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Building upon the inertial confinement fusion (ICF) technology developed for the National Ignition Facility (NIF) at Lawrence Livermore National Laboratory (LLNL), a laser-driven inertial fusion energy (LIFE) power plant is being designed. In this pre-conceptual design, the final optic is exposed to a variety of threats originating from the fusion target. These include prompt neutron and gamma fluxes, x-ray and ionic emissions. While x-rays and ions are stopped by the low-density chamber fill gas (6 μg/cc xenon), neutrons and gamma-rays are not significantly attenuated. In order to limit the consequences of such threats onto the penultimate optic and the rest of the laser systems, a shielding wall stands between the target chamber area and the laser bay. An optical telescope arrangement allows for the laser beam propagation from the penultimate to the final optic, through a pinhole in the shielding wall. These pinholes attenuate the neutron flux and reduce effective dose rates such that laser bay maintenance can be performed by humans. An optimum design of this laser pinhole requires a good understanding of the different design trade-offs that exist between shielding performance and survivability of the laser optical elements and are outlined in this work. This paper provides insight on the impact and influence of the pinholes on the radiation doses in the laser bay, which is located on the opposite side of the concrete shielding wall. After addressing the difficulties of evaluating shields containing penetrations, it establishes a guideline for the selection of different variables linked to the pinhole's design and gives a preliminary evaluation of the radiation fields in the laser bay. The study also helps identify the requirements to enable manual and/or remote maintenance during operation, by determining the minimum achievable effective dose rates for different shield wall designs. Since the ability to perform maintenance during plant operation is an important contributor to high laser availability, we will propose the use of non-aligned double shield walls with pinholes.

The Laser Inertial Fusion Energy (LIFE) concept is being designed to operate as either a pure fusion or hybrid fusion-fission system. The present work focuses on the pure fusion option. A key component of a LIFE engine is the fusion chamber subsystem. It must absorb the fusion energy, produce fusion fuel to replace that burned in previous targets, and enable both target and laser beam transport to the ignition point. The chamber system also must mitigate target emissions, including ions, x-rays and neutrons and reset itself to enable operation at 10-15 Hz. Finally, the chamber must offer a high level of availability, which implies both a reasonable lifetime and the ability to rapidly replace damaged components. An integrated design that meets all of these requirements is described herein.

This paper presents our conceptual design for laser drivers used in Laser Inertial Fusion Energy (LIFE) power plants. Although we have used only modest extensions of existing laser technology to en...

LLNL is developing the nuclear fusion based Laser Inertial Fusion Energy (LIFE) power plant concept. The baseline design uses a depleted uranium (DU) fission fuel blanket with a flowing molten salt coolant (flibe) that also breeds the tritium needed to sustain the fusion energy source. Indirect drive targets, similar to those that will be demonstrated on the National Ignition Facility (NIF), are ignited at ∼13 Hz providing a 500 MW fusion source. The DU is in the form of a uranium oxycarbide kernel in modified TRISO-like fuel particles distributed in a carbon matrix forming 2-cm-diameter pebbles. The thermal power is held at 2000 MW by continuously varying the 6 Li enrichment in the coolants. There are many options to be considered in the engine design including target yield, U-to-C ratio in the fuel, fission blanket thickness, etc. Here we report results of design variations and compare them in terms of various figures of merit such as time to reach a desired burnup, full-power years of operation, time and maximum burnup at power ramp-down and the overall balance of plant utilization.

We are carrying out a multidisciplinary multi-institutional program to develop the scientific and technical basis for inertial fusion energy (IFE) based on laser drivers and direct-drive targets. The key components are developed as an integrated system, linking the science, technology, and final application of a 1000-MWe pure-fusion power plant. The science and technologies developed here are flexible enough to be applied to other size systems. The scientific justification for this work is a family of target designs (simulations) that show that direct drive has the potential to provide the high gains needed for a pure-fusion power plant. Two competing lasers are under development: the diode-pumped solid-state laser (DPPSL) and the electron-beam-pumped krypton fluoride (KrF) gas laser. This paper will present the current state of the art in the target designs and lasers, as well as the other IFE technologies required for energy, including final optics (grazing incidence and dielectrics), chambers, and target fabrication, injection, and tracking technologies. All of these are applicable to both laser systems and to other laser IFE-based concepts. However, in some of the higher performance target designs, the DPPSL will require more energy to reach the same yield as with the KrF laser.

A systems model has been developed for the Laser Inertial Fusion-Fission Energy (LIFE) power plant. It combines cost-performance scaling models for the major subsystems of the plant including the laser, inertial fusion target factory, engine (i.e., the chamber including the fission and tritium breeding blankets), energy conversion systems and balance of plant. The LIFE plant model is being used to evaluate design trade-offs and to identify high-leverage R&D. At this point, we are focused more on doing self consistent design trades and optimization as opposed to trying to predict a cost of electricity with a high degree of certainty. Key results show the advantage of large scale (>1000 MWe) plants and the importance of minimizing the cost of diodes and balance of plant cost.

Molten salt with dissolved uranium is being considered for the Laser Inertial Confinement Fusion Fission Energy (LIFE) fission blanket as a backup in case a solid-fuel version cannot meet the performance objectives, for example because of radiation damage of the solid materials. Molten salt is not damaged by radiation and therefore could likely achieve the desired high burnup (>99%) of heavy atoms of 238 U. A perceived disadvantage is the possibility that the circulating molten salt could lend itself to misuse (proliferation) by making separation of fissile material easier than for the solid-fuel case. The molten salt composition being considered is the eutectic mixture of 73 mol% LiF and 27 mol% UF4, whose melting point is 490°C. The use of 232 Th as a fuel is also being studied. ( 232 Th does not produce Pu under neutron irradiation.) The temperature of the molten salt would be ~550°C at the inlet (60°C above the solidus temperature) and ~650°C at the outlet. Mixtures of U and Th are being considered. To minimize corrosion of structural materials, the molten salt would also contain a small amount (~1 mol%) of UF3. The same beryllium neutron multiplier could be used as in the solid fuel case; alternatively, a liquid lithium or liquid lead multiplier could be used. Insuring that the solubility of Pu 3+ in the melt is not exceeded is a design criterion. To mitigate corrosion of the steel, a refractory coating such as tungsten similar to the first wall facing the fusion source is suggested in the highneutron-flux regions; and in low-neutron-flux regions, including the piping and heat exchangers, a nickel alloy, Hastelloy, would be used. These material choices parallel those made for the Molten Salt Reactor Experiment (MSRE) at ORNL. The nuclear performance is better than the solid fuel case. At the beginning of life, the tritium breeding ratio is unity and the plutonium plus 233 U

During the past two years a team at Lawrence Livermore National Laboratory (LLNL) has used Raytheon’s TopAct code to convert a variety of CAD models into TART and MCNP Monte Carlo input files. TopAct offers the possibility of enormous savings by largely eliminating the need for manual generation of models via combinatorial geometry. Also, TopAct is expected to deliver improvements in quality assurance and configuration management. We detail our experiences with various test problems. The reader will see the steady improvements that have been made in the conversion process and understand our expectations for further progress. Finally, we explain how TopAct will become a cornerstone of our future neutronics efforts.

The design of a wide range of components in and near the target bay of the National Ignition Facility (NIF) must allow for significant radiation from neutrons and gammas. Detailed 3D Monte Carlo simulations are critical to determine neutron and gamma fluxes for all target-bay components to allow optimization of location and auxiliary shielding. Demonstration of ignition poses unique challenges because of the large range (∼3 orders of magnitude) in the yield for any given attempt at ignition. Some diagnostics will provide data independent of yield, while others will provide data for lower yields and only survive high yields with little or no damage. In addition, for a given yield there is a more than 10 orders of magnitude range in neutron and gamma fluxes depending on location in the facility. For example, sensitive components in the diagnostic mezzanines and switchyards require auxiliary shielding for high-yield shots even though they are greater than 17 meters from target chamber center (TCC) and shielded by the 2 m-thick target-bay wall. In contrast, there are components 0.2 to 2 m from TCC with little or no shielding. For these components, particular attention is being made to use low-activation material because of the extremely high neutron loading levels. Many of the components closest to target center are designed to be single use to reduce worker dose from having to refurbish highly activated components. The cryogenic target positioner is an example where activation and ease of component replacement is an important part of the design. We are developing a design process for all target-bay systems that will assure reliable operation for the full range of planned yields.

Z-pinch driven IFE (Z-IFE) requires the design of a repetitive target insertion system that allows coupling of the pulsed power to the target with adequate standoff, and a chamber that can withstand blast and radiation effects from large yield targets. The present strategy for Z-IFE is to use high yield targets (∼2–3 GJ/shot), low repetition rate per chamber (∼0.1 Hz), and 10 chambers per power plant. In this study, we propose an alternative power plant configuration that uses very high yield targets (20 GJ/shot) in a single chamber operating at 0.1 Hz. A thick-liquid-wall chamber is proposed to absorb the target emission (X-rays, debris and neutrons) and mitigate the blast effects on the chamber wall. The target is attached to the end of a conical shaped recyclable transmission line (RTL) made from a solid coolant (e.g., frozen flibe), or a material that is easily separable from the coolant (e.g., steel). The RTL/target assembly is inserted through a single opening at the top of the chamber for each shot. This study looks at the RTL material choice from a safety and environmental point of view. Materials were assessed according to waste disposal rating (WDR) and contact dose rate (CDR). Neutronics calculations, using the TART2002 Monte Carlo code from Lawrence Livermore National Laboratory (LLNL), were performed for the RTL and Z-IFE chamber, and key results reported here.

The high average power laser (HAPL) program aims at developing laser inertial fusion energy (Laser IFE) based on lasers, direct drive targets and a solid wall chamber. The preferred first wall configuration is based on tungsten and ferritic steel as armor and structural materials, respectively. A key concern is the survival of the first wall under the X-ray and ion energy deposition from the fusion micro-explosion. The HAPL design and R&D effort in the chamber and material area is focused toward understanding and resolving the key armor survival issues. This includes modeling and experimental testing of the armor thermo-mechanical behavior in facilities utilizing ion, X-rays and laser sources to simulate IFE conditions. Helium management is addressed by conducting implantation experiments along with modeling of He behavior in tungsten. This paper summarizes the HAPL chamber activities. The first wall/armor configuration and design analysis are described, key chamber issues are discussed, and the R&D to address them is highlighted.

In this paper we report on two Heavy Ion Fusion (HIF) driver point design studies. The Robust Point Design (RPD) was completed over a year ago, and the Modular Point Design (MPD) is still in progress. The goal of any point design study is to construct a detailed design that is self-consistent and integrated from injector to target. This has been the primary theme of both studies.

Dry-wall inertial fusion energy (IFE) power plants must survive repeated exposure to target threats that include x-rays, ions, and neutrons. While this exposure may lead to sputtering, exfoliation, transmutation, and swelling, more basic effects are thermomechanical in nature. In the present work, we use the newly developed RadHeat code to predict time-temperature profiles in a tungsten armor, which has been proposed for use in an IFE power plant. The XAPPER x-ray damage experiment is used to simulate thermal effects by operating at fluences that produce similar peak temperatures, temperature gradients, or thermomechanical stresses. Soft x-ray fluences in excess of 1 J/cm{sup 2} are possible. Using RadHeat, we determine the XAPPER x-ray fluence needed to match expected peak surface temperatures. Such calculations are the first step in predicting the thermomechanical effects that are expected in an IFE system. Here, we report our findings and detail directions for future experiments and modeling.

The long-range goal of the Z-Pinch IFE program is to produce an economically-attractive power plant using high-yield z-pinch-driven targets (~3GJ) with low rep-rate per chamber (~0.1 Hz). The prese...

The ARIES-IFE study was an integrated study of inertial fusion energy (IFE) chambers and chamber interfaces with the driver and target systems. Detailed analysis of various subsystems was performed parametrically to uncover key physics/technology uncertainties and to identify constraints imposed by each subsystem. In this paper, these constraints (e.g., target injection and tracking, thermal response of the first wall, and driver propagation and focusing) were combined to understand the trade-offs, to develop operational windows for chamber concepts, and to identify high-leverage research and development directions for IFE research. Some conclusions drawn in this paper are (a) the detailed characterization of the target yield and spectrum has a major impact on the chamber; (b) it is prudent to use a thin armor instead of a monolithic first wall for dry-wall concepts; (c) for dry-wall concepts with direct-drive targets, the most stringent constraint is imposed by target survival during the injection process; (d) for relatively low yield targets (

A coordinated, focused effort is underway to develop Laser Inertial Fusion Energy. The key components are developed in concert with one another and the science and engineering issues are addressed concurrently. Recent advances include: target designs have been evaluated that show it could be possible to achieve the high gains (>100) needed for a practical fusion system.These designs feature a low-density CH foam that is wicked with solid DT and over-coated with a thin high-Z layer. These results have been verified with three independent one-dimensional codes, and are now being evaluated with two- and three-dimensional codes. Two types of lasers are under development: Krypton Fluoride (KrF) gas lasers and Diode Pumped Solid State Lasers (DPSSL). Both have recently achieved repetitive 'first light', and both have made progress in meeting the fusion energy requirements for durability, efficiency, and cost. This paper also presents the advances in development of chamber operating windows (target survival plus no wall erosion), final optics (aluminium at grazing incidence has high reflectivity and exceeds the required laser damage threshold), target fabrication (demonstration of smooth DT ice layers grown over foams, batch production of foam shells, and appropriate high-Z overcoats), and target injection (new facility for target injection and tracking studies).

The Heavy Ion Fusion Virtual National Laboratory (HIF-VNL) recently initiated an effort to reach an updated, self-consistent, integrated point design for a thick-liquid inertial fusion energy power plant. We call this design the Robust Point Design. As part of this effort, the shielding design of the final focusing system has been evaluated, in an iterative fashion, with other elements of the design. The present work reports on the status of the shielding design from the perspectives of superconductor/insulator radiation lifetimes, recirculating power needed to counter nuclear heating, and neutron activation, which affects both system maintainability and waste management. Models used herein include the last three focusing magnets, and a full, three-dimensional model for the target chamber. Analyses have been performed for 9-by-9 beam arrays, with a total of 120 beams (60 per side). Results and directions for future work are presented.

Over the past several years, significant progress has been made in the analysis of safety and environmental (S and E) issues for inertial fusion energy (IFE). Detailed safety assessments have been performed for the baseline power plant concepts, as well as for a conceptual target fabrication facility. Safety analysis results are helping to drive the agenda for experiments. A survey of the S and E characteristics - both radiological and chemical - of candidate target materials has been completed. Accident initiating events have been identified and incorporated into master logic diagrams, which will be essential to the detailed safety analyses that will be needed in the future. Studies of aerosol generation and transport will have important safety implications. A Monte Carlo-based uncertainty analysis procedure has been developed for use in neutron activation calculations. Finally, waste management issues are receiving increased attention and are deserving of further discussion.

We have performed molecular dynamics simulations of radiation damage in fused silica. In this study, we discuss the role of successive cascade overlap on the saturation and self-healing of oxygen vacancy defects in the amorphous fused silica network. Furthermore, we present findings on the topological changes in fused silica due to repeated energetic recoil atoms. These topological network modifications consistent with experimental Raman spectroscopic observation on neutron and ion irradiated fused silica are indicators of permanent densification that has also been observed experimentally.

Silica is one of the candidate materials for final focusing mirrors in inertial fusion reactors. These materials will be exposed to high irradiation fluxes during operation. Radiation damage results in point defects that can lead to obscuration, that is, degradation of the optical properties of these materials. A basic understanding of defect production and migration in these materials is, however, limited. In this paper we present molecular dynamics simulations of defect production in silica glass due to energetic recoils. We compute the oxygen deficient centers generated during irradiation at energies between 1 and 5 keV and identify the mechanisms for production and recombination at short time scales.

Traditionally, waste management studies for fusion energy have used the waste disposal rating (WDR) to evaluate if radioactive material from irradiated structures could qualify for shallow land burial. However, given the space limitations and the negative public perception of large volumes of waste, there is a growing international motivation to develop a fusion waste management system that maximizes the amount of material that can be cleared or recycled. In this work, we present an updated assessment of the waste management options for the HYLIFE-II inertial fusion energy (IFE) power plant, using the concept of clearance index (CI) for radioactive waste disposal. With that purpose, we have performed a detailed neutronics analysis of the HYLIFE-II design, using the tart and acab computer codes for neutron transport and activation, respectively. Whereas the traditional version of acab only provided the user with the γ contact dose rate for recycling assessments and WDR as an index for waste disposal considerations, here we have modified the code to calculate CIs using the current international atomic energy agency (IAEA) clearance limits for radiological waste disposal. The results from the analysis are used to perform an assessment of the waste management options for the HYLIFE-II IFE design.

In an inertial fusion energy (IFE) thick-liquid chamber design such as HYLIFE-II, a molten-salt is used to attenuate neutrons and protect the chamber structures from radiation damage. In the case of a fast ignition inertial fusion system, advanced targets have been proposed that may be self-sufficient in terms of tritium breeding (i.e. the amount of tritium bred in target exceeds the amount burned). This aspect allows for greater freedom when selecting a liquid for the protective blanket, given that lithium-bearing compounds are no longer required. Materials selection may now be based upon other characteristics, such as safety and environmental (S&E), pumping power, corrosion, and vapor pressure, along with others. The present work assesses the characteristics of many single, binary, and ternary molten-salts and liquid metals using the NIST Properties of Molten Salts Database. As an initial screening, liquids were evaluated for their S&E characteristics, which included an assessment of waste disposal rating (WDR), contact dose, and radioactive afterheat. Liquids that passed the S&E criteria were then evaluated for required pumping power. The pumping power was calculated using three components: velocity head losses, frictional losses, and lifting power. The results of the assessment are used to identify those materials that are suitable for potential liquid-chamber fast-ignition IFE concepts, from both the S&E and pumping power perspective. Recommendations for further analysis are also made.

First structural wall (FSW) materials in inertial fusion energy (IFE) power reactors will be irradiated under typical repetition rates of 1–10 Hz and operation times as long as the total reactor lifetime. The main objective of the present work is to determine whether or not a continuous-pulsed (CP) approach could be an accurate and practical methodology in modeling the pulsed activation process for operating conditions of FSW materials. To do that, we assess the applicability of the CP model to predict the neutron-induced activation in the FSW material of the HYLIFE-II reactor. It is demonstrated that a CP approach consisting of a continuous irradiation period followed by a series of only a few pulses prior to shutdown can efficiently model the real pulsed operating regime of the FSW material, in terms of both accuracy and CPU time consumption. Pros and cons of the model when compared with an equivalent steady-state (ESS) method are discussed, which is useful to assess how conclusions of earlier activation studies that used the ESS approach might be different when using a more realistic model. Comparison with the exact pulsed (EP) modeling is also performed. Finally, application of the CP model to other inertial confinement fusion operating modes, such as that of the NIF facility, is suggested for future work.

Recent work in heavy-ion fusion accelerators and final focusing systems shows a trend towards less current per beam, and thus, a greater number of beams. Final focusing magnets are susceptible to nuclear heating, radiation damage, and neutron activation. The trend towards more beams, however, means that there can be less shielding for each magnet. Excessive levels of nuclear heating may lead to magnet quench or to an intolerable recirculating power for magnet cooling. High levels of radiation damage may result in short magnet lifetimes and low reliability. Finally, neutron activation of the magnet components may lead to difficulties in maintenance, recycling, and waste disposal. The present work expands upon previous, three-dimensional magnet shielding calculations for a modified version of the HYLIFE-II IFE power plant design. We present key magnet results as a function of the number of beams. # 2001 Elsevier Science B.V. All rights reserved.

Previous studies of the safety and environmental aspects of the HYLIFE-II inertial fusion energy power plant design have used simplistic assumptions in order to estimate radioactivity releases under accident conditions. Conservatisms associated with these traditional analyses can mask the actual behavior of the plant and have revealed the need for more accurate modeling and analysis of accident conditions and radioactivity mobilization mechanisms. In the present work, computer codes traditionally used for magnetic fusion safety analyses (CHEMCON, MELCOR) have been applied for simulating accident conditions in a simple model of the HYLIFE-II IFE design. Here we consider a severe loss of coolant accident (LOCA) in conjunction with simultaneous failures of the beam tubes (providing a pathway for radioactivity release from the vacuum vessel towards the confinement) and of the two barriers surrounding the chamber (inner shielding and confinement building itself). Even though confinement failure would be a very unlikely event it would be needed in order to produce significant off-site doses. CHEMCON code allows calculation of long-term temperature transients in fusion reactor first wall, blanket, and shield structures resulting from decay heating. MELCOR is used to simulate a wide range of physical phenomena including thermal-hydraulics, heat transfer, aerosol physics and fusion product transport and release. The results of these calculations show that the estimated off-site dose is less than 5 mSv (0.5 rem), which is well below the value of 10 mSv (1 rem) given by the DOE Fusion Safety Standards for protection of the public from exposure to radiation during off-normal conditions.

Target materials for inertial fusion energy (IFE) power plant designs might be selected for a wide variety of reasons including wall absorption of driver energy, material opacity, cost and ease of fabrication. While each of these issues are of great importance, target materials should also be selected based upon their safety and environmental (S&E) characteristics. The present work focuses on the recycling, waste management and accident dose characteristics of potential target materials. If target materials are recycled so that the quantity is small, isotopic separation may be economically viable. Therefore, calculations have been completed for all stable isotopes for all elements from lithium to polonium. The results of these calculations are used to identify specific isotopes and elements that are most likely to be offensive as well as those most likely to be acceptable in terms of their S&E characteristics.

CYLINDRICAL LIQUID JET GRIDS FOR BEAM-PORT PROTECTION OF THICK-LIQUID HEAVY-ION FUSION TARGET CHAMBERS R. Abbott, S. Pemberton, P.F. Peterson G.-P. Sun, P. Wright R. Holmes J. Latkowski, R. Moir, K. Springer Dept. of Nuclear Engineering Dept. of Mechanical Eng. Lawrence Livermore National Laboratory University of California University of California Livermore, CA 94551 Berkeley, CA 94720-1730 Berkeley, CA 94720-1730 ABSTRACT Thick-liquid pockets have the potential to protect structural materials and increase power density in heavy-ion fusion chambers. Here we show that cylindrical liquid jets have interesting advantages for creating shielding grids for heavy-ion beam lines. A cylindrical nozzle design with a very low convergence ratio was developed, and the fabrication methods needed for inexpensive numerically-controlled machining of large nozzle arrays demonstrated. Cylindrical jets were studied because they give the highest surface smoothness for a given degree of turbulence suppression, allow flow control to individual nozzles for control of jet pointing, and attenuate target-induced shocks effectively. Improved control of the grid geometry allows the driver energy to be delivered by a larger number of beams. These smaller beams—up to 160 in the example here—improve focusing and reduce neutron collimation up beam lines. I. INTRODUCTION The use of liquid jets for thick-liquid shielding for inertial fusion energy (IFE) target chambers remains an important goal for heavy-ion fusion, due to the potential to effectively protect target chamber structures from neutron damage. Additional advantages come from reductions in activation and waste generation, and from the capability to reduce final-focus magnet stand off distance. Liquid protection for fusion chambers was first suggested in the early 1970’s

SOMBRERO (solid moving breeder reactor) is a conceptual design of a 1000 MW e laser-driven inertial fusion energy (IFE) power plant. An important goal of the original study was the achievement of a safe and environmentally attractive reactor of relatively simple design. However, recent work has pointed out some key issues involving safety that were not completely addressed at that time, and which need to be reviewed in order to maximize the SOMBRERO design attractiveness. The present work uses a set of computer codes traditionally used for magnetic fusion safety studies (CHEMCON, MELCOR), which have been adopted and adapted for use in IFE safety analysis. Here we consider a loss of flow accident (LOFA) combined with a simultaneous loss of vacuum accident (LOVA) produced by a breach in the confinement building. Although confinement failure would be a very unlikely event, it must be postulated in order to produce significant off-site doses. The CHEMCON code is used to simulate the long-term thermal transient in the reactor structures resulting from oxidation and radioactive decay heat. MELCOR is used to simulate a wide range of physical phenomena including thermal-hydraulics, heat transfer, aerosol physics and fusion product release and transport. As specified in the DOE Fusion Safety Standards, an off-site dose below 1 rem (10 mSv) is the requirement to avoid public sheltering and evacuation. The SOMBRERO accident analysis results will be evaluated according to this limit and suggestions will be made for improvements and future work.

At the Thirteenth International Symposium on Heavy Ion Inertial Fusion (HIF Symposium), we presented magnet shielding calculations for 72-, 128, 200, and 288-beam versions of the HYLIFE-II power plant design. In all cases, we found the radiation-limited lifetimes of the last set of final focusing magnets to be unacceptably short. Since that time, we have completed follow-on calculations to improve the lifetime of the 72-beam case. Using a self-consistent final focusing model, we vary parameters such as the shielding thicknesses and compositions, focusing length, angle-of-attack to the target, and the geometric representation of the flibe pocket, chamber, and blanket. By combining many of these shielding features, we are able to demonstrate a magnet shielding design that would enable the last set of final focusing magnets to survive for the lifetime of the power plant.

We estimate possible ranges of tritium inventories for an inertial fusion energy (IFE) target fabrication facility producing various types of targets and using various production technologies. Target fill is the key subtask in determining the overall tritium inventory for the plant. By segmenting the inventory into multiple, parallel production lines--each with its own fill canister--and including an expansion tank to limit releases, we are able to ensure that a target fabrication facility would meet the accident dose goals of 10 mSv (1 rem) set forth in the Department of Energy's Fusion Safety Standards. For indirect-drive targets, we calculate release fractions for elements from lithium to bismuth and show that nearly all elements meet the dose goal. Our work suggests directions for future R&D that will help reduce total tritium inventories and increase the flexibility of target fabrication facilities.

Inertial confinement fusion (ICF) devices, both test/experimental facilities and fusion energy (IFE) power plants, will operate in a pulsed mode. However, the pulsing schedule in these devices is very different, and it could range from one shot every several days in an experimental facility to some Hz in IFE reactors. The main objective of the present work is to determine whether or not a continuous-pulsed (CP) approach could be an accurate and practical methodology in modeling the pulsed activation experienced by chamber materials of both types of devices. In testing the applicability of the CP irradiation model, we used materials and neutron environment scenarios of the HYLIFE-II reactor and the NIF experimental facility. It is demonstrated that a CP approach consisting of a continuous irradiation period followed by a series of only a few pulses prior to shutdown, can efficiently model the real pulsed operating regimes of the chamber materials, in terms of both accuracy and CPU time consumption. Pros and cons of the model when compared with an equivalent steady-state (ESS) method are discussed, and comparison with the exact pulsed (EP) modeling is also performed.

Estimating radiological risks is an essential part of an assessment of fusion as an attractive source of energy. Due to the limited data specific to radionuclides of interest to fusion reactors, one of the goals of this work is to expand the Dose Conversion Factors (DCF) library for use in the calculation of different types of off-site doses and associated health effect consequences. This expansion accounts for about 300 radionuclides included in accidental activity releases from HYLIFE-II and SOMBRERO IFE Power Plants. Furthermore, for each of the radionuclides included in the new DCF library, we address a parametric study of accident consequences by varying the atmospheric stability, wind speed, rain conditions, and thermal plume rise. The results of these calculations allow us to identify the most troublesome radionuclides in terms of safety consequences as well as the impact of the different atmospheric scenarios.

The present work continues our effort to perform an integrated safety analysis for the HYLIFE-II inertial fusion energy (IFE) power plant design. Recently we developed a base case for a severe accident scenario in order to calculate accident doses for HYLIFE-II. It consisted of a total loss of coolant accident (LOCA) in which all the liquid flibe (Li 2 BeF 4 ) was lost at the beginning of the transient. Results showed that the off-site dose was below the limit given by the DOE Fusion Safety Standards for public protection in case of accident, and that this dose was dominated by the tritium released during the accident. In order to further advance a complete safety analysis for HYLIFE-II, a range of other accident scenarios must be considered. In this work, we introduce a new version of the MELCOR thermal-hydraulics code recently developed by the Idaho National Engineering and Environmental Laboratory (INEEL) that uses flibe as the working fluid. We have focused on a loss of flow accident (LOFA), with simultaneous failure of the blanket structure and the beam tubes that connect the chamber with the outside of the confinement building. This constitutes the bypass needed to communicate the target chamber with the environment. Once the release fractions of the various radioactivity sources are known, we calculate off-site doses under different conditions as a consequence of the accident.

In December 1999 and January 2000, a 40-cm-thick spherical shell of sprayable concrete (gunite) was applied to the exterior surface of the National Ignition Facility (NIF) target chamber Glow-discharge mass spectroscopy has been used to determine the elemental composition of multiple gunite samples, which were collected at the time of application. These measured compositions are compared to the anticipated composition and both are used for neutron activation calculations. Contact dose rates are reported and implications for doses rates during operation and for the eventual facility decommissioning are discussed.

Although the safety and environmental (S & E) characteristics of fusion energy have long been emphasized, these benefits are not automatically achieved. To maximize the potential S & E attractiveness of the inertial fusion energy (IFE), analyses must be performed early in the designs so that lessons can be learned and intelligent decisions made. In this work we have introduced for the first time heat transfer and thermal-hydraulics calculations as part of a state-of-the-art set of codes and libraries in order to establish an updated methodology for IFE safety analysis. We have focused our efforts primarily on two IFE power plant conceptual designs: HYLIFE-II and SOMBRERO. To some degree, these designs represent the extremes in IFE power plant designs. Also, a preliminary safety assessment has been performed for a generic target fabrication facility producing various types of targets and using various production techniques. Although this study cannot address all issues and hazards posed by an IFE power plant, it advances our understanding of radiological safety of such facilities. This will enable better comparisons between IFE designs and competing technologies from the safety point of view.

First structural wall material (FSW) materials in inertial fusion energy (IFE) power reactors will be irradiated under typical repetition rates of 1–10 Hz, for an operation time as long as the total reactor lifetime. The main objective of the present work is to determine whether a continuous-pulsed (CP) approach can be an efficient method in modeling the pulsed activation process for operating conditions of FSW materials. The accuracy and practicability of this method was investigated both analytically and (for reaction/decay chains of two and three nuclides) by computational simulation. It was found that CP modeling is an accurate and practical method for calculating the neutron-activation of FSW materials. Its use is recommended instead of the equivalent steady-state method or the exact pulsed modeling. Moreover, the applicability of this method to components of an IFE power plant subject to repetition rates lower than those of the FSW is still being studied. The analytical investigation was performed for 0.05 Hz, which could be typical for the coolant. Conclusions seem to be similar to those obtained for the FSW. However, further future work is needed for a final answer.

The National Ignition Facility (NIF) is a US Department of Energy inertial confinement laser fusion experimental facility currently under construction at the Lawrence Livermore National Laboratory (LLNL). To ensure that decontamination and decommissioning (D&D) issues at the end-of-life are manageable, this subject has received attention from an early stage. This paper summarizes the NIF D&D issues, and the status of the D&D plan.

The Laser Inertial Fusion-based Energy (LIFE) engine design builds upon on going progress at the National Ignition Facility (NIF) and offers a near-term pathway to commercial fusion. Fusion technologies that are critical to success are reflected in the design of the first wall, blanket and tritium separation subsystems. The present work describes the LIFE engine-related components and technologies. LIFE utilizes a thermally robust indirect-drive target and a chamber fill gas. Coolant selection and a large chamber solid-angle coverage provide ample tritium breeding margin and high blanket gain. Target material selection eliminates the need for aggressive chamber clearing, while enabling recycling. Demonstrated tritium separation and storage technologies limit the site tritium inventory to attractive levels. These key technologies, along with the maintenance and advanced materials qualification program have been integrated into the LIFE delivery plan. This describes the development of components and subsystems, through prototyping and integration into a First Of A Kind power plant.

Two scenarios are typically envisioned for thorium fuel cycles: 'open' cycles based on irradiation of {sup 232}Th and fission of {sup 233}U in situ without reprocessing or 'closed' cycles based on irradiation of {sup 232}Th followed by reprocessing, and recycling of {sup 233}U either in situ or in critical fission reactors. This study evaluates a third option based on the possibility of breeding fissile material in a fusion-fission hybrid reactor and burning the same fuel in a critical reactor without any reprocessing or reconditioning. This fuel cycle requires the hybrid and the critical reactor to use the same fuel form. TRISO particles embedded in carbon pebbles were selected as the preferred form of fuel and an inertial laser fusion system featuring a subcritical blanket was combined with critical pebble bed reactors, either gas-cooled or liquid-salt-cooled. The hybrid reactor was modeled based on the earlier, hybrid version of the LLNL Laser Inertial Fusion Energy (LIFE1) system, whereas the critical reactors were modeled according to the Pebble Bed Modular Reactor (PBMR) and the Pebble Bed Advanced High Temperature Reactor (PB-AHTR) design. An extensive neutronic analysis was carried out for both the hybrid and the fission reactors in order to track the fuelmore » composition at each stage of the fuel cycle and ultimately determine the plant support ratio, which has been defined as the ratio between the thermal power generated in fission reactors and the fusion power required to breed the fissile fuel burnt in these fission reactors. It was found that the maximum attainable plant support ratio for a thorium fuel cycle that employs neither enrichment nor reprocessing is about 2. This requires tuning the neutron energy towards high energy for breeding and towards thermal energy for burning. A high fuel loading in the pebbles allows a faster spectrum in the hybrid blanket; mixing dummy carbon pebbles with fuel pebbles enables a softer spectrum in the critical reactors. This combination consumes about 20% of the thorium initially loaded in the hybrid reactor ({approx}200 GWd/tHM), partially during hybrid operation, but mostly during operation in the critical reactor. The plant support ratio is low compared to the one attainable using continuous fuel chemical reprocessing, which can yield a plant support ratio of about 20, but the resulting fuel cycle offers better proliferation resistance as fissile material is never separated from the other fuel components.« less

ICF power plants, such as the LIFE scheme at LLNL, may employ a high-Z, target-chamber gas-fill to moderate the first-wall heat-pulse due to x-rays and energetic ions released during target detonation. To reduce the uncertainties of cooling and beam/target propagation through such gas-filled chambers, we present a pulsed plasma source producing 2-5 eV plasma comprised of high-Z gases. We use a 5-kJ, 100-ns theta discharge for high peak plasma-heating-power, an electrode-less discharge for minimizing impurities, and unobstructed axial access for diagnostics and beam (and/or target) propagation studies. We will report on the plasma source requirements, design process, and the system design.

The LIFE minichamber experiment will investigate cooling of the strongly radiating Xe buffer gas protecting the LIFE chamber wall. A theta pinch will inductively heat a few cc of Xe at ion density 2e16/cc to several eV. Thomson scattering will be used to determine electron temperature and ionization state. Modeled is being done using the magnetohydrodynamic code HYDRA with an external circuit model and inductive feedback from the plasma to the external circuit. Coil stresses are being assessed using the 3D MHD code ALE3D. A major challenge to the design is the paucity of opacity and conductivity data for Xe in the buffer gas regime. Results of the modeling will be presented. *This work performed under the auspices of the U.S. Department of Energy by Lawrence Livermore National Laboratory under Contract DE-AC52-07NA27344.

The goals of the Laser Inertial Fusion Fission Energy (LIFE) is to use fusion neutrons to fission materials with no enrichment and minimum processing and have greatly reduced wastes that are not of interest to making weapons. Fusion yields expected to be achieved in NIF a few times per day are called for with a high reliable shot rate of about 15 per second. We have found that the version of LIFE using TRISO fuel discussed in other volumes of this series can be modified by replacing the molten-flibe-cooled TRISO fuel zone with a molten salt in which the same actinides present in the TRISO particles are dissolved in the molten salt. Molten salts have the advantage that they are not subject to radiation damage, and hence overcome the radiation damage effects that may limit the lifetime of solid fuels such as TRISO-containing pebbles. This molten salt is pumped through the LIFE blanket, out to a heat exchanger and back into the blanket. To mitigate corrosion, steel structures in contact with the molten salt would be plated with tungsten or nickel. The salt will be processed during operation to remove certain fission products (volatile and noble and semi-noble fission products),more » impurities and corrosion products. In this way neutron absorbers (fission products) are removed and neutronics performance of the molten salt is somewhat better than that of the TRISO fuel case owing to the reduced parasitic absorption. In addition, the production of Pu and rare-earth elements (REE) causes these elements to build up in the salt, and leads to a requirement for a process to remove the REE during operation to insure that the solubility of a mixed (Pu,REE)F3 solid solution is not exceeded anywhere in the molten salt system. Removal of the REE will further enhance the neutronics performance. With molten salt fuels, the plant would need to be safeguarded because materials of interest for weapons are produced and could potentially be removed.« less

A comprehensive assessment on the eligibility of reduced activation (RA) steels as structural chamber material in Inertial Fusion Energy (IFE) thick-liquid concepts is performed. As far as alloying elements, it is shown that the activation of tungsten is a question to debate. Regarding impurity elements, it is analyzed if they could question the possibility of obtaining real RA steels for shallow land burial (SLB). The effect of the thickness of the liquid wall on the SLB response of alloying and impurity elements is computed. And above all, we have estimated the impact of cross section uncertainties when addressing the former questions, and we identify those to be improved. The necessary improvement of some tungsten and niobium cross sections is justified.

The FY04 LLNL study of Z-IFE [1] proposed and evaluated a design that deviated from SNL's previous baseline design. The FY04 study included analyses of shock mitigation, stress in the first wall, neutronics and systems studies. In FY05, the subject of this report, we build on our work and the theme of last year. Our emphasis continues to be on alternatives that hold promise of considerable improvements in design and economics compared to the base-line design. Our key results are summarized here.

The heavy ion fusion (HIF) program is interested in developing a high brightness ion source for high energy density physics (HEDP) experiments. One possible approach to obtaining higher brightness may be to raise the surface temperature of the ion source just prior to extraction. The current ion source material being studied is a layer of potassium-doped aluminosilicate bonded to a tungsten substrate. It is speculated that if the surface temperature of the source is raised above 1200 C (from a steady-state temperature of 900 C) for time periods on the order of 100's of nanoseconds, current densities of greater than 100 mA/cm{sup 2} of ions may be achievable. Typical aluminosilicate sources produce ion current densities (either K+ or Na+ ions) of {approx}10 mA/cm{sup 2} (at 1100 C). A number of heating methods might be possible, including lasers, diode arrays, and flash lamps. Here we assume laser heating. In this preliminary study, we used the LLNL RadHeat code to model the time-temperature history of the surface when hit by laser pulses and illustrate how RadHeat can be used to optimize the surface temperature response. Also of interest is the temperature history of the interface temperature between the ceramic and the metal more » layers. This is also investigated. « less

The promise of fusion as a viable 21st century energy source requires the development of advanced structural (MFE and IFE) and optical (IFE) materials that are capable of withstanding the harsh radiation environment that leads to the degradation of physical and mechanical properties. Materials in fusion environments must be able to handle 14 MeV neutrons produced from Deuterium-Tritium nuclear reactions, as well as the insoluble He and reactive H gases that lead to swelling and embrittlement. Additionally, with the requirement of very high thermal loads makes the development of new advanced materials a formidable challenge. The scope of this study was to determine the feasibility of using atomistic simulations to predict the radiation response of novel materials engineered with potentially self-healing properties to survive in radiation environments over very long time-scales. The class of materials that shows promise is what is called a nanocrystalline material. Nanocrystalline materials are defined as those having very fine grains on the order of several to tens of nanometers in size, and consequently very high grain-boundary to volume ratio. Experimental observations [1] suggests that these grain-boundary networks can act as sinks for defects and hence promote self-repair.

PLEX is a Z-pinch based x-ray source that can produce x-rays with fluences (0.3-18 J/cm{sup 2}), pulselengths (10-30 ns), repetition rates (

In a thick-liquid protected chamber design, such as HYLIFE-II, a molten-salt is used to attenuate neutrons and protect the chamber structures from radiation damage. The molten-salt absorbs some of the material and energy given off by the target explosion. In the case of a fast ignition inertial fusion system, advanced targets have been proposed that may be Self-sufficient in the tritium breeding (i.e., the amount of tritium bred in target exceeds the amount burned). These ''tritium-lean'' targets contain approximately 0.5% tritium and 99.5% deuterium, but require a large pr of 10-20 g/cm{sup 2}. Although most of the yield is provided by D-T reactions, the majority of fusion reactions are D-D, which produces a net surplus of tritium. This aspect allows for greater freedom when selecting a liquid for the protective blanket (lithium-bearing compounds are not required). This study assesses characteristics of many single, binary, and ternary molten-salts. Using the NIST Properties of Molten Salts Database, approximately 4300 molten-salts were included in the study [1]. As an initial screening, salts were evaluated for their safety and environmental (S&E) characteristics, which included an assessment of waste disposal rating, contact dose, and radioactive afterheat. Salts that passed the S&E criteria were then evaluatedmore » for neutron shielding ability and pumping power. The pumping power was calculated using three components: velocity head losses, frictional losses, and lift. This assessment left us with 57 molten-salts to recommend for further analysis. Many of these molten-salts contain elements such as sodium, lithium, beryllium, boron, fluorine, and oxygen. Recommendations for further analysis are also made.« less