James A. Blink

Generic repository design concepts and thermal analysis (FY11).

Reference concepts for geologic disposal of used nuclear fuel and high-level radioactive waste in the U.S. are developed, including geologic settings and engineered barriers. Repository thermal analysis is demonstrated for a range of waste types from projected future, advanced nuclear fuel cycles. The results show significant differences among geologic media considered (clay/shale, crystalline rock, salt), and also that waste package size and waste loading must be limited to meet targeted maximum temperature values. In this study, the UFD RD (2) waste generated from reprocessing of LWR UOX UNF to recover U and Pu, and subsequent direct disposal of used Pu-MOX fuel (also used in LWRs) in a modified-open cycle; and (3) waste generated by continuous recycling of metal fuel from fast reactors operating in a TRU burner configuration, with additional TRU material input supplied from reprocessing of LWR UOX fuel. The geologic setting provides the natural barriers, and establishes the boundary conditions for performance of engineered barriers. The composition and physical properties of the host medium dictate design and construction approaches, and determine hydrologic and thermal responses of the disposal system. Clay/shale, salt, and crystalline rock media are selected as the basis for reference mined geologic disposal concepts in this study, consistent with advanced international repository programs, and previous investigations in the U.S. The U.S. pursued deep geologic disposal programs in crystalline rock, shale, salt, and volcanic rock in the years leading up to the Nuclear Waste Policy Act, or NWPA (Rechard et al. 2011). The 1987 NWPA amendment act focused the U.S. program on unsaturated, volcanic rock at the Yucca Mountain site, culminating in the 2008 license application. Additional work on unsaturated, crystalline rock settings (e.g., volcanic tuff) is not required to support this generic study. Reference disposal concepts are selected for the media listed above and for deep borehole disposal, drawing from recent work in the U.S. and internationally. The main features of the repository concepts are discussed in Section 4.5 and summarized in Table ES-1. Temperature histories at the waste package surface and a specified distance into the host rock are calculated for combinations of waste types and reference disposal concepts, specifying waste package emplacement modes. Target maximum waste package surface temperatures are identified, enabling a sensitivity study to inform the tradeoff between the quantity of waste per disposal package, and decay storage duration, with respect to peak temperature at the waste package surface. For surface storage duration on the order of 100 years or less, waste package sizes for direct disposal of SNF are effectively limited to 4-PWR configurations (or equivalent size and output). Thermal results are summarized, along with recommendations for follow-on work including adding additional reference concepts, verification and uncertainty analysis for thermal calculations, developing descriptions of surface facilities and other system details, and cost estimation to support system-level evaluations.

Disposal Systems Evaluations and Tool Development - Engineered Barrier System (EBS) Evaluation.

Key components of the nuclear fuel cycle are short-term storage and long-term disposal of nuclear waste. The latter encompasses the immobilization of used nuclear fuel (UNF) and radioactive waste streams generated by various phases of the nuclear fuel cycle, and the safe and permanent disposition of these waste forms in geological repository environments. The engineered barrier system (EBS) plays a very important role in the long-term isolation of nuclear waste in geological repository environments. EBS concepts and their interactions with the natural barrier are inherently important to the long-term performance assessment of the safety case where nuclear waste disposition needs to be evaluated for time periods of up to one million years. Making the safety case needed in the decision-making process for the recommendation and the eventual embracement of a disposal system concept requires a multi-faceted integration of knowledge and evidence-gathering to demonstrate the required confidence level in a deep geological disposal site and to evaluate long-term repository performance. The focus of this report is the following: (1) Evaluation of EBS in long-term disposal systems in deep geologic environments with emphasis on the multi-barrier concept; (2) Evaluation of key parameters in the characterization of EBS performance; (3) Identification of key knowledge gaps and uncertainties; and (4) Evaluation of tools and modeling approaches for EBS processes and performance. The above topics will be evaluated through the analysis of the following: (1) Overview of EBS concepts for various NW disposal systems; (2) Natural and man-made analogs, room chemistry, hydrochemistry of deep subsurface environments, and EBS material stability in near-field environments; (3) Reactive Transport and Coupled Thermal-Hydrological-Mechanical-Chemical (THMC) processes in EBS; and (4) Thermal analysis toolkit, metallic barrier degradation mode survey, and development of a Disposal Systems Evaluation Framework (DSEF). This report will focus on the multi-barrier concept of EBS and variants of this type which in essence is the most adopted concept by various repository programs. Empasis is given mainly to the evaluation of EBS materials and processes through the analysis of published studies in the scientific literature of past and existing repository research programs. Tool evaluations are also emphasized, particularly on THCM processes and chemical equilibria. Although being an increasingly important aspect of NW disposition, short-term or interim storage of NW will be briefly discussed but not to the extent of the EBS issues relevant to disposal systems in deep geologic environments. Interim storage will be discussed in the report Evaluation of Storage Concepts FY10 Final Report (Weiner et al. 2010).

Fragmentation of Suddenly Heated Liquids in ICF Reactors

Fragmentation of free liquids in Inertial Confinement Fusion reactors could determine the upper bound on reactor pulse rate because increased surface area will enhance the cooling and condensation of coolant ablated by the fusion x rays. Relaxation from the suddenly (neutron) heated state will move a liquid into the negative pressure region under the liquid-vapor P-V dome. The resulting expansion in a diverging geometry will hydrodynamically force the liquid to fragment, with vapor then forming from the new surfaces to fill the cavities. An energy minimization model is used to determine the fragment size that produces the least amount of non-fragment-center-of-mass energy; i.e., the sum of the surface and dilational kinetic energies. This model predicts fragmentation dependence on original system size and amount of isochoric heating as well as liquid density, Grueneisen parameter, surface tension, and sound speed. A two dimensional molecular dynamics code was developed to test the model at a microscopic scale for the Lennard-Jones fluid with its two adjustable constants chosen to represent lithium.

Performance of a 500-W Nd:GGG zigzag slab oscillator

Realization of practical multi-kilowatt Nd:garnet lasers will require the scale-up of crystal dimensions as well as more powerful pumping sources. A high average power zigzag slab crystal amplifier testing facility has been established at LLNL which employs two 100 kWe vortex stabilized arc lamps, cooled reflectors and a cooled, spectrally filtered, crystal slab mounting fixture. The operational characteristics of the first crystal laser to be tested in this setup, a Nd:GGG zigzag oscillator, are presented. A Nd:GGG crystal of dimensions 18x7x0.5 cm3, doped at 2x1020 cm-3 Nd3+ atomic density, was pumped by up to 40 kW of filtered argon line emission. A small-signal single pass gain (losses excluded) of 1.09 was measured with a probe laser when the DC input to the lamps was 43 kWe. Our power supply was then modified to operate in a pulsed mode and provided one to three milliseconds pulses at 120 Hz. An average optical output power of 490 watts was obtained at a lamp input power of 93 kWe in an unoptimized resonator. The laser output power declined after a few tens of seconds since the slab tips were not properly cooled. A birdhouse specular lamp reflector and a contoured diffuse reflector were tested; in both cases the pump illuminated crystal surface was smaller than the total crystal face area. Fluorescence imaging of the zigzag amplifiers output aperture registered a smoother, more uniform pumping profile when the diffuse reflector was used. Uniformity of pumping results in decreased resonator loss and yields higher laser output power. Thermo-optic distortions observed in these preliminary tests are analyzed with the aid of computer simulations of the thermal fields, stresses, and surface displacements of our crystal slab.

Inertial confinement fusion development options: Facility characteristics and schedules from a reactor physics viewpoint

The pulsed, localized fusion source in inertial confinement fusion (ICF) permits scale-down of reactor dimensions and fusion yield in development facilities while still maintaining full-scale reactor surface and volume energy loads. Hence, the power and geometric scale of ICF development facilities can be much smaller than comparable magnetic fusion facilities. The power is reduced by reducing both the pulse rate and the target gain; however, full gain and pulse rate experiments of limited duration will be possible. At least three engineering facilities will be required for the development of heavy-ion beam or short wavelength laser driven fusion power. The design and construction times required for large facilities produce a nominal plan with a demonstration (DEMO) plant operating around the year 2018, and a crash plan with DEMO operation in 2009. Fusion breeder development is expected to follow a similar time line, except that a crash (option-limited) plan could succeed as early as the turn of the century.

End of FY10 Report - Used Fuel Disposition Technical Bases and Lessons Learned Legal and Regulatory Framework for High-Level Waste Disposition in the United States

This report examines the current policy, legal, and regulatory framework pertaining to used nuclear fuel and high level waste management in the United States. The goal is to identify potential changes that if made could add flexibility and possibly improve the chances of successfully implementing technical aspects of a nuclear waste policy. Experience suggests that the regulatory framework should be established prior to initiating future repository development. Concerning specifics of the regulatory framework, reasonable expectation as the standard of proof was successfully implemented and could be retained in the future; yet, the current classification system for radioactive waste, including hazardous constituents, warrants reexamination. Whether or not consideration of multiple sites are considered simultaneously in the future, inclusion of mechanisms such as deliberate use of performance assessment to manage site characterization would be wise. Because of experience gained here and abroad, diversity of geologic media is not particularly necessary as a criterion in site selection guidelines for multiple sites. Stepwise development of the repository program that includes flexibility also warrants serious consideration. Furthermore, integration of the waste management system from storage, transportation, and disposition, should be examined and would be facilitated by integration of the legal and regulatory framework. Finally, in order to enhance acceptability of future repository development, the national policy should be cognizant of those policy and technical attributes that enhance initial acceptance, and those policy and technical attributes that maintain and broaden credibility.

Radiological Aspects of Deep-Burn Fusion-Fission Hybrid Waste in a Repository

The quantity, radioactivity, and isotopic characteristics of the spent fission fuel from a hybrid fusion-fission system capable of extremely high burnups are described. The waste generally has higher activity per unit mass of heavy metal, but much lower activity per unit energy generated. The very long-term radioactivity is dominated by fission products. Simple scaling calculations suggest that the dose from a repository containing such waste would be dominated by {sup 129}I, {sup 135}Cs, and {sup 242}Pu. Use of such a system for generating energy would greatly reduce the need for repository capacity.