by minimizing interaction layer formation and preserving the higher thermal conductivity matrix.

Internal contamination of an irradiator discovered during security enhancement.

High-risk radioactive sources regulated under Increased Controls Regulations have been protected by licensed facilities, but the federal government has placed significant emphasis on these sources and has developed initiatives to assist radioactive material licensees. The Department of Energy's Global Threat Reduction Initiative (GTRI) Domestic Threat Reduction Program is a voluntary federally funded program for security enhancements of high-risk radiological material. During the hardening or security enhancement process by the United States Department of Energy (U.S. DOE) contractors, a small amount of radioactive contamination was discovered in a Cesium irradiator. Ultimately, it was decided to pursue disposal with U.S. DOE's Off-Site Recovery Program (OSRP). Radiological devices may have a leaking source or known internal contamination that may cause difficulty during security enhancement. If the licensee understands this, it may provide facilities the opportunity to plan and prepare for unusual circumstances.

Microstructural characteristics of HIP-bonded monolithic nuclear fuels with a diffusion barrier

Due to the limitation of maximum uranium load achievable by dispersion fuel type, the Global Threat Reduction Initiative is developing an advanced monolithic fuel to convert US high-performance research reactors to low-enriched uranium. Hot-isostatic-press (HIP) bonding was the single process down-selected to bond monolithic U–Mo fuel meat to aluminum alloy cladding. A diffusion barrier was applied to the U–Mo fuel meat by roll-bonding process to prevent extensive interaction between fuel meat and aluminum-alloy cladding. Microstructural characterization was performed on fresh fuel plates fabricated at Idaho National Laboratory. Interfaces between the fuel meat, the cladding, and the diffusion barrier, as well as between the U–10Mo fuel meat and the Al-6061 cladding, were characterized by scanning electron microscopy. Preliminary results indicate that the interfaces contain many different phases while decomposition, second phases, and chemical banding were also observed in the fuel meat. The important attributes of the HIP-bonded monolithic fuel are:•A typical Zr diffusion barrier with a thickness of 25μm.•A transverse cross section that exhibits relatively equiaxed grains with an average grain diameter of 10μm.•Chemical banding, in some areas more than 100μm in length, that is very pronounced in longitudinal (i.e., rolling) direction with Mo concentration varying from 7–13wt.%.•Decomposed areas containing plate-shaped low-Mo phase.•A typical Zr/cladding interaction layer with a thickness of 1–2μm.•A visible UZr2 bearing layer with a thickness of 1–2μm.•Mo-rich precipitates (mainly Mo2Zr, forming a layer in some areas) followed by a Mo-depleted sub-layer between the visible UZr2-bearing layer and the U–Mo matrix.•No excessive interaction between cladding and the uncoated fuel edge.•Cladding-to-cladding bonding that exhibits no cracks or porosity with second phases high in Mg, Si, and O decorating the bond line.•Some of these attributes might be critical to the irradiation performance of monolithic U–10Mo nuclear fuel.There are several issues or concerns that warrant more detailed study, such as precipitation along the cladding-to-cladding bond line, chemical banding, uncovered fuel-zone edge, and the interaction layer between the U–Mo fuel meat and zirconium. Future post-irradiation examination results will focus, among other things, on identifying in-reactor failure mechanisms and, eventually, directing further fresh fuel characterization efforts.

Predicting critical flow velocity leading to laminate plate collapse—flat plates

The Oregon State University (OSU), Hydro Mechanical Fuel test Facility (HMFTF) is designed to hydro-mechanically test prototypic plate type fuel. OSU's fuel test program is a part of the Global Threat Reduction Initiative (GTRI), formerly known as the Reduced Enrichment for Research and Test Reactor program. One of the GTRI's goals is to convert all civilian research, and test reactors in the United States from highly enriched uranium (HEU) to a low enriched uranium (LEU) fuel in an effort to reduce nuclear proliferation.An analytical model has been developed and is described in detail which complements the experimental work being performed by the OSU HMFTF, and advances the science of hydro-mechanics. This study investigates two methods for determining the ‘critical flow velocity’ for a laminate plate. The objective is accomplished by incorporating a flexural rigidity term into the formulation of critical flow velocity originally derived by Donald R. Miller, and employing sandwich structure theory to determine the rigidity term. The final outcome of this study results in the developing of a single equation for each of three different edge boundary conditions which reliably and comprehensively predicts the onset of plate collapse. The two models developed and presented, are termed the monocoque analogy and the ideal laminate model. Of these two models, the ideal laminate model is the most resolved and comprehensive in its predictions.

Improved Irradiation Performance of Uranium-Molybdenum/Aluminum Dispersion Fuel by Silicon Addition in Aluminum

Uranium-molybdenum fuel particle dispersion in aluminum is a form of fuel under development for conversion of high-power research and test reactors from highly enriched to low-enriched uranium in the U.S. Global Threat Reduction Initiative program (also known as the Reduced Enrichment for Research and Test Reactors program). Extensive irradiation tests have been conducted to find a solution for problems caused by interaction layer growth and pore formation between U-Mo and Al. Adding a small amount of Si (up to ~5 wt%) in the Al matrix was one of the proposed remedies. The effect of silicon addition in the Al matrix was examined using irradiation test results by comparing side-by-side samples with different Si additions. Interaction layer growth was progressively reduced with increasing Si addition to the matrix Al, up to 4.8 wt%. The Si addition also appeared to delay pore formation and growth between the U-Mo and Al.

AN APPLICATION OF IRIDM IN THE DECISION MAKING PROCESS ON FUEL CONVERSION OF THE MARIA REACTOR

Poland, when acceded to GTRI (Global Threat Reduction Initiative) in 2004, has committed to convert the nuclear fuel of the Research Reactor MARIA, operated by the National Centre for Nuclear Research (NCBJ) in Świerk. The conversion means giving up of high enriched uranium fuel containing 36% of U-235, which was used so far, and replacing it with the low enriched uranium fuel (19.7% U-235). This article describes the potential usability of the Integrated Risk Informed Decision Making (IRIDM) methodology in optimization of the fuel conversion procedure.

Aqueous processing of U–10Mo scrap for high performance research reactor fuel

The Global Threat Reduction Initiative (GTRI) Conversion program, which is part of the US government’s National Nuclear Security Administration (NNSA), supports the conversion of civilian use of highly enriched uranium (HEU) to low enriched uranium (LEU) for reactor fuel and targets. The reason for conversion is to eliminate the use of any material that may pose a threat to the United States or other foreign countries. High performance research reactors (HPRRs) cannot make the conversion to a standard LEU fuel because they require a more dense fuel to meet their performance requirements. As a result, a more dense fuel consisting of a monolithic uranium–molybdenum alloy containing 10% (w/w) Mo with Al cladding and a Zr bonding-layer is being considered. Significant losses are expected in the fabrication of this fuel, so a means to recycle the scrap pieces is needed. Argonne National Laboratory has developed an aqueous-processing flowsheet for scrap recovery in the fuel fabrication process for high-density LEU-monolithic fuel based on data found in the literature. Experiments have been performed to investigate dissolution conditions for solutions containing approximately 20g-U/L and 50g-U/L with and without Fe(NO3)3. HNO3 and HF concentrations have been optimized for timely dissolution of the fuel scrap and prevention of the formation of the U–Zr2 intermetallic, explosive complex, while meeting the requirements needed for further processing.

IAEA assistance and training programme for transport security

The IAEA Office of Nuclear Security is working cooperatively with the U.S. Department of Energy's Global Threat Reduction Initiative, European Union and Australia to provide transport security assistance to countries throughout the world. Assistance is available to countries in reviewing and upgrading their transport security programs at all levels: (1) National level (regulatory and other government agencies); and (2) Operator level (shippers and carriers). Assistance is directed at implementing a consistent level of security throughout the life cycle of radioactive material (same level of security during transport as when in a fixed facility) Upgrade assistance can include: (1) Expert advisory missions to provide advice and guidance; (2) Training courses for regulatory, governmental and industry personnel; (3) Transport security awareness; (4) Detailed training on designing and implementing transport security programs; (5) Planning to identify and prioritize needs (developing security approaches and plans); (6) Developing model security plans and procedures; and (7) Equipment (vehicles, packages, command and control equipment, etc.). Country visits are now being scheduled to initiate transport security cooperative activities. A training course has been developed to assist countries in developing and implementing transport security programs. The training course has been given as a national training course (three times) andmore » as a Regional training course (three times). The course addresses recommended security provisions for the transport of all radioactive material.« less

From High-Enriched to Low-Enriched Uranium Fuel in Research Reactors

Since the 1970's, global efforts have been going on to replace the high-enriched (>90% 235 U), low-density UAI x research reactor fuel with high-density, low-enriched (<20% 235 U) replacements. This search is driven by the attempt to reduce the civil use of high-enriched material because of proliferation risks and terrorist threats. American initiatives, such as the Global Threat Reduction Initiative (GTRI) and the Reduced Enrichment for Research and Test Reactors (RERTR) program have triggered the development of reliable low-enriched fuel types for these reactors, which can replace the high-enriched ones without loss of performance. Most success has presently been obtained with U 3 Si 2 dispersion fuel, which is currently used in many research reactors in the world. However, efforts to search for a replacement with even higher density, which will also allow the conversion of some high flux research reactors that currently cannot change to U 3 Si 2 (eg. BR2 in Belgium), have continued and are for the moment mainly directed towards the U(Mo) alloy fuel (7-10 w% Mo). This paper provides an overview of the past efforts and presents the current status of the U(Mo) development.

Thermo-physical properties of DU–10 wt.% Mo alloys

Low-enriched uranium alloyed with 10 wt.% molybdenum is under consideration by the Global Threat Reduction Initiative reactor convert program as a very high density fuel to enable the conversion of high-performance research reactors away from highly-enriched uranium fuels. As with any fuel development program, the thermo-physical properties of the fuel as a function of temperature are extremely important and must be well characterized in order to effectively model and predict fuel behavior under normal and off-normal irradiation conditions. For the alloy system under investigation, the available thermo-physical property data is relatively inconsistent and often lacks appropriate explanation. Available literature on this alloy system comes mainly from studies done during the 1960s and 1970s, and often does not include sufficient information on fabrication history or conditions to draw conclusions for the current application. The current paper has investigated specific heat capacity, coefficient of linear thermal expansion, density, and thermal diffusivity that were then used to calculate alloy thermal conductivity as a function of temperature. The data obtained from this investigation was compared to available literature on similar U–Mo alloys, and in most cases are in good agreement.

Lack of international consensus on the disposition and storage of disused sealed sources

A lower-activity analogue of the trans-national problem of spent fuel management and disposal is the global problem of radioactive sealed source [source: The IAEA definition of a sealed source is “Radioactive material that is permanently sealed in a capsule or closely bonded and in a solid form.” Taken from glossary of Nuclear Waste Data Management found at http://www-ewmdb.iaea.org/showhelp.asp?Topic=8-1-1.] disposal. Sources are found in almost every country in the world because of their beneficial medical and commercial or industrial applications. Some of the isotopes used have short half-lives—iridium-192 (Ir-192), 73.8 days—while others have very long half-lives—americium-241 (Am-241), 432 years or plutonium-239 (Pu-239), 24,130 years. It is critically important, particularly for longer-lived isotopes, to find final disposition pathways. Lack of a permanent disposition pathway such as recycling or irretrievable disposal creates numerous problems, including the potential loss of regulatory control, which increases the risk of inadvertent or deliberate misuse of the material. The misuse of radioactive materials has the potential for substantial public health and economic damage. Disused sources also pose an inherent risk to the end-users from a liability, safety, and public health perspectives. This paper examines various disposition pathways employed by several key source manufacturing or possessing nation-states for disused sources. Examples of source disposition pathways include long-term storage, deep geological disposal, borehole disposal and shallow land burial. The Off-Site Source Recovery Project (OSRP), part of the office of Global Threat Reduction Initiative (GTRI), acts as an intermediary in the recovery and ultimate disposition of US origin sealed radiological materials. Several concepts that could help mitigate the challenge of a lack of long-term disposition options for sources are available, but these tools have not yet been applied by most nation-states. For example, regional consolidation and repatriation of sources to the country of manufacture would ease or eliminate the need for in situ disposal or storage in a number of developing nation-states.

Presidential Candidates Speak Out on Science Policies

Presidential Candidates Speak Out on Science Policies