Overview of Safety and Environmental Issues for Inertial Fusion Energy

This paper summarizes safety and environmental issues of Inertial Fusion Energy (IFE): inventories, effluents, maintenance, accident safety, waste management, and recycling. The fusion confinement approach among inertial and magnetic options affects how the fusion reaction is maintained and which materials surround the reaction chamber. The target fill technology has a major impact on the target factory tritium inventory. IFE fusion reaction chambers usually employ some means to protect the first structural wall from fusion pulses. This protective fluid or granular bed also moderates and absorbs most neutrons before they reach the first structural wall. Although the protective fluid activates, most candidate fluids have low activation hazard. Hands-on maintenance seems practical for the driver, target factory, and secondary coolant systems; remote maintenance is likely required for the reaction chamber, primary coolant, and vacuum exhaust cleanup systems. The driver and fuel target facility are well separated from the main reaction chamber.     

End product economics and fusion research program priorities

It is shown that deuterium based fusion fuels and reactors based on them face severe technological disadvantages in comparison with fission based systems as power sources for central station electric power plants. The author postulates the most plausible deuterium based fusion reactor consistent with the physics of the fusion reaction itself and compares this reactor (called OMR-DT) with existing fission reactors. Since neutrons are the main problem in fusion, the author suggests that a great deal more effort should be given to the study of non-Maxwellian plasmas with the emphasis on neutron-free fuel cycles. The author also suggests that the deuterium based fusion driver may play its best role as a fissile fuel producer.     

Safety Assessment for Inertial Fusion Energy Power Plants: Methodology and Application to the Analysis of the HYLIFE-II and SOMBRERO Conceptual Designs

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.     

Nova target physics program and the Nova upgrade laser

Summary Recent advances in ICF target design and performance have made possible the achievement of ignition and gain with 1–2 MJ laser drive energy, as against the 5–10 MJ necessary to achieve high gain in the earlier designs. Ignition and propagating burn can be achieved at the lower energy by increasing the hohlraum temperature and, thereby increasing the pressure driving the imploding fusion capsule. Nova experiments continue to address the target physics of radiatively driven targets, such as laser-plasma interaction physics, the efficiency of laser light conversion to X-rays, hohlraum characterization and design, hydrodynamic stability, and implosion physics. Recent experiments on Nova have also demonstrated 1.3 times higher hohlraum temperature than previously predicted. This latter demonstration is the key achievement leading to the Nova Upgrade proposal. These combined results, together with those from experiments to study the interaction of high-power laser light with target plasmas, indicate that the capsule drive and symmetry conditions required for ignition and net gain can be achieved with a properly designed upgrade of the existing Nova facility.Success in the Nova Upgrade objective would firmly establish target and driver requirements for achieving high yield and high gain and would support a decision to construct a Laboratory Microfusion Facility (LMF) for defense applications and an Engineering Test Facility (ETF) for energy applications by the end of the first decade of the next century. Nova Upgrade experiments would focus on the target physics necessary to determine the minimum driver energy required to achieve ignition and high-gain laser fusion. The thermonuclear yield produced (up to 20 MJ) would be used to study the effects of fusion microexplosions on potential LMF and ETF reactor chamber materials. This information would permit development of the most efficient and least costly designs for the LMF and the ETF.In collaboration with W. H. Lowdermilk, N. Frank, C. D. Henning, John R. Murray, M. T. Tobin, J. R. Smith, E. K. Storm, J. D. Lindl, J. D. Kilkenny, J. T. Hunt, and J.B. Trenholme.     

Critical Science Issues for Direct Drive Inertial Fusion Energy

There are several topics that require resolution prior to the construction of an Inertial Fusion Energy [IFE] laboratory Engineering Test Facility [ETF]: a pellet that produces high gain; a pellet fabrication system that cost-effectively and rapidly manufactures these pellets; a sufficiently uniform and durable high repetition-rate laser pellet driver; a practical target injection system that provides accurate pellet aiming; and, a target chamber that will survive the debris and radiation of repeated high-gain pellet implosions. In this summary we describe the science issues and opportunities that are involved in the design of a successful high gain direct drive Inertial Confinement Fusion [ICF] pellet.     

Experimental results of the erosion of molybdenum due to blistering by He+ bombardment

We report some preliminary measurement of the erosion rate of plasma dumps when bombarded with 100 keV He⁺ ions at high power density ( 1 MW/m²). The expected erosion rates, based on measurements of He blistering that were made at lower power density ( 0.3 MW/m²), indicate a potentially serious problem for fusion reactors. Our tests use a reactorlike power density and produce He blisters at a rate that is slower than predicted by 2 to 4 orders of magnitude, depending on the temperature of the molybdenum target.     

Complexity And Availability for Fusion Power Plants: The Potential Advantages of Inertial Fusion Energy

Probably the single largest advantage of the inertial route to fusion energy (IFE) is the perception that its power plant embodiments could achieve acceptable capacity factors. This is a result of its relative simplicity, the decoupling of the driver and reactor chamber, and the potential to employ thick liquid walls. We examine these issues in terms of the complexity, reliability, maintainability and, therefore, availability of both magnetic and inertial fusion power plants and compare these factors with corresponding scheduled and unscheduled outage data from present day fission experience. We stress that, given the simple nature of a fission core, the vast majority of unplanned outages in fission plants are due to failures outside the reactor vessel itself. Given we must be prepared for similar outages in the analogous plant external to a fusion power core, this puts severe demands on the reliability required of the fusion core itself. We indicate that such requirements can probably be met for IFE plants. We recommend that this advantage be promoted by performing a quantitative reliability and availability study for a representative IFE power plant and suggest that databases are probably adequate for this task.     

A Plan for the Development of Fusion Energy

This is the final report of a panel set up by the U.S. Department of Energy (DOE) Fusion Energy Sciences Advisory Committee (FESAC) in response to a charge letter dated September 10, 2002 from Dr. Ray Orbach, Director of the DOE's Office of Science. In that letter, Dr. Orbach asked FESAC to develop a plan with the end goal of the start of operation of a demonstration power plant in approximately 35 years. This report, submitted March 5, 2003, presents such a plan, leading to commercial application of fusion energy by mid-century. The plan is derived from the necessary features of a demonstration fusion power plant and from the time scale defined by President Bush. It identifies critical milestones, key decision points, needed major facilities and required budgets. The report also responds to a request from DOE to FESAC to describe what new or upgraded fusion facilities will best serve our purposes over a time frame of the next twenty years.     

Management of the Service Life of a Nuclear Power Plant

Nuclear power plant service life management is the modern technology for providing efficient operation of nuclear power plants at an established safety level irrespective of the initial set service life. Consequently, this technology provides a realistic possibility for nuclear power not only to maintain but also to expand its presence in the electricity production market. Conceptually, service life management is viewed as a process that encompasses the entire life cycle of a nuclear power plant, including three basic stages: preoperation, operation, and postoperation. The technical, economic, and organizational aspects and the basic directions, including continuation and decommissioning, are discussed taking account of the international experience in such practice. The target function for managing the service life, the strategy, and a program for Russian power-generating units are examined.     

The NIKE KrF laser fusion facility

NIKE is a second generation high power KrF laser now under construction at the Naval Research Laboratory. The project is a collaborative effort between NRL and Los Alamos National Laboratory. NIKE is designed to deliver more than 2 kJ of energy to target in a 600-m focal spot and a 4-ns pulse duration. Echelon free induced spatial incoherence (ISI) will be used to produce uniform target illumination. Flat targets will be ablatively accelerated to study both Rayleigh-Taylor and parametric instabilities. These results will have direct implications to direct-drive inertial confinement fusion for commercial energy applications. Reliable operation of a high power KrF laser is also an important goal of the NIKE laser, with the objective of 1000 target shots per year. This would be an important step in the development of the KrF laser as an ICF driver. NIKE is cheduled to begin target experiments in early 1994. If successful, these experiments will provide a technical basis to proceed with construction of an ignition facility.     

Advantages of high-field tokamaks for fusion reactor development

High-field designs could reduce the cost and complexity of tokamak reactors. Moreover, the certainty of achieving required plasma performance could be increased. Strong Ohmic heating could eliminate or significantly decrease auxiliary heating power requirements and high values of n_E could be obtained in modest-size plasmas. Other potential advantages are reactor operation at modest values of , capability of higher power density and wall loading, and possibility of operation with advanced fuel mixtures. Present experimental results and basic scaling relations imply that the parameterB ²a, where B is the magnetic field and a is the minor radius, may be of special importance. A superhigh-field compact ignition experiment with very high values ofB ²a (e.g.,B ²a=150 T² m) has the potential of Ohmically heating to ignition. This short-pulse device would use inertially cooled copper plate magnets. Compact engineering test reactor and/or experimental hybrid reactor designs would use steady-state, water-cooled copper magnets and provide long-pulse operation. Design concepts are also described for demonstration/commercial reactors. These devices could use high-field superconducting magnets with 7–10 T at the plasma axis.     

The fusion-supported decentralized nuclear energy system

A decentralized nuclear energy system is proposed comprising mass-produced pressurized water reactors in the size range 10 to 300 MW (thermal), to be used for the production of process heat, space heat, and electricity in applications where petroleum and natural gas are presently used. Special attention is given to maximizing the refueling interval with no interim batch shuffling in order to minimize fuel transport, reactor downtime, and opportunity for fissile diversion. The smallest reactors could be deployed as nuclear batteries, kept in the equivalent of spent-fuel shipping casks and returned to nuclear fuel centers for refueling. These objectives demand a substantial fissile enrichment (7 to 15%). The preferred fissile fuel is U-233, which offers an order of magnitude savings in ore requirements (compared with U-235 fuel), and whose higher conversion ratio in thermal reactors serves to extend the period of useful reactivity and relieve demand on the fissile breeding plants (compared with Pu-239 fuel). Application of the neutral-beam-driven tokamak fusion-neutron source to a U-233 breeding pilot plant is examined. This scheme can be extended in part to a decentralized fusion energy system, wherein remotely located large fusion reactors supply excess tritium to a distributed system of relatively smallnonbreeding D-T reactors.     

Basic Questions Concerning the Thermophysics of Nuclear Power

The term thermophysics encompasses three areas – hydrodynamics, heat transfer, and technology of coolants. These areas are closely interrelated and influence reactor physics, corrosion processes, and the reliability and safety of a nuclear power system. At the present level of knowledge, when developing prospective fourth-generation reactors and other power systems, the thermophysical processes occurring in the loops must be considered not separately but together, taking account of their mutual influence on the operation of the loops in the nuclear power system. These processes also must be taken into account in order to develop systems for controlling them.     

The Potential for Developing and the Possibility of Attaining Radiation Equivalence between Uranium and Wastes in Future Nuclear Power

The development of nuclear power with capacity up to 300 GW using thermal reactors and BREST fast reactors with small excess breeding (BR 1.05) and nuclear power operating potentially for up to 3750 yr using depleted uranium is examined on the basis of fuel materials balances. To examine the radiation balance, all wastes, which have accumulated by the time mentioned from reprocessing of spent fuel from thermal and fast reactors are included, taking account of the running nuclide composition. The change in the potential biological danger is calculated as a function of the holding time of the entire mass of the wastes taking account of the short-lived nuclide daughter products. The possibility of starting coextraction of thorium and radium with uranium starting in 2010 and 2030 or without coextraction is taken into account. If coextraction is implemented during the periods indicated, then radiation balance of the radiactive wastes which accumulate by 2100 or 2200 is reached within a holding period of 80–120 yr. Without coextraction, the fraction of plutonium going into the wastes will have to be decreased from 0.1 to 0.01%.     

A Review of the U.S. Department of Energy's Inertial Fusion Energy Program

This is the final report of a panel set up by the U.S. Department of Energy (DOE) Fusion Energy Sciences Advisory Committee (FESAC) in response to a charge letter from Dr. Ray Orbach (Appendix A). In that letter, Dr. Orbach asked FESAC for an assessment of the present status of inertial fusion energy (IFE) research carried out in contributing programs. These programs include the heavy ion (HI) beam, the high average power laser (HAPL), and Z-Pinch drivers and associated technologies, including fast ignition (FI). This report, presented to FESAC on March 29, 2004, and subsequently approved by them (Appendix B), presents FESAC's response to that charge.     

The nuclear reactor circulation circuit as a radiation source

The solution of the problem of circulation circuits with a single radioisotope, which has been found earlier [1], is applied to the general case where several radioisotopes having radioactive progeny are formed in the substance to be activated. The problems of the absolute maximum circuit power and the consumption of neutrons per unit power for a number of elements which can be used as substances to be activated in the circuit are considered. From among them, the most promising are indium and its alloys.Special attention is paid to a circulation circuit where the substance to be activated contains a fissionable isotope (uranium circuit). It is shown that the specific power of such a circuit, all other conditions being equal, is considerably lower than the specific power of circuits with metallic indium or its alloys. As a particular case of a uranium circuit, the circulation from the reactor into the radiation unit,and the reverse,of fuel elements which have not burned up completely is considered. It is shown that, in this case, the power of the unit can be increased two- to fourfold in comparison with the power of a unit, which makes a single use of completely burned-up fuel elements.     

Reactivity Behavior During RBMK Startup

During startup of an RBMK reactor, the reactivity varies from –(4–7)_eff to 0–0.1_eff. Positive reactivity is introduced locally – by extracting control rods. Since the physical dimensions of an RBMK reactor are large, a local change in the properties can produce a large change in the spatial distribution of the neutron flux in the core. The possible range of variation of the reactivity of a subcritical and a critical reactor with one control rod extracted is analyzed for the actual states of the power-generating units of a nuclear power plant with RBMK reactors. It is shown that the extraction of some rods in an RBMK reactor in subcritical and critical states can increase the reactivity by 1_eff or more.     

A low wall-loading DEMO reactor design with high priority for early and reliable realization of a tokamak fusion reactor over the cost performance

Based on scientific databases adopted for designing ITER plasmas and on the advancement of fusion nuclear technology from the recent R&D program, a low wall-loading DEMO fusion reactor has been designed, where high priority has been given to the early and reliable realization of a tokamak fusion plasma over the cost performance. Since the major radius of this DEMO reactor is chosen to be 10 m, plasma ignition is achievable with a low fusion power of 0.8 GW and an operation period of 4–5 hours is available only with inductive current drive. The low ignition power makes it possible to adopt a first wall with an austenitic stainless steel, for which significant databases and operating experience exists, due to its use in the presence of neutron irradiation in fission reactors. In step with development of advanced materials, a step-wise increase of the fusion power seems to be feasible and realistic, because this DEMO reactor has the potential to produce a fusion power of 5 GW.     

HIBALL—A conceptual design study of a heavy-ion driven inertial confinement fusion power plant

A preliminary concept for a heavy-ion beam driven inertial confinement fusion power plant is presented. The high repetition rate of the RF accelerator driver is utilized to serve four reactor chambers alternatingly. In the chambers a novel first-wall protection scheme is used. At a target gain of 83 the total net electrical output is 3.8 GW. The recirculating power fraction is below 15%.The main goal of the comprehensive HIBALL study (which is continuing) is to demonstrate the compatibility of the design of the driver, the target and the reactor chambers. Though preliminary, the present design is essentially self-consistent. Tentative cost estimates are given. The costs compare well with those found in similar studies on magnetic confinement fusion reactors.     

A programmatic framework for the Tokamak Physics Experiment (TPX)

Significant advances have been made in the confinement of reactor-grade plasmas, so that we are now preparing for experiments at the power breakeven level in the JET and TFTR experiments. In ITER we will extend the performance of tokamaks into the burning plasma regime, develop the technology of fusion reactors, and produce over a gigawatt of fusion power. Besides taking these crucial steps toward the technical feasibility of fusion, we must also take steps to ensure its economic acceptability. The broad requirements for economically attractive tokamak reactors based on physics advancements have been set forth in a number of studies. An advanced physics data base is emerging from a physics program of concept improvement using existing tokamaks around the world. This concept improvements program is emerging as the primary focus of the U.S. domestic tokamak program, and a key element of that program is the proposed Tokamak Physics Experiment (TPX). With TPX we can develop the scientific data base for compact, continuously-operating fusion reactors, using advanced steady-state control techniques to improve plasma performance. We can develop operating techniques needed to ensure the success of ITER and provide first-time experience with several key fusion reactor technologies. This paper explains the relationships of TPX to the current U.S. fusion physics program, to the ITER program, and to the development of an attractive tokamak demonstration plant for this next stage in the fusion program.Abbreviations used TFTR Tokamak Fusion Test Reactor - JET Joint European Torus - ARIES Advanced Reactor Innovations Evaluation Study - SSTR Steady State Tokamak Reactor - PBXM Princeton Beta Experiment-Modified - DIII-D Doublet III—Dee - JT60-U Japanese Tokamak 60-Upgrade