J.F. Latkowski

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.

Modeling defect production in silica glass due to energetic recoils using molecular dynamics simulations

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.

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.

Physics Issues Related to the Confinement of ICF Experiments in the U.S. National Ignition Facility

ICF experiments planned for the proposed US National Ignition Facility [NIF] will produce emissions of neutrons, x rays, debris, and shrapnel. The NIF Target Area [TA] must acceptably confine these emissions and respond to their effects to allow an efficient rate of experiments, from 600 to possibly 1500 per year, and minimal down time for maintenance. Detailed computer code predictions of emissions are necessary to study their effects and impacts on Target Area operations. Preliminary results show that the rate of debris shield transmission loss [and subsequent periodicity of change‐out] due to ablated material deposition is acceptable, neutron effects on optics are manageable, and preliminary safety analyses show a facility rating of low hazard, non‐nuclear. Therefore, NIF Target Area design features such as fused silica debris shields, refractory first wall coating, and concrete shielding are effective solutions to confinement of ICF experiment emissions.

Pulsed X-Ray Exposures and Modeling for Tungsten as an IFE First Wall Material

Abstract 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/cm2 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.

Liquid Wall Options for Tritium-Lean Fast Ignition Inertial Fusion Energy Power Plants

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.

Pulsed activation of structural materials in IFE chambers

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.

Efficient modeling for pulsed activation in inertial fusion energy reactors

Abstract 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.

Sequential Charged-Particle and Neutron Activation of Flibe in the Hylife-II Inertial Fusion Energy Power Plant Design

Most radionuclide generation/depletion codes consider only neutron reactions and assume that charged particles, which may be generated in these reactions, deposit their energy locally without undergoing further nuclear interactions. Neglect of sequential charged-particle (x,n) reactions can lead to large underestimation in the inventories of radionuclides. PCROSS code was adopted for use with the ACAB activation code to enable calculation of the effects of (x,n) reactions upon radionuclide inventories and inventory-related indices. Activation calculations were made for Flibe (2LiF + BeF{sub 2}) coolant in the HYLIFE-II inertial fusion energy (IFE) power plant design. For pure Flibe coolant, it was found that (x,n) reactions dominate the residual contact dose rate at times of interest for maintenance and decommissioning. For impure Flibe, however, radionuclides produced directly in neutron reaction dominate the contact dose rate and (x,n) reactions do not make a significant contribution. Results demonstrate potential importance of (x,n) reactions and that the relative importance of (x,n) reactions varies strongly with the composition of the material considered. Future activation calculations should consider (x,n) reactions until a method for pre-determining their importance is established.