Reactor Program Research Articles

Experimental and modeling researches of dust particles in the HL-2A tokamak**Project supported by the National Magnetic Confinement Fusion Science Program of China (Grant Nos. 2014GB107000 and 2013GB112008), the National Natural Science Foundation of China (Grant Nos. 11320101005, 11175060, 11375054, and 11075046), and the China–Korean Joint Foundation (Grant No. 2012DFG02230).

The investigation of dust particle characteristics in fusion devices has become more and more imperative. In the HL-2A tokamak, the morphologies and compositions of dust particles are analyzed by using scanning electron microscopy (SEM) and energy dispersive x-ray spectroscopy (EDX) with mapping. The results indicate that the sizes of dust particles are in a range from 1 μm to 1 mm. Surprisingly, stainless steel spheres with a diameter of 2.5 μm–30 μm are obtained. The production mechanisms of dust particles include flaking, disintegration, agglomeration, and arcing. In addition, dynamic characteristics of the flaking dust particles are observed by a CMOS fast framing camera and simulated by a computer program. Both of the results display that the ion friction force is dominant in the toroidal direction, while the centrifugal force is crucial in the radial direction. Therefore, the visible dust particles are accelerated toriodally by the ion friction force and migrated radially by the centrifugal force. The averaged velocity of the grain is on the order of ∼ 100 m/s. These results provide an additional supplement for one of critical plasma-wall interaction (PWI) issues in the framework of the International Thermonuclear Experimental Reactor (ITER) programme.

Atom probe tomography analysis of high dose MA957 at selected irradiation temperatures

Oxide dispersion strengthened (ODS) alloys are meritable structural materials for nuclear reactor systems due to the exemplary resistance to radiation damage and high temperature creep. Summarized in this work are atom probe tomography (APT) investigations on a heat of MA957 that underwent irradiation in the form of in-reactor creep specimens in the Fast Flux Test Facility–Materials Open Test Assembly (FFTF–MOTA) for the Liquid Metal Fast Breeder Reactor (LMFBR) program. The oxide precipitates appear stable under irradiation at elevated temperature over extended periods of time. Nominally, the precipitate chemistry is unchanged by the accumulated dose; although, evidence suggests that ballistic dissolution and reformation processes are occurring at all irradiation temperatures. At 412°C–109dpa, chromium enrichments – consistent with the α′ phase – appear between the oxide precipitates, indicating radiation induced segregation. Grain boundaries, enriched with several elements including nickel and titanium, are observed at all irradiation conditions. At 412°C–109dpa, the grain boundaries are also enriched in molecular titanium oxide (TiO).

核能数值反应堆国内外研究现状及进展

本文介绍了核能先进仿真建模领域的最新进展及数值反应堆的概念，对美国及欧洲针对核能数值反应堆的相关研发项目如CASL、NEAMS、NURESIM等进行了详细的调研，对项目的研究内容、研究方法等进行了阐述，结合国内外的发展情况，总结了核能数值反应堆发展的关键技术和挑战，可为我国开展核能数值反应堆研究提供技术支持。 The latest progress in the field of nuclear energy advanced modeling & simulation is introduced in this paper, in which the meaning of the nuclear virtual reactor or numerical nuclear reactor is re-viewed. A detailed research and introduction about nuclear virtual reactor programs in the United State and Europe, such as CASL, NEAMS and NURESIM are addressed. Based on the research, the key technology and challenge of developing a nuclear virtual reactor are summarized, which can give the support to the further study of the nuclear virtual reactor.

ANALYSIS OF METALLIC MATERIALS FOR ITER WITH THE EMPHASIS ON COPPER ALLOYS

The production of electrical energy from nuclear fusions is currently highly advanced field of the interest in an international scientific cooperation in the International Thermonuclear Experimental Reactor programme (ITER). Alloys based on Cu are used in the ITER project however from the future perspective in terms of the physical and mechanical properties improvement, the alloy based on CuCrZr is proposed. The CuCrZr alloy is classified as a precipitation-hardened material which is used as a functional material for various applications in ITER, for example the first wall, blanket and support, divertor and heating systems. In this study, as an experimental material was used CuCrZr alloy with the chemical composition of 98.54wt.% Cu, 1.1 wt.% Cr, 0.043 wt.% Zr and 0.317 wt.% impurities. There was used ECAP method and heat treatment to enhance the strength properties in CuCrZr alloy. The application of sophisticated treatments based on solution annealing, structure formation by ECAP on nanoscale level and additional artificial ageing producing precipitation strengthening can be achieved the yield strength more than 500MPa.

Criticality Benchmarks for the Epithermal Test Assembly Experiments

The Epithermal Test Assembly (ETA) experiments were performed to test the adequacy of 233U, 235U, and 232Th cross sections in epithermal spectra in support of the Light Water Breeder Reactor (LWBR) Program. The ETA design contained a central heavy water–moderated test region surrounded by a light water–moderated annular driver region. Two series of experiments were performed: ETA-I with 235UO2-ThO2 fuel rods in the test region and ETA-II with 233UO2-ThO2 fuel rods in the test region. The dominant uncertainties in the critical configurations include the test-rod pitch, fuel density, and ThO2 mass for ETA-I; the test-region fuel-rod fuel density and 233U to (233U + Th) weight ratio for ETA-II; and the driver-region fuel-rod outer diameter, uranium enrichment, and pitch for both ETA experiments. Benchmark model results using MCNP5 are provided for ENDF/B-V, ENDF/B-VI, ENDF/B-VII.0, and ENDF/B-VII.1 cross sections with only the ENDF/B-VII.0 results falling within three standard deviations of the benchmark model keff. The ETA-I and ETA-II benchmark evaluations have been included in the International Handbook of Evaluated Criticality Safety Benchmark Experiments and are replicated in the International Handbook of Evaluated Reactor Physics Benchmark Experiments.

Interdiffusion and reactions between U–Mo and Zr at 650 °C as a function of time

Development of monolithic U–Mo alloy fuel (typically U–10wt.%Mo) for the Reduced Enrichment for Research and Test Reactors (RERTR) program entails a use of Zr diffusion barrier to eliminate the interdiffusion–reactions between the fuel alloy and Al-alloy cladding. The application of Zr barrier to the U–Mo fuel system requires a co-rolling process that utilizes a soaking temperature of 650°C, which represents the highest temperature the fuel system is exposed to during both fuel manufacturing and reactor application. Therefore, in this study, development of phase constituents, microstructure and diffusion kinetics of U–10wt.%Mo and Zr was examined using solid-to-solid diffusion couples annealed at 650°C for 240, 480 and 720h. Phase constituents and microstructural development were analyzed by scanning electron microscopy (SEM) and transmission electron microscopy (TEM). Concentration profiles were mapped as diffusion paths on the isothermal ternary phase diagram. Within the diffusion zone, single-phase layers of β-Zr and β-U were observed along with a discontinuous layer of Mo2Zr between the β-Zr and β-U layers. In the vicinity of Mo2Zr phase, islands of α-Zr phases were also found. In addition, acicular α-Zr and U6Zr3Mo phases were observed within the γ-U(Mo) terminal alloy. Growth rate of the interdiffusion–reaction zone was determined to be 7.75(±5.84)×10−16m2/s at 650°C, however with an assumption of a certain incubation period.

A summary of sodium-cooled fast reactor development

Much of the basic technology for the Sodium-cooled fast Reactor (SFR) has been established through long term development experience with former fast reactor programs, and is being confirmed by the Phénix end-of-life tests in France, the restart of Monju in Japan, the lifetime extension of BN-600 in Russia, and the startup of the China Experimental Fast Reactor in China. Planned startup in 2014 for new SFRs: BN-800 in Russia and PFBR in India, will further enhance the confirmation of the SFR basic technology. Nowadays, the SFR development has advanced to aiming at establishment of the Generation-IV system which is dedicated to sustainable energy generation and actinide management, and several advanced SFR concepts are under development such as PRISM, JSFR, ASTRID, PGSFR, BN-1200, and CFR-600. Generation-IV International Forum is an international collaboration framework where various R&D activities are progressing on design of system and component, safety and operation, advanced fuel, and actinide cycle for the Generation-IV SFR development, and will play a beneficial role of promoting them thorough providing an opportunity to share the past experience and the latest data of design and R&D among countries developing SFR.

Contact Stress Evaluation Based on Thermal-Structural Analysis in KO HCCR TBM Submodules for ITER

The Korean Helium Cooled Ceramic Reflector Test Blanket Module (TBM) consisted of four submodules for International Thermonuclear Experimental Reactor program. The thermal-structural analysis considered a design pressure (10 MPa) and a thermal stress in the changed first wall design, which has a 15 mm \(\times 11\) mm rectangular U-shape, was carried out according to the RCC-MR code. The contact stress between the TBM submodules was also evaluated to confirm the structural integrity. The results showed that the design requirements were satisfied.

Generating Tafel Parameters in Support of Electrorefining Modeling for Uranium Recovery from Scrap U-Mo Foils

A model is under development to simulate operation of an electrorefiner for the recovery of uranium from scrap U-Mo foils. U-Mo foils are proposed as a low enriched uranium (LEU) fuel alternative, for research and test reactors currently using highly enriched fuels, as part of the Reduced Enrichment for Research and Test Reactors program. The conversion to LEU fuels will decrease the risk of proliferation at research and test reactors. The model being developed will be used to design and optimize the operation of electrorefining for scrap uranium recovery from the LEU fuel manufacture, thereby improving efficiency of the manufacturing process.Electrorefining for uranium recovery from metal fuel scrap is traditionally conducted in a molten eutectic LiCl/KCl salt electrolyte with a dilute amount of UCl3 in the salt at 500°C. The uranium is then dissolved at the anode forming U3+ ions and deposited at the cathode as metallic uranium in response to an applied potential. The advantageous choice of operating potential ensures uranium is the only species being transported out of the scrap fuel material. The uranium collected at the cathode is then available for recycle to fuel manufacturing.The model implements classic corrosion mathematics to determine polarization at the anode and cathode. Polarization at the anode is complicated by formation of a passive layer. Polarization at the cathode is complicated by mixed polarization, caused by the low concentration of uranium in the electrolyte. Transport across the cell is driven by electronic migration and diffusion. The concentration at each electrode is calculated using the diffusion layer thickness. The model is constructed in MatLab with a user interface and simulation results are exported to Excel for analysis.The electrolyte, LiCl/KCl 5-10 weight percent UCl3, used in electrorefining of uranium has not been studied extensively via corrosion research techniques. Therefore, the Tafel constants and exchange current densities are currently unavailable for uranium in this electrolyte. Experiments have been conducted to generate these parameters for materials of interest during the electrorefining of U-Mo foils. A three electrode cell with a pseudo-reference tungsten electrode was used to scan potentiodynamically across both anodic and cathodic overpotentials. The potentiodynamic data were normalized to zero overpotential at reaction. The normalized data were examined in the ±4 to 10 mV of overpotential range for Tafel constants, both anodic and cathodic. This overpotential range was chosen in accordance with the standard method for analyzing potentiodynamic data outlined by Scully [1]. The exchange current density was found using the intersection of Tafel lines constructed from the same region of overpotential.The generation of these parameters will contribute significantly to the basic chemistry knowledge of the electrorefiner system for these materials in this electrolyte. It will also contribute significantly to the development of the mathematical model. Simulations of electrorefining of several types of metallic LEU fuels using the model are proposed to validate the mathematical methods used, including the Tafel parameters.

The Small Modular Reactor Program in the United States

The United States is developing clean energy alternatives to meet President Barack Obama’s “All of the Above Clean Energy Strategy” with a goal of generating 80 percent of electricity from clean energy and thereby reducing carbon emissions in half by the year 2035. The U.S. Department of Energy believes that nuclear energy is essential to reach these goals and Small Modular Reactors are a very innovative technology option. Small Modular Reactors have many desirable attributes and the markets look promising. Demonstration of this technology over the next decade can lead to the next generation of safe, reliable, and clean nuclear energy.

Radiation Exposure of U.S. Military Individuals

The U.S. military consists of five armed services: the Army, Navy, Marine Corps, Air Force, and Coast Guard. It directly employs 1.4 million active duty military, 1.3 million National Guard and reserve military, and 700,000 civilian individuals. This paper describes the military guidance used to preserve and maintain the health of military personnel while they accomplish necessary and purposeful work in areas where they are exposed to radiation. It also discusses military exposure cohorts and associated radiogenic disease compensation programs administered by the U.S. Department of Veterans Affairs, the U.S. Department of Justice, and the U.S. Department of Labor. With a few exceptions, the U.S. military has effectively employed ionizing radiation since it was first introduced during the Spanish-American War in 1898. The U.S military annually monitors 70,000 individuals for occupational radiation exposure: ~2% of its workforce. In recent years, the Departments of the Navy (including the Marine Corps), the Army, and the Air Force all have a low collective dose that remains close to 1 person-Sv annually. Only a few Coast Guard individuals are now routinely monitored for radiation exposure. As with the nuclear industry as a whole, the Naval Reactors program has a higher collective dose than the remainder of the U.S. military. The U.S. military maintains occupational radiation exposure records on over two million individuals from 1945 through the present. These records are controlled in accordance with the Privacy Act of 1974 but are available to affected individuals or their designees and other groups performing sanctioned epidemiology studies.Introduction of Radiation Exposure of U.S. Military Individuals (Video 2:19, http://links.lww.com/HP/A30).

Beyond ITER: Neutral beams for a demonstration fusion reactor (DEMO) (invited)

In the development of magnetically confined fusion as an economically sustainable power source, International Tokamak Experimental Reactor (ITER) is currently under construction. Beyond ITER is the demonstration fusion reactor (DEMO) programme in which the physics and engineering aspects of a future fusion power plant will be demonstrated. DEMO will produce net electrical power. The DEMO programme will be outlined and the role of neutral beams for heating and current drive will be described. In particular, the importance of the efficiency of neutral beam systems in terms of injected neutral beam power compared to wallplug power will be discussed. Options for improving this efficiency including advanced neutralisers and energy recovery are discussed.

Joining Silicon Carbide Plates by Titanium Disilicide-Based Compound

The present work is part of the Fourth Generation Reactor Program, which describes the methodology and results for joining SiC substrates by metallic silicides with SiC powder reinforcements. The severest temperatures in service are in the range of 1000 °C but short-time incursions at 1600 or 2000 °C have to be anticipated. One of the key issues is the joining of SiC f /SiC m composites to seal the combustible cladding. We describe the results for joining SiC substrates in liquid state using TiSi2. Joint integrity and joint strength can be improved by adding small SiC particles to the silicides powders. The assemblies are obtained in an inductive furnace. Cross-sections of the assembly, wettability tests, thermo-mechanical properties, and four-point bending tests are presented.

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.

How Fundamentals of Phase Equilibria and Diffusion Contribute to the RERTR program

One of the oldest programs of the U.S. Department of Energy is the Reduced Enrichment for Research and Test Reactors (RERTR) program. This program develops technology necessary to convert both research and test reactors, including some medical purpose reactors, from using high enriched uranium (HEU) to using low enriched uranium (LEU) fuels and targets. This conversion is related to non-proliferation by making weapons grade nuclear material less available. In order to pursue their goal, the RERTR Program, managed by the Office of Nuclear Material Threat Reduction in the National Nuclear Security Administration, has held annual international conferences since its inception in 1978.

Diffusion Barrier Selection from Refractory Metals (Zr, Mo and Nb) Via Interdiffusion Investigation for U-Mo RERTR Fuel Alloy

U-Mo alloys are being developed as low enrichment monolithic fuel under the Reduced Enrichment for Research and Test Reactor (RERTR) program. Diffusional interactions between the U-Mo fuel alloy and Al-alloy cladding within the monolithic fuel plate construct necessitate incorporation of a barrier layer. Fundamentally, a diffusion barrier candidate must have good thermal conductivity, high melting point, minimal metallurgical interaction, and good irradiation performance. Refractory metals, Zr, Mo, and Nb are considered based on their physical properties, and the diffusion behavior must be carefully examined first with U-Mo fuel alloy. Solid-to-solid U-10 wt.%Mo versus Mo, Zr, or Nb diffusion couples were assembled and annealed at 600, 700, 800, 900 and 1000 °C for various times. The interdiffusion microstructures and chemical composition were examined via scanning electron microscopy and electron probe microanalysis, respectively. For all three systems, the growth rate of interdiffusion zone were calculated at 1000, 900 and 800 °C under the assumption of parabolic growth, and calculated for lower temperature of 700, 600 and 500 °C according to Arrhenius relationship. The growth rate was determined to be about 103 times slower for Zr, 105 times slower for Mo and 106 times slower for Nb, than the growth rates reported for the interaction between the U-Mo fuel alloy and pure Al or Al-Si cladding alloys. Zr, however was selected as the barrier metal due to a concern for thermo-mechanical behavior of UMo/Nb interface observed from diffusion couples, and for ductile-to-brittle transition of Mo near room temperature.

Indian nuclear power programme – Past, present and future

The country is facing energy shortages at present for the connected load. A large population even now does not have access to electricity. As the economy expands and the population increases, the country will need to generate even more power to meet the growing demand. Resources of energy are experiencing constraints to meet the current demand. The generating capacities in the country are underperforming at present because of fuel supply constraint. Analysis in the Energy Policy document hints that conventional fuel resources will begin to exhaust by middle of the century. All these indicate that for India to be able to meet its growing demand, nuclear and solar power could provide energy security in the long term. The first stage of commercially successful nuclear power programme has indicated that country has command on the technology through its own R&D base built since the beginning of establishing Department of Atomic Energy. On this strength it could withstand the technology denial regime for years. The second stage (Fast Breeder Reactor) programme has been initiated. Given the scientific and technological capability demonstrated so far, the technology required for optimizing the second stage programme and launching the third stage programme can be developed to assure long term energy security at the desired capacity.

Indian fast reactor technology: Current status and future programme

The paper brings out the advantages of fast breeder reactor and importance of developing closed nuclear fuel cycle for the large scale energy production, which is followed by its salient safety features. Further, the current status and future strategy of the fast reactor programme since the inception through 40 MWt/13 MWe Fast Breeder Test Reactor (FBTR), is highlighted. The challenges and achievements in science and technology of FBRs focusing on safety are described with the particular reference to 500 MWe capacity Prototype Fast Breeder Reactor (PFBR), being commissioned at Kalpakkam. Roadmap with comprehensive R&D for the large scale deployment of Sodium Cooled Fast Reactor (SFRs) and timely introduction of metallic fuel reactors with emphasis on breeding gain and enhanced safety are being brought out in this paper.

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.

Development of IFAC-1 SS: An Advanced Austenitic Stainless Steel for Cladding and Wrapper Tube Applications in Sodium-Cooled Fast Reactors

Fuel cycle cost of sodium cooled fast reactors (SFRs) is strongly dependent on the in-reactor performance of core structural materials, i.e., cladding and wrapper tube materials of the fuel subassembly, which are subjected to intense neutron irradiation during service, leading to unique materials problems like void swelling, irradiation creep and helium embrittlement. In order to increase the burnup of the fuel and thereby reduce the fuel cycle cost, it is necessary to employ materials which have high resistance to void swelling as well as better high temperature mechanical properties. The Indian fast reactor program began with the commissioning of the 40 MWtFast Breeder Test Reactor (FBTR). The core structural material of FBTR is 20% cold worked 316 austenitic stainless steel (SS). For the 5000 MetPrototype Fast Breeder Reactor (PFBR) which is in an advanced stage of construction at Kalpakkam, 20% cold-worked alloy D9 (14Cr-15Ni-Ti SS) has been selected as the cladding and wrapper tube material for the initial core. The target burnup of the fuel is 100 GWd/t. Advanced austenitic stainless steel and oxide dispersion strengthened steels are being developed for achieving fuel burnup higher than 100 GWd/t. An advanced alloy D9 containing controlled amounts of titanium, silicon and phosphorous has been developed. This alloy named as IFAC-1 (Indian Fast Reactor advanced Clad-1) SS is aimed at thermal creep properties comparable to that of alloy D9, and superior void swelling resistance upto a target burn-up of about 150 GWd/t. The nominal chemical composition of IFAC-1 SS is 14Cr-15Ni-.25Ti-.75Si-.04P. The chemical composition has been optimized after extensive evaluation of the tensile, creep and microstructural stability of fifteen laboratory heats with different amounts of titanium, silicon and phosphorous. Void swelling behavior was studied using ion irradiation. IFAC-1 SS contains higher levels of low melting eutectic phase forming elements such as phosphorous, and so is susceptible to solidification cracking. Extensive pulsed TIG welding trials have been carried out on IFAC-1 SS/316LN SS weld joints with varied weld parameters to find out the feasibility of obtaining solidification crack-free welds and the optimum welding parameters have been established. This paper gives an overview of the development of this advanced core structural material for SFRs.