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Materials-engineering challenges for the fusion core and lifetime components of the fusion nuclear science facility
Nuclear Materials and Energy 16 (2018) 82–87
Contents lists available at ScienceDirect
Nuclear Materials and Energy journal homepage: www.elsevier.com/locate/nme
Materials-engineering challenges for the fusion core and lifetime components of the fusion nuclear science facility☆
T
⁎
A.F. Rowcliffea, , C.E. Kesselb, Y. Katoha, L.M. Garrisona, L. Tana, Y. Yamamotoa, F.W. Wiffena a b
Oak Ridge National Laboratory, Oak Ridge, TN, USA Princeton Plasma Physics Laboratory, Princeton, NJ, USA
A B S T R A C T
From the perspective of materials research and development (R&D) for the fusion core and near-core lifetime components of deuterium-tritium fusion power systems, the Fusion Neutron Science Facility (FNSF) concept plays a very important function by generating the complete fusion in-service environment and providing a platform for materials component-level testing. The FNSF provides the critical link between the ITER-era and the electricity- producing facilities, DEMO and the commercial power plant. The main features of the FNSF are described and the rationale presented for the selection of structural materials to meet the challenges of the power core components and also for the system lifetime components. The calculated radiation damage parameters and potential operating temperatures requirements for each of the operational phases are presented ranging from nuclear break-in up to DEMO relevant conditions. The interdependence of the FNSF and fusion nuclear materials research are discussed, and examples of near-term materials R&D activities are outlined which could address several current FNSF-related design issues.
1. Introduction The US Fusion Nuclear Science Facility (FNSF) concept [1] is envisioned as a critical intermediate step between ITER and an electricityproducing fusion demonstration power plant (DEMO). It will ultimately operate with a deuterium-tritium (D-T) plasma and is designed to achieve major advances in plasma duration and fusion nuclear environment through a phased program reaching power plant operating regimes over period of ∼30 years. Successful design, construction, licensing, and operation of any next-step fusion nuclear facility, such as FNSF or DEMO, will require solutions to the materials challenges and basic science questions discussed in [2,3]. Fusion materials programs world-wide are making significant progress towards addressing these challenges. There is a strong inter-dependence between the FNSF and development of materials and components since the FNSF itself would provide the first opportunity for assessing the performance of materials and components in the fully integrated fusion environment, a necessary step before proceeding to a US DEMO and commercial power plants. While a comprehensive approach to an integrated materials-design effort is not currently funded in the US, there are opportunities to address some of the materials-related issues identified by the on-going FNSF
design activities. Several of these intermediate term needs, and issues are discussed with focus on the first wall and blanket (FW/B), the structural ring (SR), and the vacuum vessel (VV), shown in Fig. 2. 2. The inter-dependence of the FNSF and fusion nuclear materials research The development and testing of fusion- relevant materials, ultimately for fusion power plant applications, requires a program ranging from basic material behavior to operating experience in fusion nuclear plasma confinement devices that produce the complex environment of the fusion core. Fusion materials research has already spanned a few decades, but is now entering a more critical and focused phase, evidenced by the production of large heats of reduced activation ferritic martensitic RAFM steels [4], the push to construct fusion relevant neutron sources [5,6], the examination of testing strategies in the integrated fusion environment [1,7], broader research into multiple material issues including corrosion [8–11], plasma facing requirements [12], functional material aspects [13], and recent DEMO and next step facility designs [1,14–17]. This paper emphasizes the critical need to test materials in their full-
☆ The United States Government retains and the publisher, by accepting the article for publication, acknowledges that the United States Government retains a non-exclusive, paid-up, irrevocable worldwide license to publish or reproduce the published form of this manuscript, or allow others to do so for United States Government purposes. The Department of Energy will provide public access to these results of federally sponsored research in accordance with the DOE public Access Plan (http://energy.gov/downloads/doe-public-access-plan). ⁎ Corresponding author. E-mail address: rowcliff[email protected] (A.F. Rowcliffe).
https://doi.org/10.1016/j.nme.2018.05.025 Received 29 November 2017; Received in revised form 29 March 2018; Accepted 30 May 2018 2352-1791/ © 2018 The Authors. Published by Elsevier Ltd. This is an open access article under the CC BY license (http://creativecommons.org/licenses/BY/4.0/).
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Fig. 1. The variation in dpa and appm helium in the vicinity of the first wall for the FNSF outboard DCLL blanket, (left). Thermo-mechanics analysis of a section through the inboard blanket shows the temperature variations between the first wall, inside the large Li-Pb conduits and the back of the blanket. In addition, zooming into the helium-cooled side wall of the LiPb conduit the Von Mises stress can be seen to oscillate between approximately 30 and 60 MPa, from Ref [19]. (Reprinted from Ref. [1] with permission from Elsevier).
highest fluences, but otherwise will not evaluate the simultaneous factors such as stress, hydrogen, coolant/breeder interactions or gradients. Testing in the 6-liter medium flux zone of IFMIF, will enable some level of integrated testing with limited sample numbers [20]. While the DONES-IFMIF database will be sufficient to pursue a research step such as the FNSF, a database on the materials in their component form exposed to the actual fusion environment is critical to providing the design and licensing basis for a US DEMO and commercial power plants. The integrated testing in non-nuclear environments is similarly critical to developing the confidence in component and material performance for an effective FNSF operations program. Fission reactor experience has shown that exposure to an integrated core nuclear environment can provide complex and unexpected material responses [21–23]. To meet this requirement for fusion, the FNSF provides the necessary combination of the multi-physics and neutron environments and is designed to allow the replacement of the fusion core components after each operational phase. In-depth post -irradiation examination of components will provide a unique basis for understanding and improving materials and advancing component design. In addition, a
size component form and in the complete fusion environment of a fusion core. This test program on the FNSF will be timed to follow a rigorous fusion materials R&D program supported by surrogate fusion neutron irradiation (e.g. fission reactors, multi-beam ion accelerators, spallation neutron sources), industrial-scale production of all fusion core components, non-nuclear partially integrated testing, and testing in fusion-relevant DT neutron irradiation facilities, such as the International Fusion Materials Irradiation Facility (IFMIF) [18]. The fusion core environment is complex due to the multiplicity of variables including neutrons, high temperatures, fluid pressures and flow rates, tritium, helium and solid transmutation production and corrosion interactions. Each of these variables has gradients poloidally and radially into the blanket and divertor which results in additional complexity. Fig. 1 shows the gradients in both displacement per atom (dpa) and helium production in atomic parts per million (appm) for the FNSF; temperature and stress gradients in the solid blanket structure are also shown. Fusion relevant neutron testing in facilities such as DONES (DEMO Oriented Neutron Source – IFMIF) [5,6] will provide single effects materials tests at appropriate temperatures and the
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materials test module (MTM) in one of the sectors will provide surveillance samples for the power core and for the permanent components such as the vacuum vessel (VV).
Table 1 Neutron damage (dpa), helium and hydrogen production per full power year in FNSF components based on a DCLL power plant ARIES-ACT2 [30].
3. FNSF structural materials and radiation damage parameters The FNSF project selected the Dual Coolant Lead Lithium (DCLL) [24,25] blanket concept as the primary choice and identified two backup blanket concepts, the Helium Cooled Lead Lithium (HCLL) [26] and Helium Cooled Ceramic Breeder (HCCB) [27]. These blankets share the same structural materials and the same main coolant (He), but they address the most challenging aspect of the blanket, the breeder material, and its interactions. The first two use Pb-Li liquid metal as the tritium breeder, while the third concept uses solid Li4SiO4 or Li2TiO3. The structural material chosen for the nuclear break-in phase of operation is a RAFM steel (e.g. EUROFER, F82H, or CLAM). The FNSF program envisions a progression from these first-generation steels towards the more advanced steels, (currently in the proof- of- concept development stage), for the subsequent phases. Advanced RAFMs such as the cast nano-structured alloys (CNAs), and the oxide dispersion strengthened (ODS) RAFM steels [4,28] will have microstructures designed to confer higher creep strengths and an increased capacity to sequester the transmutation-produced helium via high number densities of nano-scale dispersoids. These microstructures have the potential to mitigate helium-related property degradation mechanisms such as void swelling and intergranular embrittlement and to exhibit greater thermal stability during off-normal events. The DCLL concept relies on an electrical and thermal insulator between the Pb-Li liquid metal and the RAFM structural material in the form of a flow channel insert (FCI). The frontrunner candidate is a SiC-SiC composite [29], although several technical feasibility issues remain to be addressed including impurity effects on corrosion in flowing PbLi, the impact of MHD-related phenomena, neutron irradiation-induced changes in electrical properties and the effects of solid transmutants. Tungsten shells are located in the blanket as electrical conductors to provide stabilization of the plasma
RAFM components
Neutron damage dpa/ FPY
He generation appm/FPY
H generation appm/FPY
Powerplant: dpa/ yr / He/yr
First Wall - OB
15.3
154.6
692.0
21.0 / 275.0
Struct. Ring OB VV - OB LTS - OB First Wall - IB
0.12
0.01
0.05
0.01 0.008 13.7
0.003 0.0003 137.3
0.003 0.0012 613.5
Struct. Ring IB VV - IB LTS - IB
2.6
1.7
7.9
0.15 0.024
0.31 0.013
0.25 0.067
20.0 / 170.0
Note: The FPY for each phase are shown in Table 2.
vertical position. Fig. 2 shows these various regions in the FNSF to illustrate the location and size. The displacement damage, helium and hydrogen production rates for the FNSF components are summarized in Table 1. Some of the values shown here differ from earlier published data [31] reflecting further refinement of the analysis. Behind the blanket is a helium-cooled structural ring (SR), shown in Fig. 2, which provides the mechanical support for the blanket and divertors. The coolant channels contain shielding materials such as tungsten carbide or borated ferritic steel and the component serves as both a thermal shield and a neutron shield to protect the magnets. The SR experiences greatly reduced rates of displacement damage and helium production compared to the first wall and both RAFM and bainitic steels are being evaluated as potential structural materials [31]. The divertor is assumed to utilize a tungsten-based material both as the plasma facing armor and the structure, with interfacing to an advanced RAFM steel at the coolant inlet/outlet interface. This tungsten-steel
Fig. 2. The cross-section of the FNSF out to the toroidal and poloidal field magnets, identifying the various components in and near the fusion core. (Reprinted from Ref. [1], with permission from Elsevier). 84
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bond may require an additional interface material such as tantalum. The vacuum vessel (VV) contains the vacuum for plasma operation and forms the primary radionuclide and pressure barrier for the fusion core. The VV accumulates orders of magnitude lower damage and helium production rates than the first wall, and requires careful design to eliminate neutron streaming between sectors or modules that can lead to local spikes in the fusion neutron exposure. The leading candidate structural material is a reduced activation 3Cr-3WV bainitic steel [32], which has the potential for developing a high-toughness microstructure during post-weld cooling and eliminating the need for post-weld heat treatment (PWHT) during fabrication of the large and complex structure. The low temperature shield (LT shield), located immediately outside the VV, is a double-walled structure fabricated from a Gen1RAFM or bainitic steel. The water coolant temperature is maintained below boiling to avoid the complications of introducing a pressurized system. Although there are many other materials required throughout the FNSF, ranging from magnet superconductor, to tritium permeation windows, to thermal shields, we focus here on materials for components from the low temperature shield inward to the fusion core. The fusion nuclear environment in a D-T tokamak confinement device is complex and although the basic displacement damage from a D-T fusion neutron spectrum is similar to that of a fission neutron spectrum [30], the helium and hydrogen production rates are quite different. The various components of the fusion core, as well as the vacuum vessel and low-temperature shield, experience different levels of exposure leading to significantly different responses thus presenting an opportunity to optimize materials choices for specific locations in the system. The uncollided neutron flux at the first wall (wall load) drops from the outboard (OB) midplane to the divertor by a factor of 2.2, while on the inboard (IB) side this drop is a factor of 3.7. The maximum neutron fluxes on the divertor components are ≤0.5 times the OB peak value.
etc. are anticipated to evolve towards higher performance versions via advanced materials, innovative designs, advances in manufacturing technologies and phased component testing and replacement.
4. Basic materials pre-FNSF testing sequence; non-nuclear, fission and fusion
While a comprehensive approach to an integrated materials-design effort is not currently funded within the existing US program, there are opportunities to address some of the materials-related issues identified by the on-going FNSF design activities. Several examples related to the first wall and blanket (FW/B), the structural ring, and the vacuum vessel are discussed below.
5. Planned operations program of the FNSF The program of FNSF operation includes a He/H shakedown phase, a DD plasma pulse extension phase, and 5 phases of DT operations. The DT phases simultaneously ramp-up the neutron fluence (dpa), the operating temperatures of the blanket (and possibly also the divertor or special plasma-facing components, plasma pulse length and plasma duty cycle (which lead to longer plasma facing material exposures). The program is shown in Table 2, noting several critical parameters that describe the FNSF evolution. The last column shows the same parameters for a power plant producing 1000 MWe. The first DT blanket (Phase3), reaches only 7 dpa and 73 appm He, and provides an early experience of the integrated environment of the fusion core. The next DT phase goal of 20 dpa is considered to be the maximum capability of the current RAFM steels [2,33,34] before significant property degradation occurs. It is a required risk mitigation strategy that fully qualified advanced steels (ADV. RAFM) should be available to meet the ∼20 dpa exposure goal for Phase4. In subsequent DT phases, the displacement levels increase to 30, 40 and 80 dpa. Meeting these targets requires the availability and deployment of improved blanket materials, (ADV. RAFMs and ODS alloys), to meet the higher operating temperatures and damage levels. Although not shown in the table, the performance capability of the blanket FCIs would also have to be advanced, and tungsten-based divertor and RF launcher materials with improved radiation damage tolerance and heat flux and particle flux tolerance would also be a necessity. 6. FNSF-relevant materials near term R&D opportunities
The FNSF will deploy components (divertor, blanket, structural ring, VV) fabricated with materials developed and qualified using a range of platforms. These include non-nuclear characterization, industrial-scale material production and component fabrication technology, multiple surrogate fusion neutron testing, and fusion relevant neutron exposure testing. The FNSF study has identified two critical qualification requirements for fusion core components, before being installed on the device. Firstly, the materials in any fusion core component must have fusion relevant neutron exposure data (e.g. IFMIF) for the individual material up to the anticipated neutron fluence levels in the FNSF, and secondly, the fusion core component must demonstrate successful non-nuclear testing in as fully- integrated as possible configuration at prototypical parameters (e.g. temperature, pressure, flow rate, and magnetic fields). The vacuum vessel and many other components outside of the VV are lifetime components while the components inside the VV are replaceable and designed to accommodate a program of material advancement. The materials development program preceding the FNSF must anticipate and incorporate this material evolution into its development. Fig. 3 is a simplified diagram showing the possible component and material timelines geared to specific phases in the FNSF program. Three different blankets are deployed, progressively using more advanced RAFM steels, more advanced SiC-SiC composite flow channel inserts, operating at higher temperatures, and receiving higher levels of displacement damage and transmutation generation. These blanket components must be delivered to the FNSF in time for each operational phase. As noted in the figure, the VV is delivered at the start of the FNSF construction and is never replaced, and similarly for virtually all components outside the VV. The divertor and plasma facing special structures, including plasma heating RF launchers, diagnostic systems
6.1. Threshold helium concentrations for initiation of intergranular fracture regimes in RAFMs The FNSF wall and blanket operates at temperatures >350 °C, which is at the high end of the radiation damage-induced hardening regime where non-hardening embrittlement mechanisms driven by helium accumulation at interfaces and radiation-induced segregation/ precipitation phenomena, could become significant [35]. The threshold helium concentrations required to weaken interfaces and initiate a transition to inter-granular fracture are unknown. Further development of predictive microstructurally-based performance models is needed, supported by irradiation data derived from simultaneous production of displacement damage and implantation of helium at FNSF-relevant levels. These conditions could possibly be attained via low damage rate regimes in spallation proton facilities such as SINQ/STIP [36] or via the development of other accelerator-based facilities. 6.2. Development of Pb-Li corrosion-resistant advanced alloys There are significant uncertainties in current estimates of the impact of Pb-Li corrosion- related phenomena on the operational lifetime of the RAFM blanket structure, see [37] for a review. Information on mass transfer and precipitation of corrosion products, liquid metal MHD effects and the overarching role of radiation effects on corrosion mechanisms are largely unanswered questions. The Pb-Li corrosion issues have the potential to become a lifetime -limiting factor in the integrated 85
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Fig. 3. Timelines for the FNSF operations and the pre-FNSF R&D to deliver components to the FNSF for the 5 phases of neutron fluence and operating temperature regimes. (MTM is material test module and TBM is test blanket module).
Table 2 The baseline program for FNSF operations showing the various phases and blanket operating conditions. Phase
1 He/H
2 DD
3 DT
4 DT
5 DT
6 DT
7 DT
Power Plant DT
Years Peak wall load MW/m2 Plasma on-time, %/year FPY a Peak dpa Peak appm He Peak appm H Max blanket structure op. temp, °C He Tin/Tout (°C) PbLi Tin/Tout (°C) Blanket Structural Material
1–2
2–3
2.75 1.75 15 0.41 7.2 73 327 350–550 350/475 350/550 Gen I - RAFM
4.5 1.75 25 1.13 19.7 200 894 350–550 350/475 350/550 Gen I RAFM ADV.RAFM
5.0 1.75 35 1.75 30.6 310 1388 400–600 400/525 400/600 ADV. RAFM
6.5 1.75 35 2.28 39.8 403 1806 450–650 450/575 450/650 ADV RAFM/ODS
6.5 1.75 35 2.28–4.55 79.6 806 3612 450–650 450/575 450/650 ODS
40 FPY 2.25 85
a
15–50
<550 >350 >350 Gen I - RAFM
<550 >350 >350 Gen I - RAFM
150–200 1500–2000 6800–9100 600
ODS
Peak refers to outboard midplane at the first wall, and refers to the Fe approximation for dpa/MW-yr/m2.
RAFM, advanced RAFM, and SiC-SiC composite structures in a tungsten monoblock concept [39]. This structure would serve as the first wall section of the blanket although in principle it might also withstand divertor-like heat fluxes. Potential R&D opportunities in the area of SiCSiC composite materials to support and validate this design concept include (a) development of hermetic coatings for high pressure containment piping, (b) joining technologies for pipe geometries, and (c) the impact of neutron-irradiation-induced solid and gaseous transmutants on physical and mechanical properties. Other important areas include the development of compliant layers for the interface between steel piping and the tungsten plasma-facing elements and continued efforts to improve the materials and design of tungsten-based plasmafacing and thermo-mechanical elements.
FW/blanket environment and impact the goal of deploying advanced blankets with maximum temperatures of 600–650 °C. In parallel with the DCLL R&D program described in [37] there is a need for a riskmitigation strategy based on the development of aluminum-modified corrosion resistant materials [10,11] pursued in conjunction with the on-going development of advanced RAFM and ODS alloys with increased tolerance for radiation damage [4,28]. 6.3. Materials R&D to support FW high heat and particle flux solutions The current FNSF concept for the FW is a thin tungsten layer on the RAFM steel structure to control plasma erosion and transient high heat flux deposition [38]. An alternative concept to withstand heat fluxes >2 MW/m2 has been proposed which utilizes helium-cooled 86
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of Science, Fusion Energy Sciences. This manuscript has been authored by UT-Battelle, LLC, under Contract no. DE-AC05-00OR22725 with the US Department of Energy, and PPPL under contract DE-AC0209CH11466. The authors wish to acknowledge the members of the FNSF design team for enlightening discussions on materials-engineering issues for the FNSF.
6.4. Low dose neutron irradiation effects in 3Cr-3WV bainitic steel for the VV structure The 3Cr-3WV bainitic steel is a strong candidate material for the VV for the FNSF. The operating temperature is a trade-off between minimizing tritium permeation and minimizing radiation induced hardening. To support this selection, data are needed on radiation-induced microstructural and mechanical property changes for the 3Cr-3WV bainitic steels (including weldments) in the 250–450 °C range for FNSFrelevant VV lifetime dpa and helium generation levels. The VV also plays a critical role in safety, and tritium containment, and its successful performance during loss of coolant accidents, where temperatures can rise above normal operation and remain there for ∼100 h, is a critical factor in the qualification process for the VV structure. Relevant data could be obtained via irradiation experiments in the low damage rate and softer neutron spectrum regimes in fission test reactors to investigate irradiation behavior for lifetime neutron doses up to ∼1.0 dpa and helium levels up to ∼2 appm.
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7. Conclusions The FNSF will be a critical bridge in the development of fusion power plants by providing the fusion nuclear break-in experience for materials, components, and fusion technologies in the actual multiphysics environment. The FNSF will produce the materials performance database and demonstration of the structural integrity of key components which forms the essential basis for the engineering design, licensing and operation of DEMO and commercial power plants. The program and mission of the FNSF is designed to advance the fusion core component technologies in terms of neutron exposure, operating temperature, plasma duration/exposure, and material performance. Simultaneously the FNSF schedule dictates the R&D timeline and delivery requirements for materials and components with more advanced capability to facilitate a progressive approach to increasing power plant levels. The basic material combinations from the first wall to the vacuum vessel, together the nuclear displacement damage, helium and hydrogen production levels have been identified. Examples of near-term materials R&D which could provide valuable input to the FNSF design activity include (1) mapping helium and dpa thresholds for initiation of intergranular fracture regimes for RAFMs under FNSF-relevant conditions, (2) materials R&D to advance novel high heat and particle fluxes for the FNSF first wall, (3) development of aluminum-bearing RAFM and ODS alloys for increased corrosion resistance in flowing Pb-Li, (4) determination of the effects of neutron irradiation on 3Cr-3WV bainitic steels under simulated vacuum vessel operating conditions. Acknowledgments This research was supported by the US Department of Energy, Office
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Nuclear Materials and Energy journal homepage: www.elsevier.com/locate/nme
Materials-engineering challenges for the fusion core and lifetime components of the fusion nuclear science facility☆
T
⁎
A.F. Rowcliffea, , C.E. Kesselb, Y. Katoha, L.M. Garrisona, L. Tana, Y. Yamamotoa, F.W. Wiffena a b
Oak Ridge National Laboratory, Oak Ridge, TN, USA Princeton Plasma Physics Laboratory, Princeton, NJ, USA
A B S T R A C T
From the perspective of materials research and development (R&D) for the fusion core and near-core lifetime components of deuterium-tritium fusion power systems, the Fusion Neutron Science Facility (FNSF) concept plays a very important function by generating the complete fusion in-service environment and providing a platform for materials component-level testing. The FNSF provides the critical link between the ITER-era and the electricity- producing facilities, DEMO and the commercial power plant. The main features of the FNSF are described and the rationale presented for the selection of structural materials to meet the challenges of the power core components and also for the system lifetime components. The calculated radiation damage parameters and potential operating temperatures requirements for each of the operational phases are presented ranging from nuclear break-in up to DEMO relevant conditions. The interdependence of the FNSF and fusion nuclear materials research are discussed, and examples of near-term materials R&D activities are outlined which could address several current FNSF-related design issues.
1. Introduction The US Fusion Nuclear Science Facility (FNSF) concept [1] is envisioned as a critical intermediate step between ITER and an electricityproducing fusion demonstration power plant (DEMO). It will ultimately operate with a deuterium-tritium (D-T) plasma and is designed to achieve major advances in plasma duration and fusion nuclear environment through a phased program reaching power plant operating regimes over period of ∼30 years. Successful design, construction, licensing, and operation of any next-step fusion nuclear facility, such as FNSF or DEMO, will require solutions to the materials challenges and basic science questions discussed in [2,3]. Fusion materials programs world-wide are making significant progress towards addressing these challenges. There is a strong inter-dependence between the FNSF and development of materials and components since the FNSF itself would provide the first opportunity for assessing the performance of materials and components in the fully integrated fusion environment, a necessary step before proceeding to a US DEMO and commercial power plants. While a comprehensive approach to an integrated materials-design effort is not currently funded in the US, there are opportunities to address some of the materials-related issues identified by the on-going FNSF
design activities. Several of these intermediate term needs, and issues are discussed with focus on the first wall and blanket (FW/B), the structural ring (SR), and the vacuum vessel (VV), shown in Fig. 2. 2. The inter-dependence of the FNSF and fusion nuclear materials research The development and testing of fusion- relevant materials, ultimately for fusion power plant applications, requires a program ranging from basic material behavior to operating experience in fusion nuclear plasma confinement devices that produce the complex environment of the fusion core. Fusion materials research has already spanned a few decades, but is now entering a more critical and focused phase, evidenced by the production of large heats of reduced activation ferritic martensitic RAFM steels [4], the push to construct fusion relevant neutron sources [5,6], the examination of testing strategies in the integrated fusion environment [1,7], broader research into multiple material issues including corrosion [8–11], plasma facing requirements [12], functional material aspects [13], and recent DEMO and next step facility designs [1,14–17]. This paper emphasizes the critical need to test materials in their full-
☆ The United States Government retains and the publisher, by accepting the article for publication, acknowledges that the United States Government retains a non-exclusive, paid-up, irrevocable worldwide license to publish or reproduce the published form of this manuscript, or allow others to do so for United States Government purposes. The Department of Energy will provide public access to these results of federally sponsored research in accordance with the DOE public Access Plan (http://energy.gov/downloads/doe-public-access-plan). ⁎ Corresponding author. E-mail address: rowcliff[email protected] (A.F. Rowcliffe).
https://doi.org/10.1016/j.nme.2018.05.025 Received 29 November 2017; Received in revised form 29 March 2018; Accepted 30 May 2018 2352-1791/ © 2018 The Authors. Published by Elsevier Ltd. This is an open access article under the CC BY license (http://creativecommons.org/licenses/BY/4.0/).
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Fig. 1. The variation in dpa and appm helium in the vicinity of the first wall for the FNSF outboard DCLL blanket, (left). Thermo-mechanics analysis of a section through the inboard blanket shows the temperature variations between the first wall, inside the large Li-Pb conduits and the back of the blanket. In addition, zooming into the helium-cooled side wall of the LiPb conduit the Von Mises stress can be seen to oscillate between approximately 30 and 60 MPa, from Ref [19]. (Reprinted from Ref. [1] with permission from Elsevier).
highest fluences, but otherwise will not evaluate the simultaneous factors such as stress, hydrogen, coolant/breeder interactions or gradients. Testing in the 6-liter medium flux zone of IFMIF, will enable some level of integrated testing with limited sample numbers [20]. While the DONES-IFMIF database will be sufficient to pursue a research step such as the FNSF, a database on the materials in their component form exposed to the actual fusion environment is critical to providing the design and licensing basis for a US DEMO and commercial power plants. The integrated testing in non-nuclear environments is similarly critical to developing the confidence in component and material performance for an effective FNSF operations program. Fission reactor experience has shown that exposure to an integrated core nuclear environment can provide complex and unexpected material responses [21–23]. To meet this requirement for fusion, the FNSF provides the necessary combination of the multi-physics and neutron environments and is designed to allow the replacement of the fusion core components after each operational phase. In-depth post -irradiation examination of components will provide a unique basis for understanding and improving materials and advancing component design. In addition, a
size component form and in the complete fusion environment of a fusion core. This test program on the FNSF will be timed to follow a rigorous fusion materials R&D program supported by surrogate fusion neutron irradiation (e.g. fission reactors, multi-beam ion accelerators, spallation neutron sources), industrial-scale production of all fusion core components, non-nuclear partially integrated testing, and testing in fusion-relevant DT neutron irradiation facilities, such as the International Fusion Materials Irradiation Facility (IFMIF) [18]. The fusion core environment is complex due to the multiplicity of variables including neutrons, high temperatures, fluid pressures and flow rates, tritium, helium and solid transmutation production and corrosion interactions. Each of these variables has gradients poloidally and radially into the blanket and divertor which results in additional complexity. Fig. 1 shows the gradients in both displacement per atom (dpa) and helium production in atomic parts per million (appm) for the FNSF; temperature and stress gradients in the solid blanket structure are also shown. Fusion relevant neutron testing in facilities such as DONES (DEMO Oriented Neutron Source – IFMIF) [5,6] will provide single effects materials tests at appropriate temperatures and the
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materials test module (MTM) in one of the sectors will provide surveillance samples for the power core and for the permanent components such as the vacuum vessel (VV).
Table 1 Neutron damage (dpa), helium and hydrogen production per full power year in FNSF components based on a DCLL power plant ARIES-ACT2 [30].
3. FNSF structural materials and radiation damage parameters The FNSF project selected the Dual Coolant Lead Lithium (DCLL) [24,25] blanket concept as the primary choice and identified two backup blanket concepts, the Helium Cooled Lead Lithium (HCLL) [26] and Helium Cooled Ceramic Breeder (HCCB) [27]. These blankets share the same structural materials and the same main coolant (He), but they address the most challenging aspect of the blanket, the breeder material, and its interactions. The first two use Pb-Li liquid metal as the tritium breeder, while the third concept uses solid Li4SiO4 or Li2TiO3. The structural material chosen for the nuclear break-in phase of operation is a RAFM steel (e.g. EUROFER, F82H, or CLAM). The FNSF program envisions a progression from these first-generation steels towards the more advanced steels, (currently in the proof- of- concept development stage), for the subsequent phases. Advanced RAFMs such as the cast nano-structured alloys (CNAs), and the oxide dispersion strengthened (ODS) RAFM steels [4,28] will have microstructures designed to confer higher creep strengths and an increased capacity to sequester the transmutation-produced helium via high number densities of nano-scale dispersoids. These microstructures have the potential to mitigate helium-related property degradation mechanisms such as void swelling and intergranular embrittlement and to exhibit greater thermal stability during off-normal events. The DCLL concept relies on an electrical and thermal insulator between the Pb-Li liquid metal and the RAFM structural material in the form of a flow channel insert (FCI). The frontrunner candidate is a SiC-SiC composite [29], although several technical feasibility issues remain to be addressed including impurity effects on corrosion in flowing PbLi, the impact of MHD-related phenomena, neutron irradiation-induced changes in electrical properties and the effects of solid transmutants. Tungsten shells are located in the blanket as electrical conductors to provide stabilization of the plasma
RAFM components
Neutron damage dpa/ FPY
He generation appm/FPY
H generation appm/FPY
Powerplant: dpa/ yr / He/yr
First Wall - OB
15.3
154.6
692.0
21.0 / 275.0
Struct. Ring OB VV - OB LTS - OB First Wall - IB
0.12
0.01
0.05
0.01 0.008 13.7
0.003 0.0003 137.3
0.003 0.0012 613.5
Struct. Ring IB VV - IB LTS - IB
2.6
1.7
7.9
0.15 0.024
0.31 0.013
0.25 0.067
20.0 / 170.0
Note: The FPY for each phase are shown in Table 2.
vertical position. Fig. 2 shows these various regions in the FNSF to illustrate the location and size. The displacement damage, helium and hydrogen production rates for the FNSF components are summarized in Table 1. Some of the values shown here differ from earlier published data [31] reflecting further refinement of the analysis. Behind the blanket is a helium-cooled structural ring (SR), shown in Fig. 2, which provides the mechanical support for the blanket and divertors. The coolant channels contain shielding materials such as tungsten carbide or borated ferritic steel and the component serves as both a thermal shield and a neutron shield to protect the magnets. The SR experiences greatly reduced rates of displacement damage and helium production compared to the first wall and both RAFM and bainitic steels are being evaluated as potential structural materials [31]. The divertor is assumed to utilize a tungsten-based material both as the plasma facing armor and the structure, with interfacing to an advanced RAFM steel at the coolant inlet/outlet interface. This tungsten-steel
Fig. 2. The cross-section of the FNSF out to the toroidal and poloidal field magnets, identifying the various components in and near the fusion core. (Reprinted from Ref. [1], with permission from Elsevier). 84
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bond may require an additional interface material such as tantalum. The vacuum vessel (VV) contains the vacuum for plasma operation and forms the primary radionuclide and pressure barrier for the fusion core. The VV accumulates orders of magnitude lower damage and helium production rates than the first wall, and requires careful design to eliminate neutron streaming between sectors or modules that can lead to local spikes in the fusion neutron exposure. The leading candidate structural material is a reduced activation 3Cr-3WV bainitic steel [32], which has the potential for developing a high-toughness microstructure during post-weld cooling and eliminating the need for post-weld heat treatment (PWHT) during fabrication of the large and complex structure. The low temperature shield (LT shield), located immediately outside the VV, is a double-walled structure fabricated from a Gen1RAFM or bainitic steel. The water coolant temperature is maintained below boiling to avoid the complications of introducing a pressurized system. Although there are many other materials required throughout the FNSF, ranging from magnet superconductor, to tritium permeation windows, to thermal shields, we focus here on materials for components from the low temperature shield inward to the fusion core. The fusion nuclear environment in a D-T tokamak confinement device is complex and although the basic displacement damage from a D-T fusion neutron spectrum is similar to that of a fission neutron spectrum [30], the helium and hydrogen production rates are quite different. The various components of the fusion core, as well as the vacuum vessel and low-temperature shield, experience different levels of exposure leading to significantly different responses thus presenting an opportunity to optimize materials choices for specific locations in the system. The uncollided neutron flux at the first wall (wall load) drops from the outboard (OB) midplane to the divertor by a factor of 2.2, while on the inboard (IB) side this drop is a factor of 3.7. The maximum neutron fluxes on the divertor components are ≤0.5 times the OB peak value.
etc. are anticipated to evolve towards higher performance versions via advanced materials, innovative designs, advances in manufacturing technologies and phased component testing and replacement.
4. Basic materials pre-FNSF testing sequence; non-nuclear, fission and fusion
While a comprehensive approach to an integrated materials-design effort is not currently funded within the existing US program, there are opportunities to address some of the materials-related issues identified by the on-going FNSF design activities. Several examples related to the first wall and blanket (FW/B), the structural ring, and the vacuum vessel are discussed below.
5. Planned operations program of the FNSF The program of FNSF operation includes a He/H shakedown phase, a DD plasma pulse extension phase, and 5 phases of DT operations. The DT phases simultaneously ramp-up the neutron fluence (dpa), the operating temperatures of the blanket (and possibly also the divertor or special plasma-facing components, plasma pulse length and plasma duty cycle (which lead to longer plasma facing material exposures). The program is shown in Table 2, noting several critical parameters that describe the FNSF evolution. The last column shows the same parameters for a power plant producing 1000 MWe. The first DT blanket (Phase3), reaches only 7 dpa and 73 appm He, and provides an early experience of the integrated environment of the fusion core. The next DT phase goal of 20 dpa is considered to be the maximum capability of the current RAFM steels [2,33,34] before significant property degradation occurs. It is a required risk mitigation strategy that fully qualified advanced steels (ADV. RAFM) should be available to meet the ∼20 dpa exposure goal for Phase4. In subsequent DT phases, the displacement levels increase to 30, 40 and 80 dpa. Meeting these targets requires the availability and deployment of improved blanket materials, (ADV. RAFMs and ODS alloys), to meet the higher operating temperatures and damage levels. Although not shown in the table, the performance capability of the blanket FCIs would also have to be advanced, and tungsten-based divertor and RF launcher materials with improved radiation damage tolerance and heat flux and particle flux tolerance would also be a necessity. 6. FNSF-relevant materials near term R&D opportunities
The FNSF will deploy components (divertor, blanket, structural ring, VV) fabricated with materials developed and qualified using a range of platforms. These include non-nuclear characterization, industrial-scale material production and component fabrication technology, multiple surrogate fusion neutron testing, and fusion relevant neutron exposure testing. The FNSF study has identified two critical qualification requirements for fusion core components, before being installed on the device. Firstly, the materials in any fusion core component must have fusion relevant neutron exposure data (e.g. IFMIF) for the individual material up to the anticipated neutron fluence levels in the FNSF, and secondly, the fusion core component must demonstrate successful non-nuclear testing in as fully- integrated as possible configuration at prototypical parameters (e.g. temperature, pressure, flow rate, and magnetic fields). The vacuum vessel and many other components outside of the VV are lifetime components while the components inside the VV are replaceable and designed to accommodate a program of material advancement. The materials development program preceding the FNSF must anticipate and incorporate this material evolution into its development. Fig. 3 is a simplified diagram showing the possible component and material timelines geared to specific phases in the FNSF program. Three different blankets are deployed, progressively using more advanced RAFM steels, more advanced SiC-SiC composite flow channel inserts, operating at higher temperatures, and receiving higher levels of displacement damage and transmutation generation. These blanket components must be delivered to the FNSF in time for each operational phase. As noted in the figure, the VV is delivered at the start of the FNSF construction and is never replaced, and similarly for virtually all components outside the VV. The divertor and plasma facing special structures, including plasma heating RF launchers, diagnostic systems
6.1. Threshold helium concentrations for initiation of intergranular fracture regimes in RAFMs The FNSF wall and blanket operates at temperatures >350 °C, which is at the high end of the radiation damage-induced hardening regime where non-hardening embrittlement mechanisms driven by helium accumulation at interfaces and radiation-induced segregation/ precipitation phenomena, could become significant [35]. The threshold helium concentrations required to weaken interfaces and initiate a transition to inter-granular fracture are unknown. Further development of predictive microstructurally-based performance models is needed, supported by irradiation data derived from simultaneous production of displacement damage and implantation of helium at FNSF-relevant levels. These conditions could possibly be attained via low damage rate regimes in spallation proton facilities such as SINQ/STIP [36] or via the development of other accelerator-based facilities. 6.2. Development of Pb-Li corrosion-resistant advanced alloys There are significant uncertainties in current estimates of the impact of Pb-Li corrosion- related phenomena on the operational lifetime of the RAFM blanket structure, see [37] for a review. Information on mass transfer and precipitation of corrosion products, liquid metal MHD effects and the overarching role of radiation effects on corrosion mechanisms are largely unanswered questions. The Pb-Li corrosion issues have the potential to become a lifetime -limiting factor in the integrated 85
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Fig. 3. Timelines for the FNSF operations and the pre-FNSF R&D to deliver components to the FNSF for the 5 phases of neutron fluence and operating temperature regimes. (MTM is material test module and TBM is test blanket module).
Table 2 The baseline program for FNSF operations showing the various phases and blanket operating conditions. Phase
1 He/H
2 DD
3 DT
4 DT
5 DT
6 DT
7 DT
Power Plant DT
Years Peak wall load MW/m2 Plasma on-time, %/year FPY a Peak dpa Peak appm He Peak appm H Max blanket structure op. temp, °C He Tin/Tout (°C) PbLi Tin/Tout (°C) Blanket Structural Material
1–2
2–3
2.75 1.75 15 0.41 7.2 73 327 350–550 350/475 350/550 Gen I - RAFM
4.5 1.75 25 1.13 19.7 200 894 350–550 350/475 350/550 Gen I RAFM ADV.RAFM
5.0 1.75 35 1.75 30.6 310 1388 400–600 400/525 400/600 ADV. RAFM
6.5 1.75 35 2.28 39.8 403 1806 450–650 450/575 450/650 ADV RAFM/ODS
6.5 1.75 35 2.28–4.55 79.6 806 3612 450–650 450/575 450/650 ODS
40 FPY 2.25 85
a
15–50
<550 >350 >350 Gen I - RAFM
<550 >350 >350 Gen I - RAFM
150–200 1500–2000 6800–9100 600
ODS
Peak refers to outboard midplane at the first wall, and refers to the Fe approximation for dpa/MW-yr/m2.
RAFM, advanced RAFM, and SiC-SiC composite structures in a tungsten monoblock concept [39]. This structure would serve as the first wall section of the blanket although in principle it might also withstand divertor-like heat fluxes. Potential R&D opportunities in the area of SiCSiC composite materials to support and validate this design concept include (a) development of hermetic coatings for high pressure containment piping, (b) joining technologies for pipe geometries, and (c) the impact of neutron-irradiation-induced solid and gaseous transmutants on physical and mechanical properties. Other important areas include the development of compliant layers for the interface between steel piping and the tungsten plasma-facing elements and continued efforts to improve the materials and design of tungsten-based plasmafacing and thermo-mechanical elements.
FW/blanket environment and impact the goal of deploying advanced blankets with maximum temperatures of 600–650 °C. In parallel with the DCLL R&D program described in [37] there is a need for a riskmitigation strategy based on the development of aluminum-modified corrosion resistant materials [10,11] pursued in conjunction with the on-going development of advanced RAFM and ODS alloys with increased tolerance for radiation damage [4,28]. 6.3. Materials R&D to support FW high heat and particle flux solutions The current FNSF concept for the FW is a thin tungsten layer on the RAFM steel structure to control plasma erosion and transient high heat flux deposition [38]. An alternative concept to withstand heat fluxes >2 MW/m2 has been proposed which utilizes helium-cooled 86
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of Science, Fusion Energy Sciences. This manuscript has been authored by UT-Battelle, LLC, under Contract no. DE-AC05-00OR22725 with the US Department of Energy, and PPPL under contract DE-AC0209CH11466. The authors wish to acknowledge the members of the FNSF design team for enlightening discussions on materials-engineering issues for the FNSF.
6.4. Low dose neutron irradiation effects in 3Cr-3WV bainitic steel for the VV structure The 3Cr-3WV bainitic steel is a strong candidate material for the VV for the FNSF. The operating temperature is a trade-off between minimizing tritium permeation and minimizing radiation induced hardening. To support this selection, data are needed on radiation-induced microstructural and mechanical property changes for the 3Cr-3WV bainitic steels (including weldments) in the 250–450 °C range for FNSFrelevant VV lifetime dpa and helium generation levels. The VV also plays a critical role in safety, and tritium containment, and its successful performance during loss of coolant accidents, where temperatures can rise above normal operation and remain there for ∼100 h, is a critical factor in the qualification process for the VV structure. Relevant data could be obtained via irradiation experiments in the low damage rate and softer neutron spectrum regimes in fission test reactors to investigate irradiation behavior for lifetime neutron doses up to ∼1.0 dpa and helium levels up to ∼2 appm.
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7. Conclusions The FNSF will be a critical bridge in the development of fusion power plants by providing the fusion nuclear break-in experience for materials, components, and fusion technologies in the actual multiphysics environment. The FNSF will produce the materials performance database and demonstration of the structural integrity of key components which forms the essential basis for the engineering design, licensing and operation of DEMO and commercial power plants. The program and mission of the FNSF is designed to advance the fusion core component technologies in terms of neutron exposure, operating temperature, plasma duration/exposure, and material performance. Simultaneously the FNSF schedule dictates the R&D timeline and delivery requirements for materials and components with more advanced capability to facilitate a progressive approach to increasing power plant levels. The basic material combinations from the first wall to the vacuum vessel, together the nuclear displacement damage, helium and hydrogen production levels have been identified. Examples of near-term materials R&D which could provide valuable input to the FNSF design activity include (1) mapping helium and dpa thresholds for initiation of intergranular fracture regimes for RAFMs under FNSF-relevant conditions, (2) materials R&D to advance novel high heat and particle fluxes for the FNSF first wall, (3) development of aluminum-bearing RAFM and ODS alloys for increased corrosion resistance in flowing Pb-Li, (4) determination of the effects of neutron irradiation on 3Cr-3WV bainitic steels under simulated vacuum vessel operating conditions. Acknowledgments This research was supported by the US Department of Energy, Office
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