S. Reyes

Timely Delivery of Laser Inertial Fusion Energy (LIFE)

Abstract The National Ignition Facility (NIF), the world’s largest and most energetic laser system, is now operational at Lawrence Livermore National Laboratory. A key goal of the NIF is to demonstrate fusion ignition for the first time in the laboratory. Its flexibility allows multiple target designs (both indirect and direct drive) to be fielded, offering substantial scope for optimization of a robust target design. In this paper we discuss an approach to generating gigawatt levels of electrical power from a laser-driven source of fusion neutrons based on these demonstration experiments. This “LIFE” concept enables rapid time-to-market for a commercial power plant, assuming success with ignition and a technology demonstration program that links directly to a facility design and construction project. The LIFE design makes use of recent advances in diode-pumped, solid-state laser technology. It adopts the paradigm of Line Replaceable Units utilized on the NIF to provide high levels of availability and maintainability and mitigate the need for advanced materials development. A demonstration LIFE plant based on these design principles is described, along with the areas of technology development required prior to plant construction.

CHAMBER DESIGN FOR THE LASER INERTIAL FUSION ENERGY (LIFE) ENGINE

Abstract 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.

Selection of IFE target materials from a safety and environmental perspective

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 and 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 and E characteristics.

Updated Modeling of Postulated Accident Scenarios in ITER

Abstract This paper gives an overview of the latest work on ITER accident analysis, describing the methodology and presenting some updated results. There are currently 25 ITER Reference Events, divided into two categories: incidents and accidents. Starting from the 2001 list of events, several new scenarios have been added, including fire events. Other former Reference Events have been updated and in some cases fully re-analyzed due to design modifications, such as changes in the confinement arrangements. The results demonstrate that the ITER General Safety Objectives are met and that the safety features of the ITER design are successful in minimizing the potential public and environmental consequences of off-normal events.

Threats to ICF reactor materials: computational simulations of radiation damage induced topological changes in fused silica

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.

Use of clearance indexes to assess waste disposal issues for the HYLIFE-II inertial fusion energy power plant design

Abstract 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.

Monte carlo uncertainty analyses of pulsed activation in the national ignition facility gunite shielding

The global effect of activation cross-section uncertainties on calculated radiological quantities is investigated for the first time using a methodology based on Monte Carlo random sampling. The method is applied to the calculation of the uncertainty in the contact dose rate from the gunite shielding of the National Ignition Facility chamber after 30 yr of pulsed irradiation. Some critical cross section contributing significantly to the overall uncertainty are identified. By a reasonable reduction of the uncertainty in those cross sections, the accuracy in the calculated total contact dose rate is greatly improved.

Materials-related issues in the safety and licensing of nuclear fusion facilities

Fusion power holds the promise of electricity production with a high degree of safety and low environmental impact. Favourable characteristics of fusion as an energy source provide the potential for this very good safety and environmental performance. But to fully realize the potential, attention must be paid in the design of a demonstration fusion power plant (DEMO) or a commercial power plant to minimize the radiological hazards. These hazards arise principally from the inventory of tritium and from materials that become activated by neutrons from the plasma. The confinement of these radioactive substances, and prevention of radiation exposure, are the primary goals of the safety approach for fusion, in order to minimize the potential for harm to personnel, the public, and the environment. The safety functions that are implemented in the design to achieve these goals are dependent on the performance of a range of materials. Degradation of the properties of materials can lead to challenges to key safety functions such as confinement. In this paper the principal types of material that have some role in safety are recalled. These either represent a potential source of hazard or contribute to the amelioration of hazards; in each case the related issues are reviewed. The resolution of these issues lead, in some instances, to requirements on materials specifications or to limits on their performance.

Safety analysis of the US dual coolant liquid lead-lithium ITER test blanket module

The US is proposing a prototype of a dual coolant liquid lead-lithium (DCLL) DEMO blanket concept for testing in the International Thermonuclear Experimental Reactor (ITER) as an ITER Test Blanket Module (TBM). Because safety considerations are an integral part of the design process to ensure that this TBM does not adversely impact the safety of ITER, a safety assessment has been conducted for this TBM and its ancillary systems as requested by the ITER project. Four events were selected by the ITER International Team (IT) to address specific reactor safety concerns, such as VV pressurization, confinement building pressure build-up, TBM decay heat removal capability, tritium and activation products release from the TBM system, and hydrogen and heat production from chemical reactions. This paper summarizes the results of this safety assessment conducted with the MELCOR computer code.

Safety Issues of Hg and Pb as IFE Target Materials: Radiological Versus Chemical Toxicity

Abstract We have performed a safety assessment of mercury and lead as possible hohlraum materials for Inertial Fusion Energy (IFE) targets, including for the first time a comparative analysis of the radiological and toxicological consequences of an accidental release. In order to calculate accident doses to the public, we have distinguished between accidents at the target fabrication facility and accidents at other areas of the power plant. Regarding the chemical toxicity assessment, we have used the U.S. DOE regulations to determine the maximum allowable release in order to protect the public from adverse health effects. Opposite to common belief, it has been found that the chemical safety requirements for these materials appear to be more stringent than the concentrations that would result in an acceptable radiological dose.