Fusion Power Plant Design Research Articles

Loss of Coolant Accident and Loss of Flow Accident Analysis of the ARIES-AT Design

Loss of coolant accident (LOCA) and loss of flow accident (LOFA) analysis is performed for ARIES-AT, an advanced fusion power plant design (1000 MWe). ARIES-AT employs a high performance, high temperature blanket system. It uses the high temperature SiC/SiC for structural material and LiPb for coolant-breeder. Due to the large difference between the time scale of plasma shutdown and the coolant or power loss, it is assumed that the plasma is immediately quenched at the onset of the LOCA/LOFA and the chamber components' temperature begins to rise due to the decay heat generated. A 2-D transient finite element model is established to examine the thermal behavior of the in-vessel components to determine the maximum temperature reached, the time, and duration of the peak. The model is axisymmetric in (r-z) around the reactor axis to show the details of temperature distribution in the vertical direction. The vacuum vessel is assumed adiabatic in the inboard side and radiates to the maintenance port located on the outboard side. The maximum temperature of steel in the reactor is about (600 °C - 700°C) after about 4 days from the onset of the accident. The highest temperature in the reactor is in the divertor region and it reaches ≈1050°C after about 2-3 hours. The analysis indicates that the reactor does not need any special scheme for decay heat removal.

High-temperature carbon-irradiation issues for the Sombrero ICF reactor

In order to assess the feasibility of carbon materials for the first-wall of the Sombrero KrF laser-driven ICF fusion reactor [W.R. Meier et al., OSIRIS and SOMBRERO Inertial Fusion Power Plant Designs, W.J. Schafer Associates Report, WSJA-92-01, DOE/ER/54100-1 (1992); W.R. Meier, et al., Fus. Eng. Des. 25 (1994) 145–157; I.N. Sviatoslavsky, et al., Fus. Tech. 21 (1992) 1470–1474], published experimental results relating to mechanical and thermal properties of graphites and carbon-fiber-composites (CFC's) under neutron irradiation and high heat loads are reviewed. Results are compared to published design requirements for the Sombrero ICF reactor, with particular attention to three separate issues of concern: (1) Erosion rates of the first wall are highly sensitive to the thermal conductivity value, which is itself environment-sensitive (radiation and high temperature). Erosion rates at the first wall are calculated here using a high-temperature post-irradiation conductivity value of 50 W/m K, with complete erosion of the first wall layer predicted within 1.25 full-power years (FPY), illustrating the sensitivity of erosion rates to thermal conductivity assumptions; (2) Radiation-induced swelling in 2-D and ‘pseudo 3-D’ CFC's is consistently ∼20% under high-temperature neutron damage of 5 dpa (0.33 FPY). This level of swelling would pose technical challenges to the engineering of the target chamber modules; (3) Total tritium retention is predicted to be ∼0.5–5 kg in the Sombrero chamber within 0.67 FPY, which may call into question safety-status assumptions of the CFC-based chamber design. These results indicate the urgency of high-temperature neutron-irradiation tests of fully symmetric 3-D CFC's in order to support the plausibility of a carbon first-wall IFE chamber such as proposed for Sombrero.

Energetic-economic analysis of inertial fusion plants with tritium commercial production

The realization of nuclear power plants based on fusion principles is expected to be, at the moment, very expensive. As a result the expected cost of electricity (COE) of fusion power plants is much higher than the COE of fission and fossil power plants. Thus it is necessary to study new solutions for fusion power plant designs to reduce the COE. An interesting solution for the first generation of fusion plants is to produce a surplus of tritium for commercial purposes. The present paper is concerned with the study of whether such a tritium surplus production can improve the plant economic balance, so that the COE is reduced, and to what extent. The result was that such a production allows a considerable reduction of COE and seems to be a good direction for development for the first generation of fusion power plants. To give an example, for a reference inertial confinement fusion (ICF) power plant the rise of the plant net tritium breeding ratio (TBR n) from 1 to 1.2 would allow, in the conservative estimate of a tritium market price ( C T) of 5 M$/kg, a COE reduction of about 20%. In the estimate of a TBR n rise from 1 to 1.3 and of a C T value of 10 M$/kg, COE reduction could be more than 50%! In conclusion, the present paper points out the influence of TBR increase on COE reduction. Such a conclusion, which holds true for every fusion plant, is much more valid for ICF plants in which it is possible to reach higher TBR values and to use tritium extraction systems easily. Thus, considering the relevant economic advantages, a commercial tritium surplus production should not be disregarded for first generation fusion power plant designs, in particular for ICF plant designs.

Re-evaluation of the use of low activation materials in waste management strategies for fusion

The world fusion programs have had a long goal that fusion power stations should produce only low level waste and thus not pose a burden for the future generations. However, the environmental impact of waste material is determined not only by the level of activation, but also the total volume of activated material. Since a tokamak power plant is large, the potential to generate a correspondingly large volume of activated material exists. The adoption of low activation materials, while important for reducing the radiotoxicity of the most active components, should be done as part of a strategy that also minimizes the volume of waste material that might be categorized as radioactive, even if lower in level. In this paper we examine different fusion blanket and shield designs in terms of their ability to limit the activation of the large vessel/ex-vessel components (e.g. vacuum vessel, magnets) and we identify the trends that allow improved in-vessel shielding to result in reduced vessel/ex-vessel activation. Recycling and clearance are options for reducing the volume of radioactive waste in a fusion power plant. Thus, the performance of typical fusion power plant designs with respect to recycling and clearance criteria are also assessed, to show the potential for improvement in waste volume reduction by careful selection of materials’ combinations. We discuss the impact of these results on fusion waste strategies and on the development of fusion power in the future.

SEAFP-2 bounding accident analyses

Analyses have been performed of the potential consequences to the public of hypothetical loss-of-coolant accidents in conceptual fusion power plant designs. In order to establish upper bounds to the consequences of such events, a case has been studied in which total loss of all active cooling has been assumed, with no remedial intervention for the duration of the accident sequence. The analyses are based on three conceptual power plant designs, two of them similar to those assumed in the earlier safety and environmental assessment of fusion power (SEAFP) study (Raeder et al., 1995), with updating of assumed structural materials. The three models studied provide a broad range of design options. In all cases the decay-heat driven temperature transients are well below the level at which structural melting would begin. Based on conservative assumptions, mobilisation, release and dose calculations show that potential maximum doses to the public are very far below the levels at which evacuation might be considered.

Loss of Coolant Accident (LOCA) Analysis of the ARIES-ST Design

Loss of coolant accident (LOCA) analysis is performed for ARIES-ST. ARIES-ST is a fusion power plant design based on the spherical tokamak concept to serve as a commercial power plant. It is assumed that the plasma is immediately quenched at the onset of the LOCA and the chamber components begin to increase in temperature due to the afterheat generated. This analysis examines the thermal behavior of the in-vessel components to determine the maximum temperature reached and addresses various schemes of afterheat removal. The thermal behavior of the reactor following a LOCA is simulated using a transient two-dimensional finite element model.

Engineering design of the ARIES-RS power plant

ARIES-RS is a fusion power plant design study which has examined the ability of an advanced tokamak-based power plant to compete with future energy sources and play a significant role in the future energy market. A steady-state reversed shear tokamak currently appears to offer the best combination of good economic performance and physics credibility for a tokamak-based power plant. In this article, the engineering design features of the ARIES-RS power core are described, with an emphasis on features adopted to meet the top-level requirements, on design integration advances, and on the key issues for further research and development.

The ARIES-RS power core—recent development in Li/V designs

The ARIES-RS fusion power plant design study is based on reversed-shear (RS) physics with a Li/V (lithium breeder and vanadium structure) blanket. The reversed-shear discharge has been documented in many large tokamak experiments. The plasma in the RS mode has a high beta, low current, and low current drive requirement. Therefore, it is an attractive physics regime for a fusion power plant. The blanket system based on Li/V has high temperature operating capability, good tritium breeding, excellent high heat flux removal capability, long structural life time, low activation, low after heat and good safety characteristics. For these reasons, the ARIES-RS reactor study selected Li/V as the reference blanket. The combination of attractive physics and attractive blanket engineering is expected to result in a superior power plant design. This paper summarizes the power core design of the ARIES-RS power plant study.

Inertial Fusion Energy: A Clearer View of the Environmental and Safety Aspects

A consistent method for the comparison of environmental, safety, and health (ES&H) characteristics of fusion power plant designs is developed. Such comparisons enable identification of trends in fusion ES&H characteristics and can be used to increase the likelihood of fusion achieving its full potential with respect to ES&H characteristics.A large number of radiological hazard indexes are defined in three different categories of hazard: accidents, occupational and routine exposures, and waste disposal. Using a consistent set of computer codes, data libraries, and assumptions, these radiological hazard indexes are calculated and compared for five inertial and two magnetic fusion energy power plant designs.The results of the analysis are threefold: The utility of low-activation materials (LAMs) is confirmed, the tremendous potential for the use of a thick-liquid system for first-wall protection is validated, and the strong need for materials that can last for the lifetime of the power plant is shown.The conservative radionuclide release fractions that are used show that all but one of the designs has only a minute chance of producing any early fatalities during an accident. The need for remote maintenance in most designs is confirmed, and the possibility of disposing of most fusion wastes via shallow land burial (if the methodology of current regulations is applied to fusion wastes) is shown.The need for more research in LAMs and for the experimental measurement of radionuclide release fractions under accident conditions is emphasized.

A Comparison of Steady-State ARIES and Pulsed PULSAR Tokamak Power Plants

The multi-institutional ARIES team has completed a series of four steady-state and two pulsed cost-optimized conceptual designs of commercial tokamak fusion power plants. The level of assumed advances in technology and physics was varied from one design to the next. The cost benefits of various design options are compared quantitatively with an emphasis placed on pulsed versus steady-state operation. Possible means to improve the economic competitiveness of fusion are suggested.

Improvements to the HYLIFE-II Inertial Fusion Power Plant Design

If the present research program is successful, heavy-ion beams can be used to ignite targets and to produce high gain for yields of about 350 MJ. HYLIFE-II is a power plant design based on surrounding such targets with thick liquid (Flibe, Li{sub 2}BeF{sub 4}) so that the chamber and other apparatus can not only stand up to these 350 MJ bursts of energy but do so without replacing components during the plant`s 30-year life. The capacity factor will be increased and the cost of component replacement will be decreased. Continuous improvements to the design are being made to increase safety, decrease the generation of radioactive material, and reduce the cost of electricity (COE). Improvements discussed in this paper decreased COE for each effect by the amount in parentheses: increased plant size (22%), increased capacity factor and reduced component replacement (20%), reduced remote maintenance equipment (3.2%), use of non-nuclear grade chamber, pumps and piping (2.9%), reduced tritium inventory by a factor of 2.4, reduced excess tritium production with attendant increase energy release in the blanket (1.8%), corrected treatment of Flibe inventory costs (3.4%).

Inertial fusion commercial power plants

This presentation discusses the motivation for inertial fusion energy, a brief synopsis of five recently-completed inertial fusion power plant designs, some general conclusions drawn from these studies, and an exmaple of an IEE hydrogen synfuel plant to suggest that future fusion studies consider broadening fusion use to low-emission fuels production as well as electricity.

Osiris and SOMBRERO inertial fusion power plant designs–summary, conclusions, and recommendations

Abstract An 18 month study to evaluate the potential of inertial fusion energy (IFE) for electric power production has been completed. The primary objective of the study was to provide the US Department of Energy with an evaluation of the potential of inertial fusion for electric power production. The study included the conceptual design of two inertial fusion power plants. Osiris uses an induction linac heavy ion beam driver, and SOMBRERO uses a krypton fluoride laser driver. Conceptual designs were completed for the reactors, power conversion and plant facilities, and drivers. Environmental and safety aspects, technical issues, technology development needs, and economics of the final point designs were assessed and compared. This paper summarizes the results and conclusions of the conceptual designs and results of the assessment studies. We conclude that IFE has the potential of producing technically credible designs with environmental, safety, and economics characteristics that are just as attractive as magnetic fusion. Realizing this potential will require additional research and development on target physics, chamber design, target production and injection systems, and drivers.

Hylife-II Inertial Fusion Energy Power Plant Design

The HYLIFE-II inertial fusion power plant design study uses a liquid fall, in the form of jets, to protect the first structural wall from neutron damage, x rays, and blast to provide a 30-y lifetime. HYLIFE-I used liquid lithium. HYLIFE-II avoids the fire hazard of lithium by using a molten salt composed of fluorine, lithium, and beryllium (Li2BeF4) called Flibe. Access for heavy-ion beams is provided. Calculations for assumed heavy-ion beam performance show a nominal gain of 70 at 5 MJ producing 350 MJ, about 5.2 times less yield than the 1.8 GJ from a driver energy of 4.5 MJ with gain of 400 for HYLIFE-I. The nominal 1 GWe of power can be maintained by increasing the repetition rate by a factor of about 5.2, from 1.5 to 8 Hz. A higher repetition rate requires faster re-establishment of the jets after a shot, which can be accomplished in part by decreasing the jet fall height and increasing the jet flow velocity. In addition, although not adequately considered for FIYLIFE-I, there is liquid splash that must be forcibly cleared because gravity is too slow, at higher repetition rates than 1 Hz. Splash removal is accomplished in the central region by oscillating jet flows. The cost of electricity is estimated to be 0.09 $/kW-h in constant 1988 dollars, about twice that of future coal and light water reactor nuclear power. The driver beam cost is about one-half the total cost, that is, a zero cost driver would give a calculated cost of electricity of 0.045 $/kWh.

HYLIFE-II Inertial Confinement Fusion Reactor Design

The HYLIFE-II inertial fusion power plant design study uses a liquid fall, in the form of jets to protect the first structural wall from neutron damage, x-rays, and blast to provide a 30-y lifetime. HYLIFE-I used liquid lithium. HYLIFE-II avoids the fire hazard of lithium by using a molten salt composed of fluorine, lithium, and beryllium (Li{sub 2}BeF{sub 4}) called Flibe. Access for heavy-ion beams is provided. Calculations for assumed heavy-ion beam performance show a nominal gain of 70 at 5 MJ producing 350 MJ, about 5.2 times less yield than the 1.8 GJ from a driver energy of 4.5 MJ with gain of 400 for HYLIFE-I. The nominal 1 GWe of power can be maintained by increasing the repetition rate by a factor of about 5.2, from 1.5 to 8 Hz. A higher repetition rate requires faster re-establishment of the jets after a shot, which can be accomplished in part by decreasing the jet fall height and increasing the jet flow velocity. Multiple chambers may be required. In addition, although not considered for HYLIFE-I, there is undoubtedly liquid splash that must be forcibly cleared because gravity is too slow, especially at high repetition rates. Splash removal can be accomplishedmore » by either pulsed or oscillating jet flows. The cost of electricity is estimated to be 0.09$/kW{center dot}h in constant 1988 dollars, about twice that of future coal and light water reactor nuclear power. The driver beam cost is about one-half the total cost. 12 refs., 9 figs., 5 tabs.« less

Creep Comparison between Type 316 Stainless Steel and Nimonic PE-16 Used as the First Wall in a Fusion Reactor

Damage caused by neutrons in the first wall of a fusion reactor is studied in detail for two different structural materials (Type 316 stainless steel and nimonic PE-16) in a fusion power plant design. The first-wall neutron spectrum is calculated by the ANISN code, and a dosimetric analysis is performed for both materials. Moreover, a complete study of damage (swelling and creep) during reactor life is done for type 316 stainless steel. On the basis of 1% creep strain, the lifetime of a type 316 stainless steel first wall is calculated to be {approximately} 2 yr, compared with the {approximately} 22-yr lifetime obtained for nimonic in the power plant design. In this way, the economic advantage of a superalloy, compared to stainless steel, has been confirmed.

Economic Comparison of Fusion Power Plant Designs

Over the past 10 yr, a number of studies have been developed for fusion power plants of various types (tokamaks, mirrors, etc.) complete with figures of merit such as cost estimates and estimates of the cost of generating electricity (COE). Each of these designs involves unresolved physics and engineering problems which, it is assumed, will eventually be worked out. Because of such uncertainties the figures of merit associated with such designs are not to be compared as absolute measures of worth but as relative indicators of progress within a given concept type. As part of Grumman's involvement in fusion energy development, an effort has been undertaken to compare economic indicators from the referenced studies in order to determine the cost trend in recent reactor design activities. Comparisons of the economic figures of merit for the several fusion reactor designs in the reference documents involved converting all cost estimates to the same dollar type and deciding on a common figure for construction time, ratio of direct to indirect costs, interest during construction, fixed rate of annual payment, etc. Various combinations of these factors were explored, and differences between selected values and referenced values are discussed. Also comparisons were made of themore » direct cost elements among the various designs to show there is essential agreement on balance-of-plant cost. 10 refs., 2 figs.« less

A Methodology for Cost/Benefit Safety Analysis for Fusion Reactors

AbstractA methodology is presented that can be used to determine if a proposed fusion power plant design modification, directed at improving plant safety, is cost-effective. Both normal and accident conditions can be handled. An approach for evaluating the maximum justified spending on safety is outlined. The incremental costs involved with a dose reduction measure are identified, and models for their assessment are given. By comparing the spending on the design modification to the justified expenditure ceiling, the cost-effectiveness of the design can be assessed.The utility of this approach is illustrated through two examples. For normal plant conditions, the cost-effectiveness of replacing the steel alloy PCA by low-activation silicon carbide (SiC) in the STARFIRE design is assessed. Based on a specified set of assumptions, it was determined that if the installed cost of SiC components is less than $110/kg, then the low-activation design is cost-effective.The second example illustrates the applicabilit...

An Overview of Inertial Fusion Reactor Design

Recent progress in the conceptual design of inertial fusion reaction chambers and power plants is reviewed. A discussion of expected operating parameters and a brief historical perspective are provided to organize the rich array of chamber and driver concepts. The technical feasibility of several reaction chamber concepts is discussed, along with technical issues that require future analysis, experiment, and development. Where these chambers have been integrated into a power plant design, the characteristics are described. Finally, requirements on the future development of inertial fusion reactor technology are discussed.

Nuclear fusion: Focus on Tokamak: U.S. engineers and physicists will team to attempt commercial demonstration by the year 2000

The general aspects of fusion power systems as well as some approaches to their development, are addressed from the viewpoints of engineering and technology. Focusing on a conceptual fusion power plant design, topics such as structure, materials and maintenance as well as environmental, safety, and economic factors are dealt with. Still in common with all such large-scale undertakings that try to project into the future, fusion power has built-in uncertainties. In this case, they are primarily in the area of plasma physics performance. Within the next five to six years, however, these uncertainties should be resolved and a national commitment could be made to a plan for demonstrating commercial feasibility in this century with a reasonable expenditure of money.