ARIES-ST safety design and analysis

Steiner, D.; El-Guebaly, L.; Herring, Stephen; Khater, H. Y.; Mogahed, E.A.; Thayer, R.; Tillack, M. S.

doi:10.1016/s0920-3796(02)00307-1

Cited by 19 publications

(4 citation statements)

References 23 publications

Supporting

Mentioning

Contrasting

Order By: Relevance

“…One of the requirements for ARIES-ST is to limit the WDR of the TF coils to Class C waste (low-level waste qualified for shallow land burial). With a 20-cm helium-cooled, ferritic steel shield, the design life of the centerpost in ARIES-ST is 3 FPY using 10CFR61 limits and 6 FPY using Fetter limits [4]. The design life of the ferritic shield structures surrounding the plasma is 3 FPY so the centerpost would have to be replaced either every replacement or every second replacement of the ferritic steel structures, depending on which limits were used, in order to satisfy WDR requirements.…”

Section: Physical Propertiesmentioning

confidence: 99%

The toroidal field coil design for ARIES-ST

Reiersen

Dahlgren

Fan

et al. 2003

Fusion Engineering and Design

View full text Add to dashboard Cite

An evolutionary process was used to develop the toroidal field (TF) coil design for the ARIES-spherical torus (ST). Design considerations included fabricability, assembly, maintenance, energy efficiency, and structural robustness. Design options were identified early in the process. Trade studies were carried out to identify preferred choices. Design points were re-optimized based on the design choices in the TF and other systems. An attractive design for the ARIES-ST TF coil system evolved. This design addresses a number of the concerns (complexity) and criticisms (high cost, high recirculating power) of fusion. It does this by: applying advanced, but available laser forming and spray casting techniques for manufacturing the TF coil system; adopting a simple single turn TF coil system to make assembly and maintenance much easier. The single turn design avoids the necessity of using the insulation as a structural component of the TF coils, and hence, is much more robust than multi-turn designs; using a high conductivity copper alloy and modest current densities to keep the recirculating power modest. #

show abstract

Section: Physical Propertiesmentioning

confidence: 99%

The toroidal field coil design for ARIES-ST

Reiersen

Dahlgren

Fan

et al. 2003

Fusion Engineering and Design

View full text Add to dashboard Cite

show abstract

“…A PbLi breeding blanket was proposed for a number of European power plant concepts PPCS-A, PPCS-C and PPCS-D and is currently being considered for a European DEMO [135] and US ARIES power plants ARIES advanced tokamak (ARIES-AT) [2], ARIES spherical torus (ARIES-ST) (see [136,137]), ARIES compact stellarator (ARIES-CS) [138], and ARIES aggressive and conservative tokamak (ARIES-ACT) [11,12]. In China, the FDS series of power plant concepts were studied, also based on PbLi breeding blankets [139,140].…”

Section: Lead-lithiummentioning

confidence: 99%

“…To avoid, or to limit as much as possible, the spreading of these activation products in the TES, dedicated filters need to be considered in the design. Highly toxic 210 Po, Pu, 115 Cd, as well radioactivity hazardous Tritium, 137 Cs, 60 Co, 129 I, 90 Sr, and long-lived 243 Cm, Np, 99 Tc, and 209 Bi as the precursor of to the 210 Po production, are of main concern. Assessment of the overall PPCS-B breeder blanket, at the end of 5 full power years operation, reported the overall 239 Pu production is the range 3.7-9.9 kg [167] considering the realistic S-65 Be grade from Brush Wellman Co.…”

Section: Neutron Multiplying Materialsmentioning

confidence: 99%

Materials-related issues in the safety and licensing of nuclear fusion facilities

et al. 2017

View full text Add to dashboard Cite

Fusion power holds the promise of electricity production with a high degree of safety and low environmental impact. Favourable characteristics of fusion as an energy source provide the potential for this very good safety and environmental performance. But to fully realize the potential, attention must be paid in the design of a demonstration fusion power plant (DEMO) or a commercial power plant to minimize the radiological hazards. These hazards arise principally from the inventory of tritium and from materials that become activated by neutrons from the plasma. The confinement of these radioactive substances, and prevention of radiation exposure, are the primary goals of the safety approach for fusion, in order to minimize the potential for harm to personnel, the public, and the environment. The safety functions that are implemented in the design to achieve these goals are dependent on the performance of a range of materials. Degradation of the properties of materials can lead to challenges to key safety functions such as confinement. In this paper the principal types of material that have some role in safety are recalled. These either represent a potential source of hazard or contribute to the amelioration of hazards; in each case the related issues are reviewed. The resolution of these issues lead, in some instances, to requirements on materials specifications or to limits on their performance.

show abstract

“…However, oxide-dispersion-strengthened (ODS) low activation ferritic steel could be a more promising option given the potential for high temperature operation. 40,45,46 The High Average Power Laser (HAPL) Program currently leads laser facility studies for inertial fusion energy. Since the beginning of the program's activities the LLNL FEP has provided expert input in the areas of S&E. Chamber material selection for laser IFE has taken into account safety results from the SOMBRERO design accident assessment.…”

Section: Programmatic Impacts Of Safetymentioning

confidence: 99%