Advanced Nuclear Energy Systems Research Articles

Microstructure and mechanical properties of 16 Cr-ODS ferritic steel for advanced nuclear energy system

A 16Cr-0.5Ti-lW-0.35Y2O3 oxide dispersion strengthened (ODS) ferritic steel was fabricated by mechanically alloying and hot isostatic pressing (HIP). Subsequent thermo-mechanical treatments were performed to improve the microstructure homogeneous and service properties of the HIPed 16Cr-ODS steel. Nano-oxide particles were observed by TEM, which can be identified to be (Y, Ti) complex oxide by EDS and SAED. The mechanical property was measured by tensile test, and the oxidation behaviour of the ODS steel was performed at high temperature in a muffle.

The Supersafe© Reactor: A Small Modular Pressure Tube SCWR

The SUPERSAFE© Reactor (SSR) is proposed as a small modular version of the Canadian supercritical water-cooled reactor (SCWR). The SCWR is Canada’s primary contribution to the Generation-IV (GEN-IV) International Forum’s (GIF) research and development effort toward the study and eventual deployment of advanced nuclear energy systems. All GEN-IV concepts, including the SCWR, have enhanced safety, improved economics, improved sustainability and enhanced security compared to contemporary reactors. The SUPERSAFE© Reactor (SSR) concept incorporates the enhanced features of the SCWR in a smaller core which could be deployed in areas with sparsely distributed population bases where it is impractical to have a full scale SCWR or large centralized energy grid. An overview of the SSR concept is presented in this work.

Nanomaterials and nanotechnologies in nuclear energy chemistry

Abstract With the rapid growth of human demands for nuclear energy and in response to the challenges of nuclear energy development, the world’s major nuclear countries have started research and development work on advanced nuclear energy systems in which new materials and new technologies are considered to play important roles. Nanomaterials and nanotechnologies, which have gained extensive attention in recent years, have shown a wide range of application potentials in future nuclear energy system. In this review, the basic research progress in nanomaterials and nanotechnologies for advanced nuclear fuel fabrication, spent nuclear fuel reprocessing, nuclear waste disposal and nuclear environmental remediation is selectively highlighted, with the emphasis on Chinese research achievements. In addition, the challenges and opportunities of nanomaterials and nanotechnologies in future advanced nuclear energy system are also discussed.

Applying the GIF PR&PP Methodology for a Qualitative Analysis of a Misuse Scenario in a Notional Generation IV Example Sodium Fast Reactor

The Generation IV International Forum (GIF) Proliferation Resistance and Physical Protection (PR&PP) Working Group has developed a methodology for the PR&PP evaluation of advanced nuclear energy systems (NESs). A notional sodium-cooled fast neutron nuclear reactor system, named the Example Sodium Fast Reactor (ESFR), was used as a case study for the development and demonstration of the GIF PR&PP evaluation methodology. This paper presents some of the results of the application of the GIF PR&PP evaluation methodology to a misuse scenario involving the ESFR. The ESFR baseline design and two design variations are addressed. Rather than presenting a complete evaluation of all the possible misuse scenarios, the paper concentrates on methodological aspects and illustrates how a qualitative analysis following the GIF PR&PP evaluation methodology can generate traceable results of the considered design variations and provide useful feedback for both system and safeguards designers as well.

Proliferation Resistance and Physical Protection Evaluation Methodology: Objectives, Accomplishments, and Future Directions

An overview is presented of the objectives, accomplishments, and potential future directions of the program on the evaluation methodology for proliferation resistance and physical protection (PR&PP) of advanced nuclear energy systems. The PR&PP Working Group of the Generation IV International Forum developed the methodology through a series of demonstration and case studies. The results of the evaluations performed with the methodology are intended for three types of users: system designers, program policy makers, and external stakeholders. During the past few years, various national and international groups have applied the methodology to nuclear energy system design as well as to developing approaches to advanced safeguards. We suggest some future applications of the methodology in this paper.

Rapid Self-Assembly of Uranyl Polyhedra into Crown Clusters

Clusters built from 32 uranyl peroxide polyhedra self-assemble and crystallize within 15 min after combining uranyl nitrate, ammonium hydroxide, and hydrogen peroxide in aqueous solution under ambient conditions. These novel crown-shaped clusters are remarkable in that they form so quickly, have extraordinarily low aqueous solubility, form with at least two distinct peroxide to hydroxyl ratios, and form in very high yield. The clusters, which have outer diameters of 23 Å, topologically consist of eight pentagons and four hexagons. Their rapid formation and low solubility in aqueous systems may be useful properties at various stages in an advanced nuclear energy system.

Benchmarking of thermal hydraulic loop models for Lead-Alloy Cooled Advanced Nuclear Energy System (LACANES), phase-I: Isothermal steady state forced convection

As highly promising coolant for new generation nuclear reactors, liquid Lead–Bismuth Eutectic has been extensively worldwide investigated. With high expectation about this advanced coolant, a multi-national systematic study on LBE was proposed in 2007, which covers benchmarking of thermal hydraulic prediction models for Lead-Alloy Cooled Advanced Nuclear Energy System (LACANES). This international collaboration has been organized by OECD/NEA, and nine organizations – ENEA, ERSE, GIDROPRESS, IAEA, IPPE, KIT/IKET, KIT/INR, NUTRECK, and RRC KI – contribute their efforts to LACANES benchmarking. To produce experimental data for LACANES benchmarking, thermal–hydraulic tests were conducted by using a 12-m tall LBE integral test facility, named as Heavy Eutectic liquid metal loop for integral test of Operability and Safety of PEACER (HELIOS) which has been constructed in 2005 at the Seoul National University in the Republic of Korea. LACANES benchmark campaigns consist of a forced convection (phase-I) and a natural circulation (phase-II). In the forced convection case, the predictions of pressure losses based on handbook correlations and that obtained by Computational Fluid Dynamics code simulation were compared with the measured data for various components of the HELIOS test facility. Based on comparative analyses of the predictions and the measured data, recommendations for the prediction methods of a pressure loss in LACANES were obtained. In this paper, results for the forced convection case (phase-I) of LACANES benchmarking are described.

Neutron measurements for advanced nuclear systems: The n_TOF project at CERN

A few years ago, the neutron time-of-flight facility n_TOF was built at CERN to address some of the urgent needs of high-accuracy nuclear data for Accelerator Driven Systems and other advanced nuclear energy systems, as well as for nuclear astrophysics and fundamental nuclear physics. Thanks to the characteristics of the neutron beam, and to state-of-the-art detection and acquisition systems, high quality neutron cross-section data have been obtained for a variety of isotopes, many of which radioactive. Following an important upgrade of the spallation target and of the experimental area, a new measurement campaign has started last year. After a brief review of the most important results obtained so far at n_TOF, the new features of the facility are presented, together with the first results on the commissioning of the neutron beam. The plans for future measurements, in particular related to nuclear technology are finally discussed.

CFD ANALYSIS OF HEAVY LIQUID METAL FLOW IN THE CORE OF THE HELIOS LOOP

Lead-alloys are very attractive nuclear coolants due to their thermo-hydraulic, chemical, and neutronic properties. By utilizing the HELIOS (Heavy Eutectic liquid metal Loop for Integral test of Operability and Safety of PEACER² ) facility, a thermal hydraulic benchmarking study has been conducted for the prediction of pressure loss in lead-alloy cooled advanced nuclear energy systems (LACANES). The loop has several complex components that cannot be readily characterized with available pressure loss correlations. Among these components is the core, composed of a vessel, a barrel, heaters separated by complex spacers, and the plenum. Due to the complex shape of the core, its pressure loss is comparable to that of the rest of the loop. Detailed CFD simulations employing different CFD codes are used to determine the pressure loss, and it is found that the spacers contribute to nearly 90 percent of the total pressure loss. In the system codes, spacers are usually accounted for; however, due to the lack of correlations for the exact spacer geometry, the accuracy of models relies strongly on assumptions used for modeling spacers. CFD can be used to determine an appropriate correlation. However, application of CFD also requires careful choice of turbulence models and numerical meshes, which are selected based on extensive experience with liquid metal flow simulations for the KALLA lab. In this paper consistent results of CFX and Star-CD are obtained and compared to measured data. Measured data of the pressure loss of the core are obtained with a differential pressure transducer located between the core inlet and outlet at a flow rate of 13.57kg/s.

The application of a figure of merit for nuclear explosive utility as a metric for material attractiveness in a nuclear material theft scenario

Effective integration of nonproliferation management into the design process is key to the broad deployment of advanced nuclear energy systems, and is an explicit goal of the Laser Inertial Fusion Energy (LIFE) project at Lawrence Livermore National Laboratory. The nuclear explosives utility of a nuclear material to a state (proliferator) or sub-state (terrorist) is a critical factor to be assessed and is one aspect of material attractiveness. In this work, we approached nuclear explosives utility through the calculation of a “figure of merit” (FOM) that has recently been developed to capture the relative viability and difficulty of constructing nuclear explosives starting from various nuclear material forms and compositions. We discuss the integration of the figure of merit into an assessment of a nuclear material theft scenario and its use in the assessment. This paper demonstrates that material attractiveness is a multidimensional concept that embodies more than the FOM. It also seeks to propose that other attributes may be able to be quantified through analogous FOMs (e.g., transformation) and that, with quantification, aggregation may be possible using concepts from the risk community.

High-fidelity system modeling of advanced nuclear energy systems approaching a zero-nuclear-waste limit

This paper will present a high-fidelity system analysis of advanced nuclear energy systems approaching a zero-nuclear-waste limit. The focus will be on achievable performance characteristics. The applied high-fidelity system modeling approach will be discussed with illustrations using obtained physics simulation results. The near-term deployable dedicated small 14 MeV-neutron waste incineration back-end clean-up facilities are envisioned as advanced incinerators that are used in combination with thermal reactors (Westinghouse AP1000 and Generation IV VHTR are considered in the present study) to minimize the remaining long-lived radioactive species. These 14 MeV transmuters are viewed as small facilities designed to reduce waste inventories prior to reprocessing stages minimizing spent fuel activity levels and repository requirements. It is envisioned that under sustainable nuclear energy scenario small clean-up facilities will eventually evolve into self-consistent complexes producing electricity and serving as energy sources for industrial applications. The paper discusses physics features of these advanced systems and challenges of their conclusive evaluations.

Uranium Pyrophosphate/Methylenediphosphonate Polyoxometalate Cage Clusters

Despite potential applications in advanced nuclear energy systems, nanoscale control of uranium materials is in its infancy. In its hexavalent state, U occurs as (UO(2))(2+) uranyl ions that are coordinated by various ligands to give square, pentagonal, or hexagonal bipyramids. Creation and design of nanostructured uranyl materials requires interruption of the tendency of uranyl bipyramids to share equatorial edges to form infinite sheets that occur in extended structures. Where a bidentate peroxide group bridges uranyl bipyramids, the configuration is inherently bent, fostering formation of cage clusters. Here the bent configurations of four- and five-membered rings of uranyl peroxide hexagonal bipyramids are bridged by pyrophosphate or methylenediphosphonate, creating eight chemically complex cage clusters with specific topologies. Chemical complexity in such clusters provides opportunities for the tuning of cage sizes, pore sizes, and properties such as aqueous solubility. Several of these are topological derivatives of simpler clusters that contain only uranyl bipyramids, whereas others exhibit new topologies.

Study on the Coupled Neutronic and Thermal-Hydraulic Characteristics of the New Concept Molten Salt Reactor

The new concept molten salt reactor is the only liquid-fuel reactor of the six Generation IV advanced nuclear energy systems. The liquid molten salt serves as the fuel and coolant simultaneously and causes one important feature: the delayed neutron precursors are drifted by the fuel flow, which leads the spread of delayed neutrons’ distribution to noncore parts of the primary circuit, and it also results in reactivity variation depending on the flow condition of the fuel salt. Therefore, the neutronic and thermal-hydraulic characteristics of the molten salt reactor are quite different from the conventional nuclear reactors using solid fissile materials. Besides, there is no other reactor design theory and safety analysis methodologies can be used for reference. The neutronic model is derived based on the conservation of particles considering the flow effect of the fuel salt in the molten salt reactor, while the thermal-hydraulic model applies the fundamental conservation laws: the mass, momentum, and energy conservation equations. Then, the neutronic and thermal-hydraulic calculations are coupled and the influences of inflow temperature and flow velocity on the reactor physical properties are obtained. The calculated results show that the flow effect on the distributions of thermal and fast neutron fluxes is very weak, as well as on the effective multiplication factor keff, while the flow effect on the distribution of delayed neutron precursors is much stronger. The inflow temperature influences the distribution of neutron fluxes and delayed neutron precursors slightly, and makes a significant negative reactivity. Coupled calculation also reveals that the flow velocity of molten salt has little effect on the distribution of neutron fluxes in the steady-state, but affects the delayed neutron precursors’ distribution significantly.

A Once-Through Fuel Cycle for Fast Reactors

A paradigm shift has altered the design targets for advanced nuclear energy systems that use a fast neutron spectrum. Whereas designers previously emphasized the ability of fast reactors to extend global reserves of fissile fuels, the overriding desire now is for reactor technologies that are “cleaner, more efficient, less waste-intensive, and more proliferation-resistant.” (Cheney, 2001, “U.S. National Energy Policy,” National Energy Policy Development Group, Washington, DC) This shift in priorities, along with recent design advances that enable high fuel burnup even when using fuels that have been minimally enriched, creates an opportunity to use fast reactors in an open nuclear fuel cycle. One promising route to this goal exploits a phenomenon known as a traveling wave, which can attain high burnups without reprocessing. A traveling-wave reactor (TWR) breeds and uses its own fuel in place as it operates. Recent design work has demonstrated that TWRs could be fueled almost entirely by depleted or natural uranium, thus reducing the need for initial enrichment. The calculations described here show that a gigawatt-scale electric TWR can achieve a burnup of 20%, which is four to five times that realized in current light water reactors. Burnups as high as 50% appear feasible. The factors that contribute to these high burnups and the implications for materials design are discussed.

Advanced nuclear energy systems and the need of accurate nuclear data: the n_TOF project at CERN

To satisfy the world's constantly increasing demand for energy, a suitable mix of different energy sources has to be devised. In this scenario, an important role could be played by nuclear energy, provided that major safety, waste and proliferation issues affecting current nuclear reactors are satisfactorily addressed. To this purpose, a large effort has been under way for a few years towards the development of advanced nuclear systems with the aim of closing the fuel cycle. Generation IV reactors, with full or partial waste recycling capability, accelerator driven systems, as well as new fuel cycles are the main options being investigated. The design of advanced systems requires improvements in basic nuclear data, such as cross-sections for neutron-induced reactions on actinides. In this paper, the main concepts of advanced reactor systems are described, together with the related needs of new and accurate nuclear data. The present activity in this field at the neutron facility n_TOF at CERN is discussed.

Fuels for Advanced Nuclear Energy Systems

AbstractFuels for advanced nuclear reactors differ from conventional light water reactor fuels and also vary widely because of the specific architectures and intended missions of the reactor systems proposed to deploy them. Functional requirements of all fuel designs for advanced nuclear energy systems include (1) retention of fission products and fuel nuclides, (2) dimensional stability, and (3) maintenance of a geometry that can be cooled. In all cases, anticipated fuel performance is the limiting factor in reactor system design, and cumulative effects of increased utilization and increased exposure to inservice environments degrade fuel performance. In this article, the current status of each fuel system is reviewed, and technical challenges confronting the implementation of each fuel in the context of the entire advanced reactor fuel cycle (fabrication, reactor performance, recycle) are discussed.

ENDOTHELIALISATION AND CELL RETENTION ON BIOLOGICAL AND SYNTHETIC MATERIALS

Designing radiation tolerant materials is one of the primary challenges associated with advanced nuclear energy systems. One attractive route that has received much attention world-wide is to introduce a high density of sinks, often in the form of interfaces or secondary phases. Here, we develop a simple model of such nanocomposites and examine the ramifications of various factors on the overall radiation stability of the material. In particular, we determine how the distribution of secondary phases, the relative sink strength of those phases, and the irradiation temperature influence the radiation tolerance of the matrix. We find that the best scenario is one in which the sinks have intermediate strength, transiently trapping defects before releasing them back into the matrix. Neither perfect sinks nor the complete absence of sinks perform as well. This provides new insight into the optimal properties of nanocomposites for radiation damage environments.

Fusion–fission hybrids for nuclear waste transmutation: A synergistic step between Gen-IV fission and fusion reactors

Energy demand and GDP per capita are strongly correlated, while public concern over the role of energy in climate change is growing. Nuclear power plants produce 16% of world electricity demands without greenhouse gases. Generation-IV advanced nuclear energy systems are being designed to be safe and economical. Minimizing the handling and storage of nuclear waste is important. NIF and ITER are bringing sustainable fusion energy closer, but a significant gap in fusion technology development remains. Fusion–fission hybrids could be a synergistic step to a pure fusion economy and act as a technology bridge. We discuss how a pulsed power-driven Z-pinch hybrid system producing only 20 MW of fusion yield can drive a sub-critical transuranic blanket that transmutes 1280 kg of actinide wastes per year and produces 3000 MW. These results are applicable to other inertial and magnetic fusion energy systems. A hybrid system could be introduced somewhat sooner because of the modest fusion yield requirements and can provide both a safe alternative to fast reactors for nuclear waste transmutation and a maturation path for fusion technology. The development and demonstration of advanced materials that withstand high-temperature, high-irradiation environments is a fundamental technology issue that is common to both fusion–fission hybrids and Generation-IV reactors.

SSTAR: The US lead-cooled fast reactor (LFR)

It is widely recognized that the developing world is the next area for major energy demand growth, including demand for new and advanced nuclear energy systems. With limited existing industrial and grid infrastructures, there will be an important need for future nuclear energy systems that can provide small or moderate increments of electric power (10–700 MWe) on small or immature grids in developing nations. Most recently, the global nuclear energy partnership (GNEP) has identified, as one of its key objectives, the development and demonstration of concepts for small and medium-sized reactors (SMRs) that can be globally deployed while assuring a high level of proliferation resistance. Lead-cooled systems offer several key advantages in meeting these goals. The small lead-cooled fast reactor concept known as the small secure transportable autonomous reactor (SSTAR) has been under ongoing development as part of the US advanced nuclear energy systems programs. It is a system designed to provide energy security to developing nations while incorporating features to achieve nonproliferation goals, anticipating GNEP objectives. This paper presents the motivation for development of internationally deployable nuclear energy systems as well as a summary of one such system, SSTAR, which is the US Generation IV lead-cooled fast reactor system.

Fabrication of Cermet Nuclear Fuels Designed for the Transmutation of Transuranic Isotopes

The Uranium Extraction (UREX) family of processes uses solvent extraction techniques designed to partition spent uranium and transuranic (TRU) isotopes from fission product waste. Once separated, the collective TRU elements (Np, Pu, Am, and Cm) can be recycled in advanced nuclear energy systems. A zirconium matrix cermet is proposed as a fuel form for this application. Processing methods have been designed to convert the TRU product and spent Zircaloy cladding into feed materials for the hot extrusion of the cermet fuel pins. The TRU conversion process is being developed using a surrogate mixture of uranium and cerium nitrate solutions to generate mixed oxide microspheres. The Zircaloy recovery process is a hydride-dehydride method that is being demonstrated at the bench scale. The powder products from these methods may be combined through hot extrusion into a cermet composite; demonstration experiments using zirconium powder and zirconia microspheres have been completed.