Advanced Fuel Cycle Initiative Research Articles

Technological Assessment of National Nuclear Fuel Cycle Using Transmutations of High-Level Nuclear Waste (Hlw)

The advanced fuel cycle initiative (AFCI) has been investigated for the safe processing of the spent nuclear fuels (SNFs), which has focused mainly on the economic factor. The simulation of the political factor is suggested, which is introduced by the political susceptibility factor (PSF), because the political situation is much more important in the treatment of the SNFs considering the characteristics of the nuclear material. The system dynamics (SD) algorithm is used in the dynamical simulation where the political aspect is emphasized. There are 5 classifications of the PSF as president, party, vote, term, and feedback. The degree of the possibility is impacted by 5 steps. Eventually, the dynamic simulation is quantified for the SNFs. The possibility of the high-level nuclear waste (HLW) repository construction increases slowly in the early stage and faster in the later stage. The importance of the political aspect for the SNFs treatment is shown as numerical values with easy estimations.

Neutron Cross Section Covariances for Structural Materials and Fission Products

We describe neutron cross section covariances for 78 structural materials and fission products produced for the new US evaluated nuclear reaction library ENDF/B-VII.1. Neutron incident energies cover full range from 10 − 5 eV to 20 MeV and covariances are primarily provided for capture, elastic and inelastic scattering as well as (n,2n). The list of materials follows priorities defined by the Advanced Fuel Cycle Initiative, the major application being data adjustment for advanced fast reactor systems. Thus, in addition to 28 structural materials and 49 fission products, the list includes also 23Na which is important fast reactor coolant. Due to extensive amount of materials, we adopted a variety of methodologies depending on the priority of a specific material. In the resolved resonance region we primarily used resonance parameter uncertainties given in Atlas of Neutron Resonances and either applied the kernel approximation to propagate these uncertainties into cross section uncertainties or resorted to simplified estimates based on integral quantities. For several priority materials we adopted MF32 covariances produced by SAMMY at ORNL, modified by us by adding MF33 covariances to account for systematic uncertainties. In the fast neutron region we resorted to three methods. The most sophisticated was EMPIRE-KALMAN method which combines experimental data from EXFOR library with nuclear reaction modeling and least-squares fitting. The two other methods used simplified estimates, either based on the propagation of nuclear reaction model parameter uncertainties or on a dispersion analysis of central cross section values in recent evaluated data files. All covariances were subject to quality assurance procedures adopted recently by CSEWG. In addition, tools were developed to allow inspection of processed covariances and computed integral quantities, and for comparing these values to data from the Atlas and the astrophysics database KADoNiS.

Dynamic Simulations of Advanced Fuel Cycles

Nothing in life is static, so why compare fuel cycle options using only static, equilibrium analyses? Competitive industry looks at how new technology options might displace existing technologies and change how existing systems work. So too, our years of performing dynamic simulations of advanced nuclear fuel cycle options provide insights into how they might work and how one might transition from the current once-through fuel cycle. This paper summarizes those insights within the context of the 2005 objectives and goals of what was then the U.S. Advanced Fuel Cycle Initiative (AFCI). The intent here is not to compare options, assess options versus those objectives and goals, nor recommend changes to those objectives and goals. (The specific options change over time; the objective in this paper is to look for more generic insights.) We organize what we have learned from dynamic simulations in the context of the AFCI objectives for waste management, proliferation resistance, uranium utilization, and economics. Thus, we do not merely describe “lessons learned” from dynamic simulations but attempt to answer the “so what” question by using this context; i.e., how do the lessons learned matter relative to goals and objectives not just to technological observations? The analyses have been performed using the Verifiable Fuel Cycle Simulation of Nuclear Fuel Cycle Dynamics (VISION). We observe that the 2005 objectives and goals do not address many of the inherently dynamic discriminators among advanced fuel cycle options and transitions thereof.

REACTOR-BASED MANAGEMENT OF USED NUCLEAR FUEL: ASSESSMENT OF MAJOR OPTIONS

This paper discusses the current status of the ongoing Advanced Fuel Cycle Initiative (AFCI) program in the U.S. Department of Energy that is investigating the potential for using the processing and recycling of used nuclear fuel to improve radioactive waste management, including used fuel. A key element of the strategies is to use nuclear reactors for further irradiation of recovered chemical elements to transmute certain long-lived highly-radioactive isotopes into less hazardous isotopes. Both thermal and fast neutron spectrum reactors are being studied as part of integrated nuclear energy systems where separations, transmutation, and disposal are considered. Radiotoxicity is being used as one of the metrics for estimating the hazard of used fuel and the processing of wastes resulting from separations and recycle-fuel fabrication. Decay heat from the used fuel and/or wastes destined for disposal is used as a metric for use of a geologic repository. Results to date indicate that the most promising options appear to be those using fast reactors in a repeated recycle mode to limit buildup of higher actinides, since the transuranic elements are a key contributor to the radiotoxicity and decay heat. Using such an approach, there could be much lower environmental impact from the high-level waste as compared to direct disposal of the used fuel, but there would likely be greater generation of low-level wastes that will also require disposal. An additional potential waste management benefit is having the ability to tailor waste forms and contents to one or more targeted disposal environments (i.e., to be able to put waste in environments best-suited for the waste contents and forms).

Verifiable Fuel Cycle Simulation Model (VISION): A Tool for Analyzing Nuclear Fuel Cycle Futures

The nuclear fuel cycle consists of a set of complex components that are intended to work together. To support the nuclear renaissance, it is necessary to understand the impacts of changes and timing of events in any part of the fuel cycle system such as how the system would respond to each technological change, a series of which moves the fuel cycle from where it is to a postulated future state. The system analysis working group of the United States research program on advanced fuel cycles (formerly called the Advanced Fuel Cycle Initiative) is developing a dynamic simulation model, VISION, to capture the relationships, timing, and changes in and among the fuel cycle components to help develop an understanding of how the overall fuel cycle works. This paper is an overview of the philosophy and development strategy behind VISION. The paper includes some descriptions of the model components and some examples of how to use VISION. For example, VISION users can now change yearly the selection of separation or reactor technologies, the performance characteristics of those technologies, and/or the routing of material among separation and reactor types—with the model still operating on a PC in <5 min.

Thermal analysis of proposed transport cask for three advanced burner reactor used fuel assemblies

Preliminary studies of used fuel generated in the US Department of Energy's Advanced Fuel Cycle Initiative have indicated that current used fuel transport casks may be insufficient for the transportation of said fuel. This work considers transport of three 5-year-cooled oxide advanced burner reactor used fuel assemblies with a burn-up of 160 MWD kg–1. A transport cask designed to carry these assemblies is proposed. This design employs a 7-cm-thick lead gamma shield and a 20-cm-thick NS-4-FR composite neutron shield. The temperature profile within the cask, from its centre to its exterior surface, is determined by two-dimensional computational fluid dynamics simulations of conduction, convection and radiation within the cask. Simulations are performed for a cask with a smooth external surface and various neutron shield thicknesses. Separate simulations are performed for a cask with a corrugated external surface and a neutron shield thickness that satisfies shielding constraints. Resulting temperature profiles indicate that a three-assembly cask with a smooth external surface will meet fuel cladding temperature requirements but will cause outer surface temperatures to exceed the regulatory limit. A cask with a corrugated external surface will not exceed the limits for both the fuel cladding and outer surface temperatures.

Shield Design of the Materials Test Station’s Camera Room

The Materials Test Station (MTS) is a project funded by the Advanced Fuel Cycle Initiative with the goal to build a facility that allows large-scale irradiation for potential future nuclear fuel and material samples to obtain the knowledge and understanding of the nuclear processes necessary to close the nuclear fuel cycle and thereby reduce the amount and the toxicity of the nuclear waste. The MTS is proposed to be built in Area A of the Los Alamos Neutron Science Center and operated at up to 2 MW (2.5 mA at 800 MeV). As part of this operation, a so-called camera room will need to be installed upstream of the target cell. Because of the uniqueness of this functionality, the camera room requires a special shielding design, which will be discussed in this paper.

Design of the Shielding of the Materials Test Station

The Materials Test Station (MTS) is a project by the Advanced Fuel Cycle Initiative to build a facility that allows for irradiating nuclear fuel and material samples to acquire the necessary knowledge to close the nuclear fuel cycle and thereby reduce the amount and the toxicity of the nuclear waste. This facility is proposed to be located in Area A of the Los Alamos Neutron Science Center at the Los Alamos National Laboratory. The MTS is proposed to be a spallation target facility operated up to 2 MW (2.5 mA at 800 MeV). To safely operate a facility of this size, a large amount of shielding needs to be put into place. In this paper we will discuss the shielding design proposed for the MTS.

Phase characteristics of a number of U–Pu–Am–Np–Zr metallic alloys for use as fast reactor fuels

Metallic fuel alloys consisting of uranium, plutonium, and zirconium with minor additions of americium and neptunium are under evaluation for potential use to transmute long-lived transuranic actinide isotopes in fast reactors. A series of test designs for the Advanced Fuel Cycle Initiative (AFCI) have been irradiated in the Advanced Test Reactor (ATR), designated as the AFC-1 and AFC-2 designs. Metal fuel compositions in these designs have included varying amounts of U, Pu, Zr, and minor actinides (Am, Np). Investigations into the phase behavior and relationships based on the alloy constituents have been conducted using X-ray diffraction and differential thermal analysis. Results of these investigations, along with proposed relationships between observed behavior and alloy composition, are provided. In general, observed behaviors can be predicted by a ternary U–Pu–Zr phase diagram, with transition temperatures being most dependent on U content. Furthermore, the enthalpy associated with transitions is strongly dependent on the as-cast microstructural characteristics.

Actinide targets for neutron cross section measurements

The Advanced Fuel Cycle Initiative (AFCI) and the Generation IV Reactor Initiative have demonstrated a lack of detailed neutron cross-sections for certain minor actinides, those other than the most common (235U, 238U, and 239Pu). For some closed-fuel-cycle reactor designs more than 50% of reactivity will, at some point, be derived from minor actinides that currently have poorly known or in some cases not measured (n,?) and (n,f) cross sections. A program of measurements under AFCI has begun to correct this. One of the initial hurdles has been to produce well-characterized, highly isotopically enriched, and chemically pure actinide targets on thin backings. Using a combination of resurrected techniques and new developments, we have made a series of targets including highly enriched 239Pu, 240Pu, and 242Pu. Thus far, we have electrodeposited these actinide targets. In the future, we plan to study reductive distillation to achieve homogeneous, adherent targets on thin metal foils and polymer backings. As we move forward, separated isotopes become scarcer, and safety concerns become greater. The chemical purification and electodeposition techniques will be described.

Engineering-Scale Liquid Cadmium Cathode Experiments

Recovery of uranium and transuranic (TRU) actinides from spent nuclear fuel by an electrorefining process was investigated as part of the U.S. Department of Energy Advanced Fuel Cycle Initiative. Experiments were performed in a shielded hot cell at the Materials and Fuels Complex at Idaho National Laboratory. The goal of these experiments was to collect, by an electrochemical process, kilogram quantities of uranium and plutonium into what is called a liquid cadmium cathode (LCC).For each experiment, a steel basket loaded with chopped spent nuclear fuel from the Experimental Breeder Reactor II acted as the anode in the electrorefiner. The cathode was a beryllium oxide crucible containing ~26 kg of cadmium metal (the LCC). In the three experiments performed to date, between 1 and 2 kg of heavy metal was collected in the LCC after passing an integrated current between 1.80 and 2.16 MC (500 and 600 A h) from the anode to the cathode. Sample analysis of the processed LCC ingots measured detectable amounts of TRUs and rare earth elements.

Status of materials handbooks for particle accelerator and nuclear reactor applications

In support of research and development for accelerator applications, a materials handbook was developed in August of 1998 funded by the Accelerator Production of Tritium Project. This handbook, presently called Advanced Fuel Cycle Initiative ( AFCI) Materials Handbook, Materials Data for Particle Accelerator Applications, has just issued Revision 5 and contains detailed information showing the effects of irradiation on many properties for a wide variety of materials. Development of a web-accessible materials database for Generation IV Reactor Programs has been ongoing for about three years. This handbook provides a single authoritative source for qualified materials data applicable to all Generation IV reactor concepts. A beta version of this Gen IV Materials Handbook has been completed and is presently under evaluation.

Engineering-Scale Liquid Cadmium Cathode Experiments

Recovery of transuranic actinides (TRU) using electrorefining is a process being investigated as part of the Department of Energy (DOE) Advanced Fuel Cycle Initiative (AFCI). TRU recovery via electrorefining onto a solid cathode is very difficult as the thermodynamic properties of transuranics are not favourable for them to remain in the metal phase while significant quantities of uranium trichloride exist in the electrolyte. Theoretically, the concentration of transuranics in the electrolyte must be approximately 106 greater than the uranium concentration in the electrolyte to produce a transuranic deposit on a solid cathode. Using liquid cadmium as a cathode contained within a LiCl-KCl eutectic salt, the co-deposition of uranium and transuranics is feasible because the activity of the transuranics in liquid cadmium is very small. Depositing transuranics and uranium in a liquid cadmium cathode (LCC) theoretically requires the concentration of transuranics to be two to three times the uranium concentration in the electrolyte. Three LCC experiments were performed in an Engineering scale elecdtrorefiner, which is located in the argon hot cell of the Fuel Conditioning Facility at the Materials and Fuels Complex on the Idaho National Laboratory. Figure 1 contains photographs of the LCC assembly in the hot cell prior to the experiment and a cadmium ingot produced after the first LCC test. Figure 1. Liquid Cadmium Cathode (left) and Cadmium Ingot (right) The primary goal of the engineering-scale liquid cadmium cathode experiments was to electrochemically collect kilogram quantities of uranium and plutonium via a LCC. The secondary goal was to examine fission product contaminations in the materials collected by the LCC. Each LCC experiment used chopped spent nuclear fuel from the blanket region of the Experimental Breeder Reactor II loaded into steel baskets as the anode with the LCC containing 26 kg of cadmium metal. In each experiment, between one and two kilograms of heavy metal was collected in the LCC after passing an integrated current over 500 amp hours. Analysis of samples from the liquid cadmium cathode ingots showed detectable amounts of transuranics and rare-earth elements. Acknowledgements K. B. Davies and D. M. Pace for the mechanical and electrical engineering needed to prepare the equipment for the engineering-scale liquid cadmium cathode experiments.

Impact on Geologic Repository Usage from Limited Actinide Recycle in Pressurized Light Water Reactors

A project has been conducted as part of the U.S. Department of Energy Advanced Fuel Cycle Initiative to evaluate the impact of limited actinide recycling in light water reactors on the utilization of a geologic repository where loading of the repository is constrained by the decay heat of the emplaced materials. In this study, it was assumed that spent PWR fuel was processed, removing the uranium, plutonium, americium, and neptunium, along with the fission products cesium and strontium. Previous work had demonstrated that these elements were responsible for limiting loading in the repository based on thermal constraints. The plutonium, americium, and neptunium were recycled in a PWR, with process waste and spent recycled fuel being sent to the repository. The cesium and strontium were placed in separate storage for 100–300 years to allow for decay prior to disposal. The study examined the effect of single and mutliple recycles of the recovered plutonium, americium, and neptunium, as well as different processing delay times. The potential benefit to the repository was measured by the increase in utilization of repository space as indicated by the allowable linear loading in the repository drifts (tunnels). The results showed that limited recycling would provide only a small fraction of the benefit that could be achieved with repeated processing and recycling, as is possible in fast neutron reactors.

Impact on Geologic Repository Usage from Limited Actinide Recycle in Pressurized Light Water Reactors

A project has been conducted as part of the U.S. Department of Energy Advanced Fuel Cycle Initiative to evaluate the impact of limited actinide recycling in light water reactors on the utilization of a geologic repository where loading of the repository is constrained by the decay heat of the emplaced materials. In this study, it was assumed that spent PWR fuel was processed, removing the uranium, plutonium, americium, and neptunium, along with the fission products cesium and strontium. Previous work had demonstrated that these elements were responsible for limiting loading in the repository based on thermal constraints. The plutonium, americium, and neptunium were recycled in a PWR, with process waste and spent recycled fuel being sent to the repository. The cesium and strontium were placed in separate storage for 100–300 years to allow for decay prior to disposal. The study examined the effect of single and mutliple recycles of the recovered plutonium, americium, and neptunium, as well as different processing delay times. The potential benefit to the repository was measured by the increase in utilization of repository space as indicated by the allowable linear loading in the repository drifts (tunnels). The results showed that limited recycling would provide only a small fraction of the benefit that could be achieved with repeated processing and recycling, as is possible in fast neutron reactors.

A Subcritical, Helium-Cooled Fast Reactor for the Transmutation of Spent Nuclear Fuel

A design concept and supporting analysis are presented for a He-cooled fast reactor for the transmutation of spent nuclear fuel. Coated transuranic (TRU) fuel particles in a SiC matrix are used. The reactor operates subcritical (k ≤ 0.95), with a tokamak D-T fusion neutron source, to achieve >90% TRU burnup in repeated five-batch fuel cycles, fissions 1.1 tonnes/full-power year, and produces 700 MW(electric) net electrical power. The reactor design is based on nuclear, fuels, materials, and separations technologies being developed in the Generation-IV, Next Generation Nuclear Plant, and Advanced Fuel Cycle Initiative programs and similar international programs, and the fusion neutron source is based on the physics and technology supporting the ITER design.

Tensile testing of EP-823 and HT-9 after irradiation in STIP II

The predicted operating conditions for a lead–bismuth eutectic target to be used in an accelerator driven system for the Advanced Fuel Cycle Initiative spans a temperature range of 300–600 °C while being irradiated by a high energy (∼600 MeV) proton beam. Such spallation conditions lead to high displacement rates coupled with high accumulation rates of helium and hydrogen up to 150 appm/dpa. Two candidate ferritic/martensitic materials for these applications include HT-9 and EP-823. To investigate the effect of irradiation on these materials, the tensile properties were measured at 25, 250, and 400 °C after irradiation in the STIP II irradiation to doses up to 20 dpa at temperatures up to 350 °C. Losses in ductility concomitant with increases in yield stress are observed in both alloys although the embrittlement is more severe in the EP-823. This stronger embrittlement is attributed to the high silicon content in EP-823.

Separations and Transmutation Criteria to Improve Utilization of a Geologic Repository

This paper describes the results of a study that uses the thermal performance of the repository to establish chemical separations and transmutation criteria for commercial spent nuclear fuel of benefit to a geologic repository, as measured by the allowable increase in utilization of repository space. The method for determining the chemical elements to be separated is based on the thermal performance of the repository. The important chemical elements are identified, the order of importance of the separated elements is established, and the relationship between the efficiency of the chemical separation and the resulting increase in utilization is determined. The proposed repository at Yucca Mountain is used as an example of a geologic repository for the purposes of illustrating the magnitude of the benefits that are possible and the implications for repository size and operation. This work is being done in support of the U.S. Department of Energy Advanced Fuel Cycle Initiative, where numerous reactor, processing, and recycling strategies are being examined to determine the impact on issues important to the viability of nuclear electricity generation, including the disposal of spent nuclear fuel and nuclear waste.

The McGNASH nuclear reaction code and its use for gas production cross-section calculations

The McGNASH nuclear reaction code being developed at Los Alamos is presented. It is a modern version of the well-known GNASH code, which implements the Hauser–Feshbach theory of the compound nucleus evaporation and takes into account corrections from the pure statistical formalism (direct reactions, pre-equilibrium processes, etc.). Its use spans neutron-, proton- and photon-induced reactions on target nuclei starting from A ≃ 20 and up, and for incident energies from tens of keV up to about 200 MeV. Its application to the determination of gas production cross-sections for the Advanced Fuel Cycle Initiative is presented.

Engineered product storage under the advanced fuel cycle initiative. Part II: Conceptual storage scenarios

The Advanced Fuel Cycle Initiative (AFCI) of the US Department of Energy stresses waste minimization within an integrated nuclear fuel cycle. In one possible future fuel cycle scenario, the radionuclides 244Cm and 241Am are placed in decay storage until a capability for fast-spectrum transmutation is available. We discuss the scale of hypothetical wet and dry storage facilities for cesium/strontium and americium/curium products. Because of the extended storage period for cesium/strontium products, underground storage appears to be the option most acceptable to regulatory policy. Several underground designs are discussed. To reduce the required underground storage space, we recommend a pre-staging period where the cesium/strontium products are cooled in a small pool for 30–50 years. Americium and curium products would require storage for not more than approximately 50 years. For this reason, and because of the high thermal power density, pool storage appears to be the best option for americium and curium.