skip to main content
OSTI.GOV title logo U.S. Department of Energy
Office of Scientific and Technical Information

Title: Overview APEX(NucSys)/ALPS(PFCs) 1996-2002 plus transition into the ITER TBM.


Abstract not provided.

Publication Date:
Research Org.:
Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)
Sponsoring Org.:
USDOE Office of Science (SC), Fusion Energy Sciences (FES) (SC-24)
OSTI Identifier:
Report Number(s):
DOE Contract Number:
Resource Type:
Resource Relation:
Conference: Proposed for presentation at the Fusion Energy Sciences Materials Workshop held July 25-29, 2016 in Oak Ridge, TN.
Country of Publication:
United States

Citation Formats

Nygren, Richard E. Overview APEX(NucSys)/ALPS(PFCs) 1996-2002 plus transition into the ITER TBM.. United States: N. p., 2016. Web.
Nygren, Richard E. Overview APEX(NucSys)/ALPS(PFCs) 1996-2002 plus transition into the ITER TBM.. United States.
Nygren, Richard E. 2016. "Overview APEX(NucSys)/ALPS(PFCs) 1996-2002 plus transition into the ITER TBM.". United States. doi:.
title = {Overview APEX(NucSys)/ALPS(PFCs) 1996-2002 plus transition into the ITER TBM.},
author = {Nygren, Richard E.},
abstractNote = {Abstract not provided.},
doi = {},
journal = {},
number = ,
volume = ,
place = {United States},
year = 2016,
month = 8

Other availability
Please see Document Availability for additional information on obtaining the full-text document. Library patrons may search WorldCat to identify libraries that hold this conference proceeding.

Save / Share:
  • A Sn-Li alloy has been identified to be a coolant/breeding material for D-T fusion applications. The key feature of this material is its very low vapor pressure, which will be very useful for free surface concepts employed in APEX, ALPS and inertial confinement fission. The vapor is dominated by lithium, which has very low Z. Initial assessment of the material indicates acceptable tritium breeding capability, high thermal conductivity, expected low tritium volubility, and expected low chemical reactivities with water and air. Some key concerns are the high activation and material compatibility issues. The initial assessment of this material, for fissionmore » applications, is presented in this paper.« less
  • In fiscal year (FY) 1998 two new fusion technology programs were initiated in the US, with the goal of making marked progress in the scientific understanding of technologies and materials required to withstand high plasma heat flux and neutron wall loads. APEX is exploring new and revolutionary concepts that can provide the capability to extract heat efficiently from a system with high neutron and surface heat loads while satisfying all the fusion power technology requirements and achieving maximum reliability, maintainability, safety, and environmental acceptability. ALPS program is evaluating advanced concepts including liquid surface limiters and divertors on the basis ofmore » such factors as their compatibility with fusion plasma, high power density handling capabilities, engineering feasibility, lifetime, safety and R and D requirements. The APEX and ALPS are three-year programs to specify requirements and evaluate criteria for revolutionary approaches in first wall, blanket and high heat flux component applications. Conceptual design and analysis of candidate concepts are being performed with the goal of selecting the most promising first wall, blanket and high heat flux component designs that will provide the technical basis for the initiation of a significant R and D effort beginning in FY2001. These programs are also considering opportunities for international collaborations.« less
  • The Department of Energy's (DOE) Savannah River Site (SRS) high-level waste program is responsible for storage, treatment, and immobilization of high-level waste for disposal. The Salt Processing Program (SPP) is the salt (soluble) waste treatment portion of the SRS high-level waste effort. The overall SPP encompasses the selection, design, construction and operation of treatment technologies to prepare the salt waste feed material for the site's grout facility (Saltstone) and vitrification facility (Defense Waste Processing Facility). Major constituents that must be removed from the salt waste and sent as feed to Defense Waste Processing Facility include actinides, strontium, cesium, and entrainedmore » sludge. In fiscal year 2002 (FY02), research and development (R&D) on the actinide and strontium removal and Caustic-Side Solvent Extraction (CSSX) processes transitioned from technology development for baseline process selection to providing input for conceptual design of the Salt Waste Processing Facility. The SPP R&D focused on advancing the technical maturity, risk reduction, engineering development, and design support for DOE's engineering, procurement, and construction (EPC) contractors for the Salt Waste Processing Facility. Thus, R&D in FY02 addressed the areas of actual waste performance, process chemistry, engineering tests of equipment, and chemical and physical properties relevant to safety. All of the testing, studies, and reports were summarized and provided to the DOE to support the Salt Waste Processing Facility, which began conceptual design in September 2002.« less
  • The ITER Conceptual Design Activity (CDA) was a three-year, 400 professional-year effort to design a next step tokamak. The Activity was conducted under the auspices of the IAEA jointly by EURATOM, Japan, the USSR and USA. The main ITER parameters are summarized in the paper. An engineering design phase (EDA) lasting 5-6 years is planned to begin in 1992. Fuel Cycle design studies carried out as part of the CDA concluded that suitable options existed or could be developed to satisfy all tritium-handling requirements for the machine within the EDA time and resource framework. During the EDA, special emphasis willmore » be required on design integration and optimization.« less
  • Components inside the ITER cryostat will become activated due to neutron irradiation such that after only a short operational period with DT fuel, maintenance will have to be done by remote means. Further remote handling operations will be required outside the cryostat, e.g. hot cells, and limited remote maintenance could be required in plant rooms adjacent to the cryostat. This paper deals, however, exclusively with maintenance operations inside the cryostat boundary. The main components inside and outside the vacuum vessel have been given a remote handling classification according to their scheduled or unscheduled maintenance requirement and their replacement has beenmore » studied. Maintenance operations inside the vacuum vessel and cryostat will in many cases be carried out under harsh environmental conditions concerning gamma flux, temperature, access and visibility and require to be verified under as close as possible simulated conditions and well rehearsed before the start of the active operations phase. The following maintenance scenarios will be reviewed, some of which require the venting of the vessel or the venting of the cryostat or both: (1) Replacement of blanket modules (2) Replacement of divertor modules (3) Replacement of vacuum pumping systems inside cryostat (4) Replacement of TF coils (require removal & re-installation of CS assembly as well as vacuum vessel sector) (5) Maintenance of PF coils, some of which are trapped underneath the vacuum vessel and support structure.« less