E.T. Cheng

Long-term Radioactive Waste from Fusion Reactors: Part II

Abstract In Part I we calculated 10 CFR 61 “Class-C” specific activity limits for all long-lived radionuclides with atomic number less than 88 (Ra). These calculations were based on the whole-body dose. We also estimated the production of these radionuclides from all naturally occurring elements with atomic numbers less than 84 (Po) in the first wall of a typical fusion reactor, and thereby derived concentration limits for these elements in first-wall materials, if the first wall is to be suitable for Class-C disposal. In Part II we use the “effective dose equivalent” (EDE), which is a much better indication of the risk from radiation exposure than the whole-body dose, to calculate specific activity limits for all long-lived radionuclides up to Cm-248. In addition, we have estimated the production of long-lived actinides and fission products from possible thorium and uranium impurities in first-wall structures. This completes our study of long-lived radionuclides that are produced from all elements that occur in the earths crust at average concentrations greater than one part per trillion.

Long-term radioactivity in fusion reactors

The specific activity limits for shallow land (“Class C”) waste disposal of all long-lived radionuclides with atomic number less than 88 have been calculated using the 10 CFR 61 methodology. These specific activity limits were used to determine the concentration limits of nearly all naturally-occurring elements in fusion reactor blanket materials. Of the elements that could be constituents of or impurities in blanket materials, aluminum, silicon, nickel, zirconium, tantalum, and tungsten were found to be limited to concentrations of 0.1 to 10%, and niobium, molybdenum, silver, gadolinium, terbium, and holmium were found to be restricted to 0.1 to 10 parts per million.

The ARIES-I Tokamak Reactor Study †

The ARIES research program is a multi-institutional effort to develop several visions of tokamak reactors with enhanced economic, safety, and environmental features. Three ARIES visions are currently planned for the ARIES program. The ARIES-I design is a DT-burning reactor based on modest extrapolation from the present tokamak physics data base; ARIES-II is a DT-burning reactor which will employ potential advances in physics; and ARIES-III is a conceptual D-3He reactor. The first design to be completed is ARIES-I, a 1000 MWe power reactor. The key features of ARIES-I are: (1) a passively safe and low environmental impact design because of choice of low activation material throughout the fusion power core, (2) an acceptable cost of electricity, (3) a plasma with performance as close as possible to present-day experimental achievements, (4) a high performance, low activation, SiC composite blanket cooled by He, and (5) an advanced Rankine power cycle as planned for near term coal-fired plants. The ARIES-I research has also identified key physics and technology areas with the highest leverage for achieving attractive fusion power system.

Initial Integration of Accident Safety, Waste Management, Recycling, Effluent, and Maintenance Considerations for Low-Activation Materials

A true “low-activation” material should ideally achieve all of the following objectives: The possible prompt dose at the site boundary from 100% release of the inventory should be <2 Sv (200 rem); ...

Radioactivity aspects of fusion reactors

Abstract Activation characteristics, including radioactivity, decay heating rate, and integrated decay energy at times after shutdown of a D—T fusion power reactor were investigated for all potential reactor materials using a recently published comprehensive activation cross-section library and decay data handbook. It was found that among the potential structural elements, the shutdown activity could vary by four orders of magnitude, with C, O, and Si producing the least radioactivity and Mo giving the highest activity within a few days after shutdown, a period of importance to the reactor operation. Vanadium, Ti and Fe are among the lower activation elements with the activity levels higher than Si by about one (for V) to two (for Ti and Fe) orders of magnitude. As far as alloying elements are concerned, Cr and Si are best for minimizing the activity level; Mn, Ni, Ta and W are among the elements giving higher radioactivity and decay heat values. These higher activity elements are furthermore subject to the neutron spectral effect resulting in an increase of activation levels in a soft spectrum with higher neutron population at lower energies. The important elements, that need to be limited in fusion reactor materials in order to meet the 10CFR61 Class C shallow-land burial disposal goal, are Al, Si, Ni, Zr and Ta as alloying elements, and Nb, Mo, Ag, Gd, Tb, and Ho as impurities. The concentration limits of some of these elements such as Nb will also become more restrictive in a soft neutron spectrum, that is typical for the present fusion experimental facilities under investigation.

The ARIES-III D-3He tokamak-reactor study

A description of the ARIES-III research effort is presented, and the general features of the ARIES-III reactor are described. The plasma engineering and fusion-power-core design are summarized, including the major results, the key technical issues, and the central conclusions. Analyses have shown that the plasma power-balance window for D-/sup 3/He tokamak reactors is small and requires a first wall (or coating) that is highly reflective to synchrotron radiation and small values of tau /sub ash// epsilon /sub e/ (the ratio of ash-particle to energy confinement times in the core plasma). Both first and second stability regimes of operation have been considered. The second stability regime is chosen for the ARIES-III design point because the reactor can operate at a higher value of tau /sub ash// tau /sub E// tau /sub E/ approximately=2 (twice that of a first stability version), and because it has a reduced plasma current (30 MA), magnetic field at the coil (14 T), mass, and cost (also compared to a first-stability D-/sup 3/He reactor). The major and minor radii are, respectively 7.5 and 2.5 m.<<ETX>>

Blanket design for the ARIES-I tokamak reactor

For the Advanced Reactor Innovation and Evaluation Study-I (ARIES-I) tokamak power reactor design, the authors evaluated two gas-cooled, low-activation ceramic blanket designs, a 5-MPa helium-cooled design, and a 0.5-MPa CO/sub 2/ gas-carried, Li/sub 4/SiO/sub 4/ particulate design. The more extensive database available for the helium-cooled option has prompted the selection of this option as the reference design. The selected ARIES-I blanket design uses SiC composite as the structural material, 5-MPa helium as coolant, Li/sub 4/SiO/sub 4/ as the solid tritium breeder, and Be metal pellets as the neutron multiplier. This combination of materials provides the design of a high nuclear performance blanket with high-outlet temperature, good neutron multiplication, and adequate tritium breeding. It is a low-activation design that satisfies the criteria for 10CFR61 Class-C shallow land waste disposal, and achieves inherent safety since it produces negligible after heat, thus virtually eliminating the possibility of exposing the public to radioactivity. The mechanical design, neutronics analysis, thermal-hydraulic analysis, power-conversion system design, tritium extraction, and safety evaluation are summarized.<<ETX>>

Introduction and synopsis of the TITAN reversed-field-pinch fusion-reactor study

Abstract The TITAN reversed-field-pinch (RFP) fusion-reactor study has two objectives: to determine the technical feasibility and key developmental issues for an RFP fusion reactor operating at high power density: and to determine the potential economic (cost of electricity), operational (maintenance and availability), safety and environmental features of high mass-power-density fusion-reactor systems. Mass power density (MPD) is defined as the ratio of net electric output to the mass of the fusion power core (FPC). The FPC includes the plasma chamber, first wall, blanket, shield, magnets, and related structure. Two different detailed designs TITAN-I and TITAN-II, have been produced to demonstrate the possibility of multiple engineering-design approaches to high-MPD reactors. TITAN-I is a self-cooled lithium design with a vanadium-alloy structure. TITAN-II is a self-cooled aqueous loop-in-pool design with 9-C ferritic steel as the structural material. Both designs use RFP plasmas operating with essentially the same parameters. Both conceptual reactors are based on the DT fuel cycle, have a net electric output of about 1000 MWe, are compact, and have a high MPD of 800 kWe per tonne of FPC. The inherent physical characteristics of the RFP confinement concept make possible compact fusion reactors with such a high MPD. The TITAN designs would meet the U.S. criteria for the near-surface disposal of radioactive waste (Class C, IOCFR61) and would achieve a high Level of Safety Assurance with respect to FPC damage by decay afterheat and radioactivity release caused by accidents. Very importantly, a “single-piece” FPC maintenance procedure has been worked out and appears feasible for both designs. Parametric system studies have been used to find cost-optimized designs. to determine the parametric design window associated with each approach, and to assess the sensitivity of the designs to a wide range of physics and engineering requirements and assumptions. The design window for such compact RFP reactors would include machines with neutron wall loadings in the range of 10–20 MW/m 2 with a shallow minimum COE at about 18 MW/m 2 . Even though operation at the lower end of the this range of wall loading (10–12 MW/m 2 ) is possible, and may be preferable, the TITAN study adopted the design point at the upper end (18 MW/m 2 ) in order to quantify and assess the technical feasibility and physics limits for such high-MPD reactors. From this work, key physics and engineering issues central to achieving reactors with the features of TITAN-I and TITAN-II have emerged.

Overview of the TITAN-I fusion-power core

The TITAN reactor is a compact (major radius of 3.9 m and plasma minor radius of 0.6 m), high neutron wall loading (~18 MW/m 2 ) fusion energy system based on the reversed-field pinch (RFP) confinement concept. The reactor thermal power is 2918 MWt resulting in net electric output of 960 MWe and a mass power density of 700 kWe/tonne. The TITAN-I fusion power core (FPC) is a lithium, self-cooled design with vanadium alloy (V-3Ti-1Si) structural material. The surface heat flux incident on the first wall is ~4.5 MW/m 2 . The magnetic field topology of the RFP is favorable for liquid metal cooling. In the TITAN-I design, the first wall and blanket consist of single pass, poloidal flow loops aligned with the dominant poloidal magnetic field. A unique feature of the TITAN-I design is the use of the integrated-blanket-coil (IBC) concept. With the IBC concept the poloidal flow lithium circuit is also the electrical conductor of the toroidal-field and divertor coils. Three dimensional neutronics analysis yields a tritium breeding ratio of 1.18 and a molten salt extraction technique is employed for the tritium extraction system. Almost every FPC component would qualify for Class C waste disposal. The compactness of the design allows the use of single-piece maintenance of the FPC. This maintenance procedure is expected to increase the plant availability. The entire FPC operates inside a vacuum tank, which is surrounded by an atmosphere of inert argon gas to impede the flow of air in the system in case of an accident. The top-side coolant supply and return virtually eliminate the possibility of a complete LOCA occurring in the FPC. The peak temperature during a LOFA is 991 °C.

Nonelectric Applications of Fusion

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. James Decker, Acting Director of the DOE Office of Science. In that letter, Dr. Decker asked FESAC to consider “whether the Fusion Energy Sciences program should broaden its scope and activities to include non-electric applications of intermediate-term fusion devices.” This report, submitted to FESAC July 31, 2003, and subsequently approved by them (Appendix B), presents FESACs response to that charge.