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ARIES Documents -- Meetings Archive

ARIES Project Meeting, 19-21 June 2000

Documented by L. Waganer

University of Wisconsin - Madison
Documented by L. Waganer

Agenda and Presentations

Action Items

Participants:
(ANL) Billone, Gohar, Hassanein, Mattas, Sze
(Boeing) Waganer
(DOE) -
(FPA) -
(GA) Goodin, Lao, Schultz
(INEEL) Petti
(LBL) Yu
(LLNL) Latkowski, Meier
(MIT) Bromberg
(NRL) Schmitt, Sethian
(PPPL) Heitzenroeder, Jardin, Kessel
(RPI) Steiner
(TSI) -
(UCSD) Mau, Miller, Najmabadi, Raffray, Tillack
(UW) A. Abdou, El-Guebaly, Haynes, Henderson, Khater, Kukcinski, Mogahed, Moses, Peterson, Wilson, Sawan, Sviatoslavsky

Ref: Agenda & Presentation Links

Administrative and General Topics

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Farrokh Najmabadi reminded the team that one of the purposes of this meeting is to finalize the operating and design parameters of the ARIES-AT facility. After the meeting, the team should commence writing the final report sections so the draft is ready for review in September. All AT work should be completed by the end of the year.

The second purpose of the meeting is to kickoff the ARIES-IFE chamber concept assessment. This meeting is to inform the team of the current technical status of the IFE driver, target, and chamber approaches and to establish the assessment approach.

Farrokh noted the Design Studies 2001 budget is unknown at present, hence specific ARIES tasks cannot be defined at this time. Bill Dove was unable to attend this meeting due to work commitments.

We tried to schedule our future meetings to secure desirable dates. The meetings will be full three-day meetings, starting on Monday morning or ending on Friday afternoon. It was recommended to hold the next meeting in late September and the following one in early to mid December. Generally a two-week window is available at both times. Les Waganer will draft a query to determine if there are any conflicts on the candidate dates. The next ARIES conference call will be on Wednesday, 19 July at 9-11 PDT. L. Waganer will obtain a call-in number.

ARIES-AT Final Design Presentations

Physics Analysis

AT Physics Overview - Steve Jardin compared the ARIES-AT physics basis to the ARIES-RS physics basis. The RS was more aggressive than prior ARIES tokamak plasma approaches but the work was well received by the fusion community and helped to establish new standards for comparison. The ARIES-ST work enhanced the modeling and predictions associated with improved plasma shaping and flux profiles. The current AT plasma analyses are more complete with respect to plasma transport, resistive wall modes, and edge temperature and density conditions. More care has been exercised to optimize the AT plasma. The 99% flux surface was used in place of the traditional 95% associated with ARIES-RS and ST plasmas. This results in higher predicted beta values. The pressure profiles used are more flexible, resulting in better bootstrap current drive profiles, higher beta values, and elimination of the higher harmonic fast wave (HHFW) current drive system. Higher triangularity and elongation AT plasmas also have improved the plasma performance, with the vertical stabilizing shell being closer to the plasma and a higher betaN. Vertical stability analyses have been conducted to examine the naturally unstable plasma modes in conjunction with active feedback coils to provide the necessary plasma stability. Lower edge density conditions have been investigated over a range of betaN values and several candidate current drive systems. Two current drive systems have been well matched for on- and off-axis current drive profiles to fit with the desired composite plasma current profile.

Summary of GA Resistive Wall Mode, Transport, Neoclassical Tearing Modes, and Divertor Analyses Lang Lao discussed key findings from the GA physics analyses on the PPPL-provided L-mode equilibrium base case. Pressure and current profiles are fitted with polynomials to computed L- and H-mode equilibria. These H-mode pressure and current profiles and X-point positions improve ballooning stability. The GATO code was used to predict that a conducting wall stabilizes low n-modes. A rotational drive ~ 0.1 WAlf is essential to keep open the stability operation window against the n = 1 ideal kink/ballooning mode and the resistive wall mode (RWM). An ICH rotational drive system would be required in the range of 50 100 MW of power. [Input from Physics Group splinter meeting: Note that this rotation requirement came from analysis of a system without active feedback stabilization. In the presence of RWM feedback coils, it is not clear that there is any requirement for driven rotation. This point will be clarified in the coming week in consultation with the NTM experts in the community; however, it appears to be born out by a paper in the recent Phys. Rev. Lett. by Liu and Bondeson.] Lang also found a skin time L/R > 10 ms is crucial to keep the power requirement for the n = 1 RWM feedback coil at a modest level. Self-consistent physics-based transport simulations indicate the optimized pressure and current profiles can be sustained with a peaked density profile.

Lang Lao suggested that adding a plasma impurity such as argon could radiate more than 50% of the total exhausted power and keep the peak heat fluxes (at the first wall?) at a manageable level (< 10 MW/m2). The team said that it had been decided to use a radiative divertor to more uniformly distribute the exhausted power to the divertor region and reduce the exhaust power to the first wall. On the ARIES-RS and ST, Tom Petrie predicted the performance of a radiative divertor and should be asked to estimate the ARIES-AT divertor performance.

Vertical Stability and PF Coil Optimization Chuck Kessel discussed the current stability and equilibrium results. A good conducting passive structure and feedback control with moderate power requirements (30 MVA) can stabilize the plasma for the n = 0 vertical instability mode (stability factor of 1.2 and a reasonable vertical growth time). The position of the AT passive stabilizing shell is closer to the plasma that has allowed higher plasma elongation and higher beta. The 4-cm thick tungsten shell is designed to operate hot and be passively cooled within the blanket and shield.

An additional shell will be required on the midplane to meet RWM requirements identified by Lao. There should be no interaction between the vertical stability shells and the RWM shell as the fields are perpendicular.

The number and position of the PF coils have been finalized and the data forwarded to Phil Heitzenroeder for mechanical coil development and analysis. Chuck explained there are 7 coils in the central solenoid stack (actually 7 x 2) plus 7 more PF coils outboard to shape the plasma (again 7 x 2). The size and currents in the outer coils are somewhat larger than optimal values to provide adequate clearance for maintenance actions.

RF Heating and Current Drive System Definition T. K. Mau presented his latest RF heating and current drive analyses assuming nominal AT plasma conditions from the latest strawman data. At least two RF systems are required to drive current on-axis and off-axis to achieve the desired plasma current profile. Previous analyses had predicted an excellent current profile fit; however, slight changes to the fields, plasma densities, and edge plasma densities have resulted in more power required to achieve a proper current fit. Chuck Kessel suggested TK define a reasonable RF system and then give the predicted current profile to C. Kessel to try to find a solution that would reasonably fit the input data. It is strongly urged that we do not incorporate either a third RF system or a neutral beam injection system for current profiling and/or plasma rotation. For the moment, the reference option is the ICRF/FW (on-axis) and LHW (off-axis). The operational power requirement is 35 MW with startup power slightly higher. Launcher designs for both systems are to be defined by T.K. Mau.

AT-Systems Analysis

Strawman Results Ron Miller reported the systems code adequately reproduced the PPPL-provided plasma and coil data. He continues to coordinate with T.K. Mau to model the RF heating and current drive power requirements. The baseline strawman remains at an aspect ratio of 4.0, Rt = 5.2 m, and BetaN = 6.0. All available plasma parameters and engineering data inputs have been incorporated into the code, including radial/vertical builds, EQDISK data, and PF coil sizes and locations. The plasma rotation power estimates have not been included. Also the cost of the vacuum vessel provided by L. Waganer during the meeting has not been incorporated. The extra cost for the high temperature power conversion system has not been included. Waganer and Miller will try to incorporate analysis and rationale for higher plant availability factors.

The net electric power for the baseline and the engineering design point is 1000 MWe, but data for higher net output powers may be reported. For the present, the COE is reported in 1992$.

Ron showed several parametric assessments for the COE metric. The major toroidal radius had a minor effect. A betaN of 6.0 was better than 5.6 or 6.8. TF and PF coils using Nb3Sn and HTS were nearly equal in cost impact. Increasing the plant net electrical output by 60% decreased the COE by 15%. Increasing the peak field had minimal impact on the overall cost.

AT- Engineering

Power Core Final Design Definition - Xueren Wang was not able to attend the meeting, but Rene Raffray and Les Waganer explained the major machine and maintenance features.

Blanket and Divertor Design Rene Raffray described his analysis and design that employed the machine and power parameters from the 3/10/00 strawman. SiC/SiC properties are consistent with the SiC/SiC Town Meeting results. Rene illustrated the coolant circuits serviced by the coaxial ring headers (I, II, and III). The general approach for the blanket segment and coolant header was explained. The maximum first wall CVD and SiC/SiC temperatures were 1009C and 996C, respectively. The maximum SiC/LiPb interface temperature at the inner wall was 994C. This analysis used a 2-D moving coordinate system. A pressure stress analysis conducted on the inner shell of the blanket module determined the stresses were less than allowable. An analysis of the thermal conditions on the counterflow annular inlet and outlet piping confirmed a LiPb inlet temperature of 653C and an outlet temperature of 1100C.

The LiPb cooled divertor was analyzed to determine the MHD effects on the coolant and the expected heat transfer capability. The analysis confirmed the divertor configuration can handle ~5 to 6 MW/m2 surface heat flux. For a 5 MW/m2 surface heat load, the 3-mm-thick tungsten is expected to remain below 1145C and the underlying SiC structure below 970C. The divertor SiC structure was also analyzed for thermal stresses. Manufacturing processes were proposed for the outer divertor plates. Rene expects to complete a 3-D structural analysis of the divertor and define more thorough blanket and divertor fabrication flow diagrams.

Liquid Jet Divertor Concept for ARIES-AT Yousry Gohar summarized the objective to develop a divertor concept for the ARIES-AT configuration and plasma conditions. Several liquid metal systems were considered with lithium, Sn-Li, and Flibe having the more attractive thermal properties for this application. It was pointed out that lithium would pose a definite safety hazard given the significant water content of the ARIES-AT power core. Results from parametric thermal analyses (jet length, jet thickness, jet velocity, and surface heat load) indicted lithium has the best thermal performance, tin-lithium is satisfactory, and flibe is deemed to be inadequate for this application. Lithium would increase the tritium recycle and would increase the safety concern in ARIES-AT. The tin-lithium metal system needs more physical and operational data to confirm the application potential.

A conceptual approach for a liquid jet divertor on the lower and upper divertor regions was presented. Several technical questions were raised. Yousry said that he would continue to work closely with the ARIES-AT team to refine the liquid jet divertor concept. Farrokh Najmabadi confirmed that if the liquid jet continued to offer promise, it would be included in the AT report as a possible divertor option.

Update of AT Vacuum Vessel Design Les Waganer described the main functions and constraints for the AT vacuum vessel design. Replacement of modular power core elements is the adopted approach, with individual doors and enclosures to gain access to the power core modules. Locking pneumatic jacks secure the door during plant operation and swing out of the way for maintenance. Steps in the door provide positive positioning of the doors and minimize neutron streaming through the door opening. A high temperature SiC/SiC wedge cooled with LiPb is located just inboard of the doorframe. Laila El-Guebaly offered to provide additional material definition to adequately shield the TF coils directly behind the frame.

Details of the vacuum vessel components were defined to assemble a bottoms-up ROM cost estimate of the welded ferritic steel structure. The estimate included the basic structure, the doors, door frames, port enclosures, and the port enclosure doors. The entire vacuum vessel system would weigh approximately 1.1 x 106 kg. The estimated direct capital cost of the vacuum system would be $39M (in 2000$) including contingency and subcontractor fees. Fabrication costs dominate, thus future work would assess more labor saving manufacturing approaches, especially involving the interior bulkheads inside the double-walled structures. The port enclosures represented a majority of the cost; hence a single walled approach will be considered. The port enclosures are considered to be a vacuum structure, therefore the cryostat could be downgraded from a vacuum structure to a simple, single-walled, containment vessel.

Material Design Limits and Issues Mike Billone summarized his recommended design guidelines for SiC/SiC as a structural and thermal transfer material in the ARIES-AT design. Given the current database for material properties, he recommended an allowable stress limit of 190 MPa for primary plus bending plus thermal stresses (assuming stresses are calculated with an isotropic elastic model), a maximum operating temperature of about 1000C, and a thermal conductivity of 35 W/mK at the beginning of life and 20 W/mK in the irradiated, end-of-life condition. The largest uncertainties associated with these numbers are the effects of helium generation on the thermal and structural performance of the SiC/SiC composite.

Lithium lead properties are based on data summarized in a German report (1986). Upper temperature limits for the properties database are in the range of 350-700C. The validity of extrapolating and correlating the properties to higher temperatures remains an unresolved issue. It was suggested that a Russian report might contain LiPb properties at higher temperatures.

Within the range of 5 to 20 dpa, ferritic steel can be used between 20C and 100C with a fracture toughness design limit of 40 MPam0.5. Preliminary estimates of the corresponding design allowable stress are in the range of 50-100 MPa. This limit is dependent on the minimum detectable flaw size during QA inspection. Bob Odette of UCSB will be consulted to recommend better stress limits.

Mike said he thought the preferred joining method of SiC to W is to spray SiC onto W, but he would investigate plasma spraying W onto SiC. The main design concern, other than the strength of the interface, is the mechanical properties of plasma-sprayed tungsten as compared to the design properties assumed in the ARIES-AT study for stress-relieved tungsten.

High Temperature Superconductor (HTS) Coil Definition Leslie Bromberg said the preferred HTS material is the YBCO highly textured tapes. Although only short tapes have been made to date, this process can be scaled to production quantities. This material performs best when the tapes are aligned parallel to field, which constrains the fabrication to certain techniques. The cooling requirement for the YBCO HTS is much reduced (higher temperature and no quench requirement.) HTS coils can operate at much higher fields than conventional superconductors, but this feature is not invoked in this design. The coil structural material will probably be Inconel 625. Coil stresses are well below the design allowables. Leslie estimates the cost will soon fall to the range of $20/kAm and ultimately to $4/kAm.

Engineering Design Overview of TF and PF Coils Phil Heitzenroeder explained the HTS TF and PF coil design features and design basis. Additional structure has been provided just inside the upper and lower TF coils to provide adequate support while enabling the large maintenance regions outboard. Details of the coil cases and supporting structure were shown. The TF coils have 9.6 MAturns per coil using a wound construction approach for proper field orientation. The power leads are introduced and extracted at the bottom of the coil. The TF coils are wound in a continuous fashion with no joints or field splices. No active cooling is required internal to the conductor, rather the HTS coils are cooled by conduction. Stress and deflection data were provided for both the TF and PF coils. Phil noted the stresses are acceptable. The coil data will have to be slightly updated to reflect the most current design data.

Vertical Stabilizer Shell Design and Tutorial on EM Pump Design Igor Sviatoslavsky described the current approach to attach vertical stabilizing shells to the second blanket element, which is also a life-of-plant component. These shells are located between the first and second blanket regions. The 4-cm-thick shells are constructed of tungsten and are passively cooled. During operation, they form a toroidally-continuous path to provide the necessary plasma vertical stability. Tungsten wool at the adjoining interfaces will provide the electrical conductivity across the sector interface gap [including the high temperature wedge].

In the discussion of the divertor plumbing and fluid pressure, Rene mentioned that it would be necessary to increase the pressure head in the upper divertor region to compensate for the static and dynamic pressure losses. Therefore a local pressure booster pump might be required. Igor described three electromagnetic (EM) pumps that might be used. The induction pump has the best system characteristics for this application.

Tritium System and Primary Loop Design Dai-Kai Sze mentioned the characteristics of the AT tritium system include both high tritium partial pressure and low tritium concentration (low inventory). The best method of tritium recovery from LiPb is a liquid-gas contactor. Helium gas will be bubbled through a 10% slipstream of LiPb. Tritium is diffused into the helium, which transports the tritium to a molecular sieve bed that removes the tritium from the helium purge stream. A palladium diffusion window will remove more tritium from the hydrogen stream. Dai-Kai will pursue the effect of tritium in the primary loop components, e.g., the turbine.

Dai-Kai discussed the selection of the materials for the primary and secondary (power conversion) loops. Most materials are not compatible with 1000-1100C LiPb. SiC is reasonably compatible with LiPb over most temperature regimes. In the high radiation zone of the power core, the SiC/SiC temperature limit of 1000C is judged to be acceptable. In the low radiation zone of the closed cycle helium coolant, it is thought a higher temperature of 1100C could be tolerated. This would allow the primary heat exchanger to use SiC/SiC as the structural material. The SiC to high temperature material joint needs to be resolved.

Final Nuclear Analyses for AT Laila El-Guebaly reported that the current blanket configuration satisfies the breeding requirement (TBR 1.1) with a 10 cm increase in the Blanket II to compensate for the addition of the tungsten vertical stabilizing and RWM shells. The blanket provides an overall energy multiplication of 1.1.

The high temperature shield and vacuum vessel are well optimized to achieve the design constraints. The design is more compact than previous ARIES designs and the HTS magnets are well protected. The vacuum vessel will not meet the criteria to allow rewelding given todays knowledge of the irradiated material properties. Therefore, if an internal power core component fails and the vacuum vessel must be removed, it will have to be replaced with a new component. Given the projected cost of the vacuum vessel provided by L. Waganer, this would not be a prohibitive expense but will increase the volume of waste, which is a concern for tokamaks. Maintenance of all the power core components is considered to be remote, with no hands-on maintenance required.

The activation analysis results indicate all components will qualify as low level waste, assuming proper impurity controls on FS component materials and purification of LiPb to lower the concentration of bismuth. Design solutions have been proposed to reduce decay heat and mitigate effects of decay heat during postulated worst case accidents. The volume of waste is small in comparison to previous ARIES designs, but the small volume results in all components having a clearance index greater than one. However, the waste material could be recycled and reused in other nuclear facilities or disposed of in near surface low level waste facilities (all components can qualify as Class C waste and most can qualify as Class A waste). The new definition of the vacuum vessel port enclosures will be integrated into the waste assessment.

Divertor Activation Analysis Ali Abdou reported the divertor activation analysis results based on the current divertor configuration, materials, and vertical build. The radioactivity level for the SiC/SiC divertor structure and manifold drops several orders of magnitude shortly after shutdown (~ 1 minute), but the tungsten coating does not significantly decrease until several decades have elapsed. The divertor region has a unique, replaceable (at 4 FPY), high temperature shield that has a lower activity level than the tungsten, but the radiation levels of the FS filler do not decrease significantly until after 50 years. The life-of-plant components show much lower levels of induced radioactivity, but their activity levels do not significantly reduce until after 50 to 100 years. Tungsten has the highest decay heat. The decay heat in SiC drops significantly after a minute. All components meet the Fetter and NRC Class C criteria and the vacuum vessel, magnets, and coil cases meet the NCR Class A limits. After 100 years, all components will meet the NRC Class A LLW criteria except for the HT shield.

LOCA and LOFA Analyses - Elsayed Mogahed described the ANSYS code used to perform the 2-D transient axisymmetric, finite-element thermodynamic LOFA and LOCA analysis of the AT power core. Due to the large difference in time scales of plasma shutdown and loss of coolant or flow, it was assumed the plasma would be shut down immediately. He used an adiabatic boundary at the inboard vacuum vessel, whereas the outboard vacuum vessel radiates to the maintenance port. Les Waganer pointed out that there is a very large mass of vacuum port enclosures (712,000 kg) in intimate contact with the outer surface of the vessel that would provide a better heat sink than radiation to a distant wall. The high temperature components usually cooled down over the time span examined (1 week), whereas the vacuum vessel continued to heat up to a peak temperature. However the vacuum vessel remained below the critical steel temperature of 700C. The most critical condition (686C in the vacuum vessel structure) occurred at 3 days after accident initiation with a LOCA in vacuum vessel water coolant and a simultaneous LOFA in LiPb, which allowed the decay heat in the LiPb to remain in the blanket and shield. Even with the present pessimistic assumptions, Elsayed concluded there are no special procedures required to address the worst LOFA and LOCA conditions.

Safety and Enviromental Update for ARIES-AT Dave Petti outlined the contents for the integrated Safety and Environmental chapter. The major objectives were to eliminate an evacuation plan (< 1 Rem) at 1 km site boundary) and to minimize the plant waste (and waste level). The main radioisotope inventories are Po in the LiPb coolant, tritium in LiPb, tritium in SiC, and activated W dust. The concentration of Po-210 in the LiPb coolant could be controlled by limiting the bismuth concentration to a level of 1 ppm that would, in turn, correspond to a Po-210 concentration of 0.1 ppb. This would equate to a total inventory of 2500 Ci. To meet the 25 Ci release limit in the case of a LiPb spill, an additional factor of 100 reduction must be provided by containment. Four separate loops would provide a factor of 4. Drain tanks would help meet the limit constraints. Dai Kai Sze has been asked to provide more specific data on loop and drain tank configurations and volumes. Additional work to be done in the near term will be on tritium retention in SiC and LiPb, assessment of LOCA impacts, and contact dose assessments.

ARIES-AT Action Items

AT Task LeadersComplete draft contents by end of September (at September meeting?)
Physics
Jardin, LaoDefine plasma rotation need or requirement
PetrieComplete divertor analysis, power distribution, and feasibility of radiative divertor
Mau Design/determine RF launcher structure, size, power requirements for startup phase
Kessel Resolve (refine) off axis seed current and profiles for Mau
Engineering
Miller, Waganer, Tillack Develop a methodology to support a higher availability in concert with AP600 proposed values
MillerMap major physics and engineering changes from RS to AT
Raffray Determine the volume (clearance) of the divertor
Raffray, Wang, El-GuebalyDetermine and show the shielding for the vacuum duct resolve one duct (UCSD) vs. two ducts (PPPL) with required conductance in ducts and behind HT shield
Waganer Convert cryostat from double-walled vacuum container to single-walled containment barrier
Heitzenroeder Transmit TF and PF coil CAD files to UCSD
" Resolve high level stresses in TF coils and toroidal displacement to coils
" Determine magnetic fields on HTS coils
" Update PF coil cross-sections
El-Guebaly, Raffray Update Blanket II thicknesses
Miller Address Engineering and configuration issues. (Ron, exactly what? radial builds, cost of shells, TF coil set compared to Kessel's recommendation)
Billone Determine method to apply a thin layer of tungsten onto SiC/SiC and estimate interface design properties
SviatoslavskyDetermine mechanical loads on vertical stability shells
Sze Determine tritium inventory and the Po (Bi) processing system details
El-Guebaly Update the blanket and shield definition and TBR
Waganer, El-Guebaly Update the high temperature wedge design
Abdou Include a 3.5 mm tungsten layer on the divertor surface in the analysis
Mogahed Update LOCA/LOFA to include kink and WRM shells and port enclosures
Kessel, Mau Define fueling injection design approaches and
All task leaders Provide data to systems code in one week
NajmabadiDistribute a final report outline in two weeks

Neutron Source Study

Don Steiner summarized the main issues of comparing neutron sources based upon fission, accelerator, and fusion drivers. Little data is available on fission neutron sources, so the best comparison was between the ATW and fusion. Don showed a table of the major system parameters. The choice of the blanket option is a very important decision that has major impact on the performance and economics of the entire system. In our study, the team examined three neutron multiplication factor levels: 5, 30 and 60.

Don has incorporated most of the teams comments into the study report draft. He is anticipating distributing the report next week to the team for review.

ARIES-IFE Study Kickoff Presentations

IFE Program Direction

Program Goals, Expectations, and Plans Farrokh Najmabadi stressed that the ARIES team has the opportunity to assess several leading chamber concepts when integrated and self-consistent with the appropriate drivers and targets. The three types of chamber concepts are dry; solid, protected wall; and thick liquid walls. Several concepts of each of these types have been proposed in the past, but an integrated assessment by a broad technology team is required. First, the dry wall will be assessed, roughly by November 2000, followed by the solid, protected wall by April 2001, and the thick liquid wall by September 2001.

The approach is not to develop and assess point designs, rather it should be to examine the critical issues and to establish credible design windows in concert with the potential drivers and targets. Many leading laboratories, universities, and industries are contributing to this study and bringing their current understanding of the system elements. The concepts will be evaluated with respect to physics, engineering, environmental, and economic parameters. Technical and programmatic risks will be identified and research and developmental needs will be proposed to mitigate identified risks.

Perspective on IFE Designs Jerry Kulcinski told the team that there have been many valuable system and conceptual design studies for IFE power plants over the years, but not since the 1980s and early 1990s. The two most recent comprehensive studies were the Osiris/Sombrero and Prometheus-L and H designs concluded in 1992. Fortunately, many of the original team members are present within the new ARIES-IFE team, and that knowledge can be applied to the current evaluation.

The major factors for IFE are the targets, drivers, and the chambers. The high cost of the driver is a dominating influence to achieve the requisite power on the target. However, once a suitable power level is achieved, the driver has the ability to achieve repetition rates to support very large or multiple chambers. Both laser and heavy ion (HI) drivers should be considered, but the light ion drivers are not presently being pursued. The veil of classification has been lifted on the targets, so physics and engineering issues can now be openly discussed and evaluated. Both direct and indirect targets should be evaluated. Many changes and improvements in IFE target design have been identified since 1990-1992.

The chamber designs from 1990 are still viable candidates. Some of the designs have been improved, but most remain to be evaluated in a self-consistent manner. In the dry wall concept, carbon-carbon has replaced woven carbon fibers. Tritium entrainment is now identified as a potential concern and constraint. New tools allow a better analysis of the opacity of the dry wall chamber gas. Ability to inject and track the target in the chamber environment is better understood. Survivability of the target also must be considered and analyzed. And pulsed radiation damage of the chamber materials is appreciated and being investigated.

Vision of Direct Drive Laser IFE Plants, Targets, and Issues John Sethian gave a brief summary of direct drive laser inertial fusion principles, advantages, and remaining challenges. Both indirect and direct drive targets may be employed with lasers. John thought direct drive (DD) was superior due to the simpler target physics, cheaper targets to fabricate on a commercial scale, more efficient coupling of the laser with the fuel, and no hohlraum debris to remove from the chamber. John discussed the effects and solutions of the laser and target surface non-uniformities that degrade the laser-to-fuel coupling. The application of KrF and diode-pumped solid state lasers (DPSSL) to IFE were described by John. NRL advanced DD target designs show promise for high gain (<100). Issues to be resolved before IFE could be a viable commercial product include: target injection, final focussing optics, and plant economics. John discussed the Integrated Research Experiment and how it would contribute to the success of IFE.

Target Systems

Summary of Target Physics Issues John Sethian provided the team a primer on high gain target design. Higher gain targets of 100 or more can be achieved by tailoring the adiabat of the fuel, zooming the laser spot size, and smoothing the laser output. Preheating the target (e.g., raising the adiabat) reduces the density, which increases the ablation velocity and lowers the instability growth rate (more stable target); but the lower density results in a lower gain. NRL has developed a "tailored adiabat" concept to preheat the ablator but keep the fuel cold and dense. One way to do this is to put an outer layer of 1 mm CH and 200Å Au on the target. The foot of the laser pulse heated the gold layer to ~70 eV, producing broadband radiation. The X-rays heat up the lower foam and DT (ablator) layers but not the inner pure DT fuel. More recent results show the gold may not be necessary. Regardless, zooming the laser focal spot to follow the collapsing target will substantially boost gains by as much as 70%. Three possible DD target designs were compared from the standpoint of target materials and layers, laser type, focal spot zooming, and target performance. ARIES should consider a spectrum of potential targets and driver features. Fast ignitor targets are a possibility for future investigation.

Target Selection Ken Schultz emphasized that the study should focus on the primary chamber issues of wall and final optic protection, blanket performance, target fabrication, and target injection and tracking. In this context, the study should explore integrated systems to determine design windows for all major elements. Ken suggested four classes of targets, including laser DD shock preheated, DD radiation preheated, and ID x-ray drive plus the HI ID close-coupled, distributed radiator. Ken listed several selection criteria for the targets: reasonable gain substantiated physics basis, suitable output spectrum, feasible fabrication process and cost, and attractive injection characteristics. Other considerations might include laser frequency, alternate materials, cost and ES&H impact, fill and layering techniques, and injection and survival.

Target Fabrication, Injection, and Tracking Dan Goodin emphasized that target injection will be a necessary and important component in all the steps to reach a commercially attractive IFE power plant. Dan reviewed the severe requirements and the strategy to inject, track, and protect the targets so they reach the correct position inside the chamber. The injection technique for ID targets was discussed and the heating during injection is thought to be negligible. Orientation of the ID target is thought to be achievable with spin stabilization. The heating of the DD targets is a more severe problem, especially on large-radius, dry-walled chambers. The heat flux on the injected target is attributed to thermal radiation and convection. It is recommended the surface heat flux be less than 1W/cm2 to prevent damage to the DT layer. A reflective outer layer (e.g., gold) is effective for reducing the thermal radiation from the blackbody wall conditions. The convection solution is more challenging, requiring tailoring the chamber environment and using special measures to protect the target.

Driver Systems

Candidate IFE Drivers (KrF and DPSSL) John Sethian summarized the two leading laser approaches for IFE, KrF (l = 0.25 mm) and DPSSL (l = 0.33 mm). At present, the KrF has better beam quality for current, high gain targets; whereas the DPSSL has greater durability and better efficiency. However, both driver approaches have the potential for meeting all the IFE requirements. The parallel architecture for IFE drivers simplifies the development of the beam line modules. John discussed the current and proposed development plans for both types of lasers to meet the demanding goals to achieve a commercial product.

Candidate IFE Drivers (HI) Simon Yu is leading the HI virtual laboratory for heavy ion beam technology with support from LBL, LLNL, and PPPL. Simon briefly summarized the heavy ion beam requirements and program plans for HI beams to be a viable IFE driver. The current thinking is to use multiple beams for the driver. Space charge is the biggest challenge. Heavy ion accelerators offer a large parameter space with a proven technology base. Final focus and protection of final focus elements need substantial development. Ballistic focus is one of the primary methods of focusing the beams but self-pinching and channel formation are also being considered. As other chambers more suitable for HI beams are being evaluated, Simon will be solicited by the team for more in-depth analyses and interaction with the team.

Driver/Chamber Interface Jeff Latkowski described the important driver issues. Lasers need to protect the final optic (perhaps a grazing incidence metal mirror (GIMM) or a transmissive wedge) in the line of sight and the focussing mirror just out of the line of sight. HI drivers need protection for their final focussing magnets. Radiation damage to these elements will reduce their effective lifetimes, increase the waste stream volume, and negatively impact the plant maintenance and availability.

Chamber Physics Analyses

Chamber Physics Bob Peterson proposed using the gas-protected dry wall chamber approach as it has the advantages of being adaptable to a variety of dry walls, accommodate multiple wall openings, allowing high temperature wall operation, and it is a relatively passive system. The gas will attenuate and slow the energy being transmitted to the first wall. But the protective chamber gas density will limit laser propagation and adversely affect the heating and position of the target during injection. Issues to be addressed are the target output, physics and radiation transport in the gas, response of the chamber wall, laser breakdown of the gas, impurity buildup, and gas transmutation. The Sombrero study used a 0.5 Torr of xenon to stop most of the 1.6  MeV carbon atoms. The UW BUCKY 1-D radiation-hydrodynamics code is used to simulate target, gas behavior, and wall response. It predicted a graphite first wall surface temperature of 2155C, which is below the sublimation temperature of graphite but higher than the 1485C predicted in the Sombrero study.

Inertial Fusion Target Output Bob discussed the typical DD and ID laser target materials. Neutrons and energetic ablator ions dominate the direct drive target output, whereas neutrons and x-rays dominate ID target output. Bob mentioned that UW has been studying the sensitivities of the Z-pinch ID target performance parameters due to variations in the fabrication process. It was questioned if this target design is applicable to laser or HI targets.

Blast Propagation Donald Haynes explained how the opacity of the chamber gas is now considered to be an important input to the evaluation of the chamber and the power plant designs. The protective gas stops most of the energetic carbon ions before they reach the wall, keeping below sublimation temperature and reducing the blast impulse and peak pressure. The key gas property is its opacity. A more sophisticated suite of EOS / Opacity codes are being used to re-investigate the effects of Xe opacity on chamber dynamics.

Wall Response in Gas-Protected Chambers Bob Peterson stated that a chamber gas could be effective in preventing thermal damage to first walls. Without gas protection, there would be significant vaporization of the Sombrero first wall. The peak FW temperatures depend on the thermal conductivity of the first few microns.

Pulsed Heating and Hydrodynamic Effects of IFE Chamber Walls During Microexplosions Ahmed Hassanein discussed the application of the High Energy Interaction with General Heterogeneous Target System (HEIGHTS) to describe first wall responses to IFE target microexplosions. Results of several material and process examples were examined to relate to the chamber wall responses.

Systems Analysis

Modeling Tools, Evaluation Metrics, and Systems Integration Ron Miller told the group that the systems analysis group intends to use elements of the formal systems engineering process (requirements analysis, functional analysis, system architecture, interface analysis, and system integration) to provide an integrated approach to the chamber assessment. The intent is to determine a design window for each chamber concept over a range of plant sizes, drivers, and targets. Each design window will be assessed for a number of parameters, including repetition rate, target gain, driver energy, etc.

Chamber Assessment

Damage Resistant Materials and Damage Mitigation for Dry-Wall Chambers Mark Tillack highlighted the need to have chamber walls that survive or are renewable. This need can be achieved with damage-resistant materials or mitigation schemes. Gas protection schemes are possible, but target propagation and accurate positioning will be difficult above 5-100 mTorr. Instead, Mark suggested reducing total x-ray yield and hardening the target spectrum to provide deeper photon penetration while diverting the debris energy. A lower pressure chamber gas might strip the softer part of the spectrum. Magnetic diversion of debris might provide adequate protection. Advanced materials (e.g., carbon velvet) may offer superior thermal properties for this application. He is proposing near term activities in dry wall concept development, thermo and mechanical wall responses, and propagation physics.

Issues for Dry-Wall Concepts Igor Sviatoslavsky presented three ways to protect chambers from x-rays and ions: distance, gas protection, and liquid walls. Propagation of ion beams or injection of the targets within a large-radius chamber can be difficult. A low-pressure gas helps to stop the x-rays and ions and radiates the energy over longer time scales. Liquid or liquid protected walls can absorb the energy on the liquid surface. Details of chamber designs from prior studies were discussed.

Igor discussed issues related to target injection through the chamber protective gas. He suggested a possible delivery system using a sacrificial injection tube that would extend some distance toward the center of the chamber. This injection tube would protect and guide the target a portion of the way to the chamber center. As the leading edge of the tube is ablated, the tube is extended a like amount toward the center. Igor also mentioned the design, protection, and maintenance of the GIMM and wedge optics.

Neutronics and Shielding Issues Mohamed Sawan informed the group that, although there appears to be a lot of similarity between MFE and IFE neutronics, there are significant differences in geometry and much higher instantaneous damage rates. The IFE blankets have less difficulty in achieving adequate tritium breeding. Special emphasis will be required to shield the final focus elements. As the target is heated and compressed, there are fuel and neutron interactions that soften the neutron spectrum, neutron multiplication, and gamma production. Energy deposited by neutrons and gammas heat the target and result in radiated x-rays and expanding ionic debris. The source term for all neutronic analyses are the spectra of neutron and gamma photons emitted from the target. It is important to couple the target neutronics and hydrodynamics to account for continuously time-varying densities, configuration, and source distribution. Results from a LIBRA-SP and X-1 target were shown to illustrate these effects. Because of the neutron and target interaction, the energy multiplication is defined differently for IFE.

The chamber neutronic results indicated the blankets developed for MFE could be used for the contemplated IFE concepts, after accounting for the geometric effects of a nearly-point IFE source. Due to the spherical geometry, the radial gradient of the radiation effects within the blanket is reduced. The neutron source spectrum is softened (10-12 MeV average). Due to the pulsed nature of IFE, it is essential to determine instantaneous damage rates for prediction of the structural lifetime. The blanket neutronics for the wetted wall concept is identical to the dry wall concept because the thin liquid sheet provides negligible neutron attenuation. The recommended thickness for the biological shield is on the order of 2.5 to 3.5 m of reinforced concrete shield to maintain a dose rate of < 2.5 mrem/h outside the shield. Line-of-sight components can be protected (mitigated) from the neutrons only by providing greater distance from the source. The radiation damage limit for the final focusing mirror is unknown. The shielding of the HI final focussing magnets must be tailored to achieve expected life of plant lifetime. A 3-D analysis is needed for IFE neutronics/shielding activity.

Pulsed Radiation Damage by IFE Neutrons Jerry Kulcinski pointed out that the differences between MFE and IFE radiation damage are caused by geometrical, spectral, temporal, and rate effects. Jerry supported this conclusion with results from several experiments. He also showed a graph of swelling observed in carbon materials as a result of radiation dpa levels. Past experimental data indicted a significant effect of pulsed neutron damage on swelling of metals (at low damage rates). He noted that there are no irradiation test facilities that will adequately simulate typical IFE pulsed radiation damage levels. There is no experimental or theoretical data on IFE pulsed effects on C-C and SiC composites.

Pulsed Activation of Target and Chamber Douglass Henderson explained that the DD and ID targets have different output spectra. The fuels burn from a central hot spot to the remainder of the fuel. Time dependent activation analyses are required, but usually quasi-static neutron transport and activation calculations are conducted. Neutrons will activate the chamber gas and structural materials. Radioactive target debris will condense or plate out on the wall and other chamber surfaces and affect biological dose rates, decay heat, WDR, and chamber maintenance approaches. It is important to model the pulse irradiation history to obtain accurate results.

Tritium Issues Dai-Kai Sze repeated the earlier statements that an IFE blanket will easily be able to breed adequate amounts of tritium to maintain the required tritium breeding ratio. He mentioned that we only need to keep a few kilograms of tritium inventory in the factory, which would represent a few days of fuel burnup. Additionally, there is a large surface area within the chamber and the confinement vessel that would significantly contribute to the plant tritium inventory. The choices of the breeding materials and coolants will affect the tritium processing option and the tritium inventory.

Safety and Environmental Activities - Dave Petti outlined the proposed approach to integrate the team activities for the concept definition phase and the detailed assessment phase. Dave suggested a tentative team makeup to address each of the topic areas. Initially the safety and environmental activity would involve the definition of the metrics and establishment of safety and environmental criteria to be applied to the IFE concepts. Metrics would be based upon:

  • Mobilized tritium, activation product, and dust/debris (inventories?)
  • Decay heat
  • Chemical reactivity and combustible gas generation
  • Waste and environmental effects

Release limits and inventories would support the key public goal of no-evacuation criteria per the DOE Fusion Safety Standard. Confinement barriers would be established to help contain the inventories that could be mobilized. A list of potential accidents was postulated for the safety assessment. Mitigation features would include decay heat removal, minimization of activation product mobilization, control of coolant energy, control of chemical energy, rapid plasma shutdown, and confinement barriers.

IFE Summary and Action Items

Farrokh Najmabadi commended the entire team for an excellent compilation of technologies to commence the ARIES-IFE chamber assessment study. The presenters proposed a good set of goals and assessment plans to form a comprehensive chamber assessment. To support the combined plan, Farrokh asked each task leader to prepare a set of near-term actions to be accomplished nominally within the period before the next project meeting. These task group actions were discussed and revised at the meeting. The final set of actions items are documented on the ARIES web site.

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