The Fusion Nuclear Science Facility (FNSF) is examined here as part of a two step program from ITER to commercial power plants. This first step is considered mandatory to establish the materials and component database in the real fusion in-service environment before proceeding to larger electricity producing facilities. The FNSF can be shown to make tremendous advances beyond ITER, toward a power plant, particularly in plasma duration and fusion nuclear environment. A moderate FNSF is studied in detail, which does not generate net electricity, but does reach the power plant blanket operating temperatures. The full poloidal Dual Coolant Lead Lithium (DCLL) blanket is chosen, with alternates being the Helium Cooled Lead Lithium (HCLL) and Helium Cooled Ceramic Breeder/Pebble Bed (HCCB/PB). Several power plant relevant choices are made in order to follow the philosophy of targeted technologies. Any fusion core component must be qualified by fusion relevant neutron testing and highly integrated non-nuclear testing before it can be installed on the FNSF in order to avoid the high probability of constant failures in a plasma-vacuum system. A range of missions for the FNSF, or any fusion nuclear facility on the path toward fusion power plants, are established and characterized by several metrics. A conservative physics strategy is pursued to accommodate the transition to ultra-long plasma pulses, and parameters are chosen to represent the power plant regime to the extent possible. An operating space is identified, and from this, one point is chosen for further detailed analysis, with R=4.8m, a=1.2m, IP=7.9 MA, BT=7.5T, βN<2.7, n/nGr=0.9, fBS=0.52, q95=6.0, H98 ∼1.0, and Q=4.0. The operating space is shown to be robust to parameter variations. A program is established for the FNSF to show how the missions for the facility are met, with a He/H, a DD and 5 DT phases. The facility requires ∼25years to complete its DT operation, including 7.8 years of neutron production, and the remaining spent on inspections and maintenance. The DD phase is critical to establish the ultra-long plasma pulse lengths. The blanket testing strategy is examined, and shows that many sectors have penetrations for heating and current drive (H/CD), diagnostics, or Test Blanket Modules (TBMs). The hot cell is a critical facility element in order for the FNSF to perform its function of developing the in-service material and component database. The pre-FNSF R&D is laid out in terms of priority topics, with the FNSF phases driving the time-lines for R&D completion. A series of detailed technical assessments of the FNSF operating point are reported in this issue, showing the credibility of such a step, and more detailed emphasis on R&D items to pursue. These include nuclear analysis, thermo-mechanics and thermal-hydraulics, liquid metal thermal hydraulics, transient thermo-mechanics, tritium analysis, maintenance assessment, magnet specification and analysis, materials assessments, core and scrape-off layer (SOL)/divertor plasma examinations.

The paper describes research needs in primary R&D areas for the family of dual-coolant lead-lithium (DCLL) blankets. Associated key scaling parameters are introduced and evaluated under conditions of FNSF, ITER and DEMO and also for the existing non-fusion MHD facilities, using the MaPLE loop at UCLA as an example. Comparisons among these parameters are recommended for measuring the R&D progress on the pathway from the present experimental facilities to FNSF. Possible experiments both in the existing facilities and FNSF are discussed along with the flow diagnostics.

In the U.S., the Fusion Nuclear Science Facility (FNSF) is viewed as an essential element of the fusion developmental roadmap. The tritium self-sufficiency, blanket testing, and materials testing are of particular interest since they define a critical element of the FNSF mission. There is a definitive need to breed the majority of, if not all, the tritium required for operation. A staged blanket testing strategy has been developed to test and enhance the blanket performance during each phase of operation. A materials testing module is critically important to include in FNSF to test large specimens of future generations of materials (for blanket, divertor, magnets, etc.) in relevant fusion environment. In this strategy, the test modules play a pivotal role and serve as “forerunners” for more advanced versions of blanket and materials that will validate their characteristics and features to assure the successful operation of DEMO and advanced power plants.

ARIES-ACT1 engineering design efforts were devoted to developing a credible configuration that allows for rapid removal of full-power core sectors followed by disassembly in hot cells during maintenance. The power core evolved with the main objective of achieving high performance while maintaining attractive design features and credible configuration, maintenance, and fabrication processes. To achieve high availability and maintainability of a fusion power plant, the power core components of a sector, including inboard and outboard first wall (FW)Iblankets, upper and lower divertors, and structural ring or high-temperature shield, were integrated into one replacement unit to minimize time-consuming handling inside the plasma chamber. As with the ARIES-AT design, the FW/blanket design was based on Pb-17Li as coolant and breeder, and low-activation SiC/SiC as structural material; however, the Pb-17Li mass flow rate control, flow path, FW and blanket cooling channels, coolant access pipes, and blanket structural configuration have been revised and improved to provide about the same thermal performance (similar to 58% thermal efficiency) while keeping the magnetohydrodynamic pressure drop and pumping power, material temperature, and stresses at an acceptable level. Helium-cooled W or W-alloy divertor concepts were developed to accommodate a peak surface heat flux up to similar to 14 MW/m(2). They include a smaller finger-based divertor and a midsized T-tube and larger plate-type divertor concepts, which take advantage of a simple configuration, and the smaller number of plate units and joints in a power plant. The two-zone divertor concept, with the combination of a finger-based divertor and plate-type divertor, was selected and integrated into the ARIES-ACT1 power core. The fingers are used to accommodate the designed peak heat flux of similar to 13 MW/m(2), while the plate-type divertor is used for the lower heat flux region. The overall power core configuration and system integration, as well as the definitions of major power core components, such as the FW/blankets, divertor, structural ring, and the vacuum vessel, are described here and the main design features are highlighted. Sector maintenance operations have been investigated and motion demonstrations for removing the power core sectors have been performed using state-of-the-art three-dimensional CAD to analyze the clearances and spaces in all directions. The maintenance sequence and procedure for removing the replacement unit from the plasma chamber to the hot cell for exchange and refurbishment are also discussed in this paper.

The breeding blanket with integrated first wall (FW) is the key nuclear component for power extraction, tritium fuel sustainability, and radiation shielding in fusion reactors. The ITER device will address plasma burn physics and plasma support technology, but it does not have a breeding blanket. Current activities to develop "roadmaps" for realizing fusion power recognize the blanket/FW as one of the principal remaining challenges. Therefore, a central element of the current planning activities is focused on the question: what are the research and major facilities required to develop the blanket/FW to a level which enables the design, construction and successful operation of a fusion DEMO? The principal challenges in the development of the blanket/FW are: (1) the Fusion Nuclear Environment - a multiple-field environment (neutrons, heat/particle fluxes, magnetic field, etc.) with high magnitudes and steep gradients and transients; (2) Nuclear Heating in a large volume with sharp gradients - the nuclear heating drives most blanket phenomena, but accurate simulation of this nuclear heating can be done only in a DT-plasma based facility; and (3) Complex Configuration with blanket/first wall/divertor inside the vacuum vessel - the consequence is low fault tolerance and long repair/replacement time.These blanket/FW development challenges result in critical consequences: (a) non-fusion facilities (laboratory experiments) need to be substantial to simulate multiple fields/multiple effects and must be accompanied by extensive modeling; (b) results from non-fusion facilities will be limited and will not fully resolve key technical issues. A DT-plasma based fusion nuclear science facility (FNSF) is required to perform "multiple effects" and "integrated" experiments in the fusion nuclear environment; and (c) the Reliability/Availability/Maintainability/Inspectability (RAMI) of fusion nuclear components is a major challenge and is one of the primary reasons why the blanket/FW will pace fusion development toward a DEMO.This paper summarizes the top technical issues and elucidates the primary challenges in developing the blanket/first wall and identifies the key R&D needs in non-fusion and fusion facilities on the path to DEMO. Published by Elsevier B.V.

The proposed Fusion Nuclear Science Facility (FNSF) represents the first facility to enter the complex fusion nuclear regime, and its technical mission and attributes are being developed. The FNSF represents one part of the fusion energy development pathway to the first commercial power plant with other major components being the pre-FNSF research and development, research in parallel with the FNSF, pre-DEMO research and development, and the demonstration power plant (DEMO). The Fusion Energy Systems Studies group is developing the technical basis for the FNSF in order to provide a better understanding of the demands on the fusion plasma and fusion nuclear science programs.

Water has both advantages and disadvantages as a coolant in conceptual designs of future fusion power plants. In the United States, water has not been chosen as a fusion power core coolant for decades. Researchers in other countries continue to adopt water in their designs, in some cases as the leading or sole candidate. In this article, we summarize the technical challenges resulting from the choice of water coolant and the differences in approach and assumptions that lead to different design decisions amongst researchers in this field.

Tokamak power plants are studied with advanced and conservative design philosophies in order to identify the impacts on the resulting designs and to provide guidance to critical research needs. Incorporating updated physics understanding, and using more sophisticated engineering and physics analysis, the tokamak configurations have developed a more credible basis compared to older studies. The advanced configuration assumes a self-cooled lead lithium (SCLL) blanket concept with SiC composite structural material with 58% thermal conversion efficiency. This plasma has a major radius of 6.25 m, a toroidal field of 6.0 T, a q95 of 4.5, a βN of 5.75, H98 of 1.65, n/nGr of 1.0, and peak divertor heat flux of 13.7 MW/m. The conservative configuration assumes a dual coolant lead lithium (DCLL) blanket concept with ferritic steel structural material and helium coolant, achieving a thermal conversion efficiency of 45%. The plasma major radius is 9.75 m, a toroidal field of 8.75 T, a q95 of 8.0, a βN of 2.5, H98 of 1.25, n/nGr of 1.3, and peak divertor heat flux of 10 MW/m. The divertor heat flux treatment with a narrow power scrape-off width has driven the plasmas to larger major radius. Edge and divertor plasma simulations are targeting a basis for high radiated power fraction in the divertor, which is necessary for solutions to keep the peak heat flux in the range of 10-15 MW/m. Combinations of the advanced and conservative approaches show intermediate sizes. A new systems code using a database approach has been used and shows that the operating point is really an operating zone with some range of plasma and engineering parameters and very similar costs of electricity. Papers in this issue provide more detailed discussion of the work summarized here.

ARIES-ACT2 is a conventional tokamak power plant conceptual design that uses a dual-coolant lead-lithium (DCLL) blanket concept with a RAFS (reduced-activation ferritic steel) first-wall (FW) and blanket structure. The design concept is the first fully integrated study of the DCLL blanket in a tokamak power plant. The major engineering efforts were to develop a credible configuration that can meet aggressive maintenance goals and achieve high availability and maintainability; to design a DCLL blanket that can meet tritium breeding requirements with reasonable helium and Pb-17Li cooling schemes to remove the surface and volumetric thermal power in the blanket while keeping the helium pressure drop, magneto-hydrodynamic (MHD) pressure drop, and total pumping power low, and material temperatures and stresses at an acceptable level; to design manifolding and access pipes to connect/disconnect the inboard and outboard blanket sectors to the ring headers located underneath the reactor without affecting maintenance operations and creating major MHD effects when feeding all the Pb-17Li He mass flow. Detailed three-dimensional finite element analysis of the DCLL blankets together with design iterations have been performed to finalize and optimize the major design parameters of the FW and blanket structure. The helium-cooled W plate-type divertor concept was adopted and integrated into the ACT2 DCLL power core to accommodate the peak surface heat flux of similar to 10 MW/m(2) predicted by edge plasma physics.

ARIES-ACT1 power plant has been designed and configured to allow for rapid removal of full power core sectors followed by disassembly in hot cells during maintenance operation. To achieve high availability and maintainablity of a fusion power plant, power core components of a sector, including inboard and outboard FW/blankets, upper and lower divertor, structural ring or high temperature shield were integrated into one replacement unit to minimize time comsuming handling inside plasma chamber. In this paper, the overall power core configuration and system integration, as well as the definitions of major power core components are described and main design features are highlighted.

A fusion pilot plant study was initiated to clarify the development needs in moving from ITER to a first of a kind fusion power plant, following a path similar to the approach adopted for the commercialization of fission. The pilot plant mission encompassed component test and fusion nuclear science missions plus the requirement to produce net electricity with high availability in a device designed to be prototypical of the commercial device. Three magnetic configuration options were developed around this mission: the advanced tokamak (AT), spherical tokamak (ST) and compact stellarator (CS).With the completion of the study and separate documentation of each design option a question can now be posed; how do the different designs compare with each other as candidates for meeting the pilot plant mission? In a pro/con format this paper will examine the key arguments for and against the AT, ST and CS magnetic configurations. Key topics addressed include: plasma parameters, device configurations, size and weight comparisons, diagnostic issues, maintenance schemes, availability influences and possible test cell arrangement schemes.

ARIES-ACT1 is an advanced tokamak power plant conceptual design that utilizes SiC composite structural material in the blanket and PbLi as the tritium breeder and coolant. This design concept represents an evolutionary step from ARIES-AT, which has guided tokamak research programs for the past decade. In conjunction with a helium Brayton power cycle, the high primary coolant outlet temperature allows thermal conversion efficiency of 58%. The self-cooled blanket and He-cooled W-alloy divertor provide the ability to survive relatively high power density with acceptable projected lifetime. In ARIES-ACT1, we attempted to add "robustness" to the design point without major sacrifices in performance. In this paper, we will discuss the main features of the power core and selected details in the design and analysis.

Research has been conducted to find the optimal steel to use in the vacuum vessel (VV) of ARIES power plants. The VV should meet several design criteria, including activation and fabrication requirements. Seven different types of steel were examined in order to determine which steel would be the best candidate for the ARIES VV. The main concerns are related to activation, properties under irradiation, and fabrication of a sizable VV. Steels generating high-level waste (such as 316-SS) were excluded from possible material choices. As a VV material, there is the necessity for a carefully controlled the post-weld-heat-treatment at similar to 750 degrees C after assembly, welding, and rewelding. For this particular reason, the F82H FS is not suitable for the ARIES VV. The newly developed 3Cr-3WV bainitic FS meets the activation requirements and has the potential to satisfy the fabrication requirements for the ARIES VV. It is recommended for further consideration because of several advantages over other candidate steels.

There is a strong indication that the dual-cooled LiPb blanket is the preferred concept for many fusion power plants being designed around the world. The ability of the blanket to provide tritium self-sufficiency is among the important issues that we investigated in detail for ARIES-ACT to pinpoint the design elements that degrade the breeding the most, using state-of-the-art neutronics codes. A novel stepwise approach was developed to identify the exact cause of the degradation in the tritium breeding ratio (TBR), initially 1.8 for an ideal system, reaching 1.05 for a practical design. More broadly, this paper gives many insights into the impact that internal components of the blanket as well as essential parts of a tokamak can have on the TBR and the more damaging or enhancing conditions or changes to the breeding. To overcome the challenges of dealing with all tritium-related uncertainties in several subsystems, we suggest adjusting the Li enrichment online during operation to mitigate concerns about the danger of placing the plant at risk due to tritium shortage as well as the problem of handling and safeguarding any surplus of tritium.

A number of advanced helium-cooled W-based divertor concepts have been proposed recently for fusion power plant applications within the framework of the ARIES Program. This paper summarizes design optimization and improvements of these concepts based on the minimum and maximum operating temperature of the W structure, pumping power and structural design limits. Re-evaluations of all concepts were performed with increased minimum operating temperature of the W structure from 700 ◦ C to 800

A number of advanced helium-cooled W-based divertor concepts have been proposed recently for fusion power plant applications within the framework of the ARIES Program. This paper summarizes design optimization and improvements of these concepts based on the minimum and maximum operating temperature of the W structure, pumping power and structural design limits. Re-evaluations of all concepts were performed with increased minimum operating temperature of the W structure from 700 degrees C to 800 degrees C in order to avoid embrittlement by neutron radiation. Design adjustments to allow for non-uniform heat flux profiles also have been considered. Comprehensive 3D thermal-fluid and 3D finite element thermo-mechanical analyses have been performed considering both elastic and plastic behavior and results are summarized in this paper. (C) 2012 Elsevier B.V. All rights reserved.

A potentially attractive next-step towards fusion commercialization is a pilot plant, i.e. a device ultimately capable of small net electricity production in as compact a facility as possible and in a configuration scalable to a full-size power plant. A key capability for a pilot-plant programme is the production of high neutron fluence enabling fusion nuclear science and technology (FNST) research. It is found that for physics and technology assumptions between those assumed for ITER and nth-of-a-kind fusion power plant, it is possible to provide FNST-relevant neutron wall loading in pilot devices. Thus, it may be possible to utilize a single facility to perform FNST research utilizing reactor-relevant plasma, blanket, coil and auxiliary systems and maintenance schemes while also targeting net electricity production. In this paper three configurations for a pilot plant are considered: the advanced tokamak, spherical tokamak and compact stellarator. A range of configuration issues is considered including: radial build and blanket design, magnet systems, maintenance schemes, tritium consumption and self-sufficiency, physics scenarios and a brief assessment of research needs for the configurations.

The use of tungsten as a plasma-facing material necessitates a transition joint to the structural material of the primary coolant loop at some location in order to transport the coolant to the heat exchanger. A critical issue in transition joints is the thermal expansion mismatch between materials, which can lead to unacceptably high thermal stresses. Detailed 2D and 3D analyses were performed to study the behavior of a transition from tungsten to ferritic steel (FS) with an intermediate layer of tantalum, located outside of the high heat flux region. This paper describes the results of FEM analyses including primary and secondary stresses under various time-dependent loading conditions such as warm and cold shutdown, and allowing for inelastic behaviors leading to stress relaxation and ratcheting. The results show that the transition joint satisfies the design requirement on maximum accumulated principal strain during operation.