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Dive into the research topics where Kathryn A. McCarthy is active.

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Featured researches published by Kathryn A. McCarthy.


Fusion Engineering and Design | 2000

ALPS–advanced limiter-divertor plasma-facing systems

R.F. Mattas; Jean Paul Allain; R. Bastasz; J.N. Brooks; Todd Evans; A. Hassanein; S Luckhardt; Kathryn A. McCarthy; P.K. Mioduszewski; R. Maingi; E.A. Mogahed; Ralph W. Moir; Sergei Molokov; N. Morely; R.E. Nygren; Thomas D. Rognlien; Claude B. Reed; David N. Ruzic; I.N. Sviatoslavsky; D.K. Sze; M. S. Tillack; M. Ulrickson; P. M. Wade; R. Wooley; Clement Wong

The advanced limiter-divertor plasma-facing systems (ALPS) program was initiated in order to evaluate the potential for improved performance and lifetime for plasma-facing systems. The main goal of the program is to demonstrate the advantages of advanced limiter:divertor systems over conventional systems in terms of power density capability, component lifetime, and power conversion efficiency, while providing for safe operation and minimizing impurity concerns for the plasma. Most of the work to date has been applied to free surface liquids. A multi-disciplinary team from several institutions has been organized to address the key issues associated with these systems. The main performance goals for advanced limiters and divertors are a peak heat flux of \ 50 MW:m 2 , elimination of a lifetime limit for erosion, and the ability to extract useful heat at high power conversion efficiency (40%). The evaluation of various options is being conducted through a combination of laboratory experiments, www.elsevier.com:locate:fusengdes


Fusion Engineering and Design | 2000

Characterization and analysis of dusts produced in three experimental tokamaks : TFTR, DIII-D, and Alcator C-Mod

William J. Carmack; R.A. Anderl; R.J. Pawelko; G.R. Smolik; Kathryn A. McCarthy

Abstract Particulate, referred to as ‘dust’, produced during operation of tokamak systems can be a large source of activated material in a D-T fusion machine. Particulate less than 10 μm in size is easily mobilized both during an accident as well as during routine maintenance activities. We have collected and analyzed dust samples from three tokamaks to determine the potential contribution to accident source terms in future fusion power plants. We have obtained dust samples from Princeton Plasma Physics Laboratorys TFTR (prior to the final run period in 1997), from MITs Alcator C-Mod (during March and April of 1998), and from General Atomics’ DIII-D (August of 1998). This paper presents the results of our analyses for particle size distribution, specific surface area, and dust composition, including the tritium content of TFTR dust.


Fusion Engineering and Design | 2000

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

David A. Petti; Kathryn A. McCarthy; N.P. Taylor; C.B.A Forty; R.A. Forrest

Abstract 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.


Fusion Engineering and Design | 1998

Collection and analysis of particulate from the DIII-D Tokamak

William J. Carmack; Kathryn A. McCarthy; David A. Petti; Arnold G. Kellman; C.P.C. Wong

Abstract Particulate (i.e. Tokamak dust) has been collected from the DIII-D Tokamak located at General Atomics in San Diego, CA, USA. Two methods were used to collect particulate with the goal of preserving the particle size distribution and physical characteristics of the particulate. Vacuum collection on substrates and adhesion removal with metallurgical replicating tape were chosen as non-intrusive sampling methods. Sampling was completed in four toroidal locations of the machine; the 0–90° area, the 90–180° area, the 180–270° area, and the 270–360° area. The 0° direction designates the north side of the DIII-D machine. Four tiles in each area were sampled; the first vertical tile above the floor on the centerpost, a floor tile next to the center post, a tile on the isolated ring, and the first tile up from the isolated ring on the divertor. In addition, samples were collected from underneath the floor tiles at the 275 and the 75° locations. The count median diameter of the samples ranged from 0.50 to 0.86 μm, with geometric standard deviations ranging from 2.0 to 3.5. Diameters of average mass calculated from Hatch–Choate equations based upon log normal distribution characteristics ranged from 2.8 to 7.5 μm.


Journal of Fusion Energy | 1997

Steam-Chemical Reactivity Studies for Irradiated Beryllium

R. A. Anderl; R. J. Pawelko; M. A. Oates; G.R. Smolik; Kathryn A. McCarthy

This paper reports the results of an experimental study to determine the influence of neutron irradiation effects on the chemical reactivity of beryllium exposed to steam. The study entailed measurements of the following: (1) swelling of irradiated Be specimens annealed at temperatures ranging from 450°C to 1200°C, (2) hydrogen generation rates for unirradiated Be control specimens exposed to steam at temperatures from 450°C to 1200°C, and (3) hydrogen generation rates and tritium mobilization rates for irradiated Be exposed to steam at temperatures from 450°C to 700°C. For irradiated Be, swelling occurred at temperatures above 600°C and it increased to about 56% for an anneal temperature of 1200°C. Tritium and 4He were released concurrently from specimens that were annealed at 800°C and above. Steam-Be reactivity measurements for the control specimens were consistent with previous work at temperatures above 700°C, and the new measurements extended the reactivity database down to 450°C. Steam-reactivity measurements for irradiated Be were comparable to control specimens for 600°C and below, but, they indicated a significant enhancement in the chemical reactivity at 700°C.


Fusion Technology | 2000

Progress in U.S. fusion safety and environmental activities over the last decade

David A. Petti; Kathryn A. McCarthy

Abstract Magnetic fusion energy has the potential for superior safety and environmental (S&E) characteristics relative to other energy options, which is one of the main reasons for developing fusion power. Excellent progress has been made in understanding the nature of the S&E concerns associated with fusion power and in demonstrating the S&E potential of fusion. Over the past 10 yr, U.S. fusion S&E activities have been largely focused on the International Thermonuclear Experimental Reactor (ITER). The design of ITER is such that the hazards addressed are similar to those of a future fusion power plant; hence, many of the safety issues addressed by ITER are relevant to commercial fusion power plants. This paper reviews the progress and accomplishments in fusion S&E activities performed largely in support of ITER over the past decade and discusses future directions in fusion safety design criteria development and implementation; characterization of the radioactive and hazardous materials in fusion and the potential energy sources that could mobilize those materials during an accident; integrated state-of-the-art safety and risk analysis tools, methods, and results; and development of environmental design criteria for radioactive and hazardous fusion waste minimization as well as the evaluation of recycle/reuse potential of fusion materials.


Fusion Engineering and Design | 1997

Chemical reactivity and mobilization of beryllium exposed to steam

Kathryn A. McCarthy; G.R. Smolik; R.A. Anderl; R.J. Pawelko; M.A. Oates; R.S. Wallace

Abstract Beryllium is used in many fusion reactor designs as either an armor for plasma facing surfaces, or as a neutron multiplier in the blanket. Beryllium used in a water-cooled design poses important safety issues related to the chemical reactivity of beryllium in steam and its toxicity. The Fusion Safety Program at the Idaho National Engineering and Environmental Laboratory has been investigating experimentally the chemical reactivity and mobilization of various forms of beryllium for the past 6 years. In this paper we present a summary of this work, including results from fully dense (irradiated and non-irradiated), plasma-sprayed, and 88% dense beryllium. Assembling this data helps us to assess where further testing is needed. Our data help guide designs such that accident temperatures stay below values necessary to ensure beryllium release limits and hydrogen generation limits are met.


Fusion Engineering and Design | 2001

Future directions in US fusion safety and environmental activities

David A. Petti; Kathryn A. McCarthy

Abstract Over the last 30 years, the US fusion program has been a program with a focused energy mission. In recent years, the mission has changed to one that is centered on advancing plasma science, fusion science and fusion technology — the knowledge base needed for an economically and environmentally attractive fusion energy source. In this paper, we discuss the future directions of US fusion safety and environmental R&D and design support activities that are being taken in response to this change in overall mission.


Other Information: PBD: 1 Jan 2000 | 2000

Oxidation, volatilization, and redistribution of molybdenum from TZM alloy in air

G.R. Smolik; David A. Petti; Kathryn A. McCarthy; Stanley Thomas Schuetz

The excellent high temperature strength and thermal conductivity of molybdenum-base alloys provide attractive features for components in advanced magnetic and inertial fusion devices. Refractory metal alloys react readily with oxygen and other gases. Oxidized molybdenum in turn is susceptible to losses from volatile molybdenum trioxide species, MoO{sub 3}(m), in air and the hydroxide, MoO{sub 2}(OH){sub 2}, formed from water vapor. Transport of radioactivity by the volatilization, migration, and re-deposition of these volatile species during a potential accident involving a loss of vacuum or inert environment represents a safety issue. In this report the authors present experimental results on the oxidation, volatilization and re-deposition of molybdenum from TZM in flowing air between 400 and 800 C. These results are compared with calculations obtained from a vaporization mass transfer model using chemical thermodynamic data for vapor pressures of MoO{sub 3}(g) over pure solid MoO{sub 3} and an expression for the vapor pressures of MoO{sub 2}(OH){sub 2} from the literature. Calculations correlate well with experimental data.


Fusion Engineering and Design | 1994

Safety analyses of the ARIES tokamak reactor designs

J. Stephen Herring; Kathryn A. McCarthy; Thomas J. Dolan

Abstract The ARIES design effort has sought to maximize the environmental and safety advantages of fusion through careful selection of materials and careful design. Three goals are that the reactor achieve inherent or passive safety, that no public evacuation plan be necessary, and that the waste be disposable as 10CFR61 class C waste. The ARIES-1 tokamak reactor design consists of an SiC composite structure for the first wall and blanket, cooled by 10 MPa helium. The breeder is Li 2 ZrO 3 , although Li 2 O and Li 4 SiO 4 were also considered. The divertor consists of SiC composite tubes coated with 2 mm tungsten. Owing to the minimal afterheat of this blanket design, loss-of-cooling accident (LOCA) calculations indicate maximum temperatures will not cause damage if the plasma is promptly extinguished. The ARIES-II design includes liquid lithium and vanadium, both of which have low activation, multiple barriers between the lithium and air and an inert cover gas to prevent lithium-air reactions. The ARIES-II reactor is passively safe with a total 1 km early dose of about 88 rem (0.88 Sv). This dose is the result of a full-scale lithium fire resulting from a LOCA with air ingress. ARIES-III was an extensive examination of the viability of a D– 3 He fueled tokamak power reactor. Because neutrons are produced only through side reactions (D + D → 3 He + n, and D + D → T + p followed by D + T → 4 He + n), the reactor has the significant advantages of reduced activation of the first wall and shield, low afterheat and class A or C low level waste disposal. Since no tritium is required for operation, no lithium-containing breeding blanket is necessary. We modeled a LOCA in which the organic coolant was burning in order to estimate the amount of radionuclides released from the first wall. Because the maximum temperature is low, below 600°C, release fractions are small. We analyzed the disposition of the 20 g per day of tritium that is produced by D-D reactions and removed by the vacuum pumps. The ARIES-IV reactor has been designed for low activation and low stored energy. The coolant is helium and the breeder is lithium oxide. The structure is silicon carbide. Since the neutron multiplier, beryllium metal, is combustible, releasing about 60 MJ kg −1 , beryllium is the chief source of chemical energy. Less than 10% of the 24 Na inventory is likely to diffuse out of the SiC during a fire in which the beryllium is consumed. Therefore, the offsite dose would be less than 200 rem.

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David A. Petti

Idaho National Laboratory

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G.R. Smolik

Idaho National Laboratory

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Claude B. Reed

Argonne National Laboratory

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D.K. Sze

Argonne National Laboratory

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E.A. Mogahed

University of Wisconsin-Madison

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I.N. Sviatoslavsky

University of Wisconsin-Madison

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M. S. Tillack

University of California

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N. Morely

University of California

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