FUSION REACTOR DESIGN AND TECHNOLOGY

Transcription

1 FUSION REACTOR DESIGN AND TECHNOLOGY Report on the Fifth IAEA Technical Committee Meeting and Workshop, held at the University of California, Los Angeles, California, United States of America September 1993 R.W. CO"*, F. NAJMABADI, M.S. TILLACK Institute of Plasma and Fusion Research, University of California, Los Angeles, California, United States of America Y. SEKI Naka Fusion Research Establishment, Japan Atomic Energy Research Institute, Naka-machi, Naka-gun, Ibaraki-ken, Japan Yu.A. SOKOLOV Russian Research Centre, Kurchatov Institute, Moscow, Russian Federation J. CLEVELAND, V.V. DEMCHENKO International Atomic Energy Agency, Vienna INTRODUCTION J. Cleveland, V.V. Demchenko The objective of fusion power reactor design and technology development is to bring to the world a safe, economical and essentially unlimited energy source. While considerable progress has been made during the last few years, there is still a great deal of research and development needed. The IAEA convened the fifth Technical Committee Meeting and Workshop on Fusion Reactor Design and Technology at the University of California, Los Angeles (UCLA), United States of America, for the purpose of reviewing the status of * Present address: School of Engineering, University of California, San Diego, California, USA. fusion reactor design and technology, and specifically to discuss the status and prospects of emerging commercial fusion reactor concepts, including aspects of plasma engineering, technology, materials development and safety, for fusion devices beyond ITER, including both magnetic and inertial confinement concepts. Another important topic was the elaboration of possible fusion development pathways to electric power demonstration (DEMO), including objectives, requirements, features and database needs. The first IAEA meeting in this series, a Workshop on Fusion Reactor Design Problems [ 11, was held in January 1974 at the Culham Laboratory in the United Kingdom. The second, a Technical Committee Meeting NUCLEAR FUSION, Vo1.34, No.5 (1994) 747

2 and Workshop on Fusion Reactor Design Concepts [2], was held in Madison, Wisconsin, USA, in October The third meeting [3] was held in Tokyo, Japan, in 1981, and the fourth meeting [4] was held in Yalta, Soviet Union, in These meetings have been extremely fruitful in assessing the state of the art of fusion reactor design and also in indicating the appropriate orientation of research and development for the future. At the 1993 meeting, there were 85 participants from six countries and three international organizations, and a total of 33 papers were presented. Invited papers described the status of fusion reactor design and technology programmes in China, the European Community, Japan, the Russian Federation and the United States of America. An invited paper was also presented on the design and status of the ITER project. Several technical papers described design and technology development activities for magnetic confinement fusion energy systems, including tokamaks, stellarators and hybrid fission-fusion reactors, Inertial confinement fusion energy reactor and technology development were covered in five technical papers. Other papers dealt with development pathways to commercial fusion power and the associated needs. The meeting provided a needed forum to discuss design and technology for several types of fusion reactor, specifically for tokamaks, stellarators, inertial confinement systems and fission-fusion hybrid systems. Also, it provided a forum for discussion of design and technology issues beyond the objectives of ITER, for example the design and further technology needs for DEMO. Workshop sessions were organized by UCLA on the following topics: (a) Fusion: The Industrial Perspective, (b) Fusion: The Magnetic Fusion Energy Programme Perspective, (c) Future Experimental Devices, (d) The Inertial Fusion Energy Programme Perspective. Each session included presentations by invited speakers well known in the field, followed by question and answer sessions. The following summary reports the strategic goals and development pathways for fusion, the status of fusion reactor design studies, the status and R&D needs of plasma physics and plasma engineering, and the status and R&D needs of fusion nuclear technology and materials. The summary draws heavily on the papers and discussions of the Technical Committee Meeting and incorporates additional information as needed in order to provide an in-depth review of these topics. REFERENCES Fusion Reactor Design Problems (Proc. Workshop Culham, 1974), Nucl. Fusion Spec. Suppl. 1974, IAEA, Vienna (1974). Fusion Reactor Design Concepts (Proc. 2nd Tech. Comm. Mtg and Workshop Madison, 1977), IAEA, Vienna (1978); CONN, R.W., et al., Nucl. Fusion 18 (1978) 985. Fusion Reactor Design and Technology (Proc. 3rd Tech. Comm. Mtg and Workshop Tokyo, 1981), 2 vols, IAEA, Vienna (1983); ISO, Y., et al., Nucl. Fusion 22 (1982) 671. Fusion Reactor Design and Technology 1986 (Proc. 4th Tech. Comm. Mtg and Workshop Yalta, 1986), 2 vols, IAEA, Vienna (1987); ABDOU, M.A., et al., Nucl. Fusion 26 (1986) STRATEGIC GOALS AND DEVELOPMENT PATHWAYS FOR FUSION R.W. Conn, F. Najmabadi The strategic goals for fusion development and the development pathways to be followed are fundamental questions of fusion policy. Paper presentations and workshop sessions at the meeting allowed the discussion of views from around the world on these issues for both the magnetic and the inertial fusion programmes. For this part of the summary, we focus on these questions and summarize first the discussions regarding magnetic fusion and secondly the discussions regarding inertial fusion Magnetic fusion energy: Strategic goals The papers presented summarized the strategic objectives and programme structure for magnetic fusion energy (MFE) programmes in China, the European Community, Japan, the Russian Federation and the United States of America. The strategic objective of the programmes of the EC, Japan, Russia and 748 NUCLEAR FUSION, Vo1.34, No.5 (1994)

3 the USA (to be referred to here as the four parties ) is economically and environmentally attractive fusion electric power plants. These four parties view their magnetic fusion programmes as energy programmes, all agree that the ITER project is the next major step on the road to their goal and all agree that the ITER programme must not fail. The US programme has a plan published by the US Department of Energy as part of a National Energy Strategy (NES). For fusion, the plan calls for a technology demonstration date of 2010 and a fusion demonstration power plant (DEMO) beginning operation in Programmes in the EC and in Japan have not set a date for such a future facility and indicate that flexibility with respect to the DEMO operating date is important. Participants did indicate, however, that construction of a DEMO could start in Japan and in Russia in the time frame, which coincides with US aspirations. It was noted that the additional funding needed for key elements of the US programme, other than for ITER, to meet the 2025 DEMO operation date has not yet been forthcoming. Hence, in practical terms, the programmes of the EC, Japan, Russia and the USA have the same strategic goal, the same major next step facility (ITER) and, de facto, a flexibility with respect to the timing of DEMO. The strategic goal of the magnetic fusion programme in China is different. In papers and at workshop sessions, the delegates from China gave the production of fissile fuel for commercial nuclear reactors as China s strategic fusion energy goal. This goal is based on the estimate that 1500 GW of electric power will be needed in China by 2050, on the environmental impact of meeting such a requirement using fossil fuels, on the view that nuclear plants will be employed to meet new needs, and on the view that the supply of fissile fuel up to and beyond 2050 in China will be insufficient MFE: Development pathways Details of development pathways and programme elements were discussed mainly for the EC, Japanese, Russian and US programmes. Discussion here is restricted to these four programmes. Development pathways and programme elements of the four parties had major similarities and some differences. All four parties have ITER as the central next step in their fusion development programme. All indicate the importance of maintaining the timetable for ITER and of achieving a decision to construct ITER by the end of the present Engineering Design Activities (EDA) phase. Beyond ITER, differences emerged in NUCLEAR FUSION, Vol.34, (1994) both philosophy and specific requirements. These include blanket, structural material, coolant and tritium production technologies. While no party, including the USA, has announced a national policy, fusion workers from each party offered different approaches to DEMO. A similarity developed in this diversity in that the choices were all driven, as are those of the USA, by a concern for public acceptance. This was particularly true of Russia and the EC. In a workshop session, participants from industrial firms in the USA expressed the view that a more comprehensive plan for fusion development is needed, having major elements in addition to ITER, to permit a more defined plan and timetable to meet the strategic goal. The industrial participants and others discussed the need for a base programme in materials development and testing that would run in parallel with the ITER EDA, in order to capitalize on ITER results and move fusion forward. In particular, the participants identified the role of a volume neutron source (VNS) for the development of fusion blanket engineering and reliability, though it is not now an agreed upon element of any of the world s fusion programmes. The VNS is a fusion machine and is distinct from, and in addition to, an accelerator based point neutron source (PNS). The PNS is to obtain high fluence data on small samples of structural materials to provide basic design data, All parties have the PNS as a major element in their programme plans. For all parties, the position on a VNS is unsettled and there was discussion about whether both a VNS and a PNS are required, particularly from the EC perspective. Another difference in development pathways relates to magnetic confinement alternatives to the tokamak. For all four parties, the tokamak is the primary and leading approach. In Japan and in the EC, there are also major programmes to develop alternative magnetic confinement approaches. In Japan, by far the largest effort is focused on the stellarator/heliotron concept as represented by the LHD project of the National Institute for Fusion Science under construction at the Toki site. In Europe, the reversed field pinch (RFP) approach is pursued at the Istituto Gas Ionizzati del CNR in Padua with the world s largest RFP (the RFX machine), and the stellarator approach is pursued at the Max-Planck- Institut fiir Plasmaphysik in Garching with the W7-AS machine and the proposed W7-X experiment. By contrast, neither the USA nor Russia has plans for altemative magnetic confinement machines on the scale of LHD, RFX or W7-X. However, the USA does have plans for an advanced tokamak, the TPX, the scale and timing of which are similar to those of the LHD and w7-x. 749

4 During the workshop, important issues were raised and discussed about each party s development strategy. For the US MFE plan, it was pointed out that ITER is the key device, but that the 2025 date for DEMO is very unlikely to be met if ITER is the only facility between now and DEMO. Some US industry participants pointed to a mismatch between the current plan and actual needs. Further, they pointed out that the justification for a date such as 2025 must and can be made stronger, and that the needs of the electric power market must be addressed over the next 30 years as part of this process. From the EC perspective, it was made clear that the ITER programme must not fail, that a successful ITER is central to overall success, and that ITER alone is probably not sufficient to provide the basis to design and construct DEMO. Further, the EC is now in the midst of considering whether or not to maintain a confinement alternative to the ITER tokamak, namely the stellarator. The EC has the largest technology development programme among the parties, though its size is still small compared with the need. The EC has recently established the Safety and Environmental Assessment of Fusion Power (SEAFP) project, driven by public and parliamentary concerns, and this will help guide future technology development important to moving towards DEMO. The modest scope of technology development in the EC, and the meagre scale in the USA and elsewhere, as well as the recently completed ARIES studies on tokamak power plants, indicate that current MFE technology development worldwide is not of sufficient size to support a DEMO on the anticipated time-scales. From the Japanese perspective, it was made clear that ITER must go forward on its planned timetable, and that ignition and a successful plasma burn are central ITER goals. Further, the need for other facilities, while recognized, must not interfere with ITER construction. The improvement of the tokamak concept from a physics point of view is seen as a desirable objective. From the perspective of Russia, ITER is the key need, and 80% of the Russian magnetic fusion budget is devoted to the ITER programme. However, the importance of a VNS facility was indicated, and the siting of a VNS device at a former defence site undergoing conversion would be highly desirable. One common refrain from participants of all four parties related to the present fusion budgets, which have been constant for some time. To proceed with construction of ITER, increased budgets are vital Inertial fusion energy: Goals and pathways Inertial fusion energy (IFE) programmes of substance exist in the USA, Japan, Russia, China and the EC, with different primary objectives. Most of the funding for inertial fusion has been, and continues to be, provided by the defence programmes in these countries. Japan has the largest non-defence-funded inertial fusion programme, while that of the EC is of a size necessary only to maintain a watching brief with respect to progress in the field. Given the disparities in objectives and in programme sizes, it is fair to say that for IFE there is less of a common view worldwide regarding development strategies than there is for MFE. However, with the exception of China, the energy oriented portions of the IFE programmes all have fusion electric power plants with attractive safety, environmental and economic features as their goal. All agree that demonstration of ignition in the laboratory (it has already been demonstrated in underground tests) is the next critical step. Papers from the USA discussed the US IFE programme and the development pathways under consideration. Since only the plans for the USA were presented, a brief summary of the presentations and discussion is provided. The NES of 1991 foresaw that the demonstration of heavy ion fusion would require the construction of five major facilities: a National Ignition Facility (NIF), the Induction Linac Systems Experiments (ILSE), an Intermediate Driver Facility (IDF), an Integrated Test Facility (ITF) and a Demonstration Power Plant (DPP). The potential to use an upgradable driver for the IDF, ITF and DPP facilities will result in the least expensive development pathway. The NIF will be used to ignite targets and produce energy gains of up to 10 with a laser driver. Other experiments will study energy transport in laser targets in attempts to estimate the minimum ion energy required to ignite a target with ions. The NIF experiments will also study the target manufacturing tolerances required for mass production. The driver development programme will demonstrate that an ion driver can be built with the required beam performance, high pulse rate capability, and necessary low cost characteristics. The ILSE will address emittance growth and beam transport issues and will test concepts for all necessary beam manipulations (beam merging, bending, compression and focusing) at significant scale. The IDF that follows ILSE will be an end-to-end test of a W driver. It is intended to provide an engineering demonstration of longitudinal instability control, provide the first beam- 750 NUCLEAR FUSION, Vo1.34, No.5 (1994)

5 target physics experiments with a heavy ion driver and provide validation of engineering performance. The ITF will be the first facility to integrate all the major subsystems required for an inertial fusion power plant, including the driver, target production and injection systems, and reactor systems (chamber, blanket, heat removal, tritium recovery, etc.). Because fusion capsule ignition and burn physics are independent of the reaction chamber size, feasibility tests of various chamber concepts can be made at much smaller sizes (about 1-2 m first wall radius) and much lower powers (tens of megawatts) than for magnetic fusion development facilities such as ITER. A reactor technology programme, using the above and other existing facilities, will demonstrate the reactor technologies needed for IFE, including target injection and tracking, beam steering and reactor-driver interface systems. Using the output of ignited targets, NIF experiments will characterize radiation, shock and debris effects on various first wall candidate materials and on reactor-driver interface systems, providing the data necessary to design multiple pulse experiments for the ITF. It was noted that near term progress and projections have not met the NES schedule for IFE development. Presentations and workshop discussions strongly suggested that meeting the goal of an IFE demonstration reactor operating by 2025 will not be possible without increased funding. The principal reasons are that the IFE driver development budget is too small, and technology development in all areas besides driver development and pellet fabrication is currently inadequate. On the other hand, recent IFE reactor studies have identified the diverse and extensive R&D needs of an IFE power programme. Finally, it was noted that security classification severely limits international co-operative development of the large facilities needed on the pathway of IFE development. Relaxation of such classification has just recently been announced by the US Department of Energy. It is to be hoped that this will bring about change so that international cooperation will now become a more open pathway for IFE power plant development. 2. FUSION REACTOR DESIGN STUDIES Y. Seki 2.1. Introduction The fusion reactor design studies presented and discussed at the meeting are summarized. This summary, however, does not represent all of the reactor design studies done in the world since the last IAEA meeting and workshop in Yalta in The reactor designs presented are based on the DT fuel cycle (both magnetic and inertial confinement concepts) with the exception of ARIES-111, which uses D-3He fuel. In Section 2.2, DT magnetic confinement fusion reactor design studies are summarized. Tokamak studies presented at the meeting are introduced first, followed by stellarator and RFP reactor studies. Six recent DT inertial confinement fusion reactor studies presented at the meeting are summarized in Section 2.3. A brief summary on D-3He reactors is given in Section 2.4. A description of fusion-fission hybrid reactor designs is given in Section 2.5. Finally, in Section 2.6, some overall observations on reactor studies are given DT magnetic confinement fusion reactors 2.2. I. Tokamaks Four tokamak power reactor studies, with steady state operating mode, are summarized in Table 2.1. These reactors, which are proposed at four different institutions, show a common trend of high aspect ratio with low plasma current and large bootstrap current fraction. They represent advanced concepts of tokamak reactors. In 1990, the Steady State Tokamak Reactor (SSTR) concept [2.1] was proposed as a realistic fusion reactor to be built in the near future. The major feature of SSTR is the maximum utilization of a bootstrap current in order to reduce the power required to maintain steady state operation. The ARIES study of four conceptual tokamak fusion reactor designs has recently been completed [2.2]. Three of the designs are based on DT and one on D-3He fuel cycles. All are steady state tokamak reactors producing - lo00 MW(e) net. ARIES-I bums DT and operates in the first stability regime. ARIES-I1 and ARIES-IV both have similar physics performance and NUCLEAR FUSION, Vo1.34, NOS (1994) 75 1

6 TABLE 2.1. STEADY STATE TOKAMAK POWER REACTORS Major radius, R (m) Plasma radius, U (m) Aspect ratio, A (=R/u) Plasma current (MA) Magnetic field on axis (T) Maximum magnetic field (T) Fusion power (MW) Current drive power (MW) Bootstrap current fraction Reference SSTR ARIES-I DEMO-S plant model [2.1] [2.2] [2.3] ~ s TABLE 2.2. PULSED TOKAMAK POWER REACTORS IDLT ~2.51 Conventional Reactor Concept G.61 DEMO-P ~2.31 Major radius, R (m) Plasma radius, U (m) Aspect ratio, A (=R/u) Plasma current (MA) Magnetic field on axis (T) Maximum magnetic field (T) Fusion power (MW) Pulse duration (h) Dwell time (min) Plasma heating power (MW) Bootstrap current fraction operate in the second stability regime; however, their engineering designs are different. ARIES-I1 employs a self-cooled design using liquid lithium as the primary coolant and a vanadium alloy as the structural material. ARIES-IV employs a helium cooled, solid breeder design. The ARIES study points towards the commercial attractiveness of tokamak fusion reactors as power plants, especially with the DT fuel cycle. Good safety performance can be expected with the helium cooled, solid breeder design using Sic as the structural material. Parameters for a steady state tokamak reactor, DEMO-S, in which plasma current is sustained by the bootstrap effect only, have been presented [2.3]. DEMO-S will be compared with the pulsed option, DEMO-P. A reference plant model is assumed for the European Safety and Environmental Assessment of Fusion Power (SEAFP) [2.4]. It is considered to be a reasonable concept for a future fusion power station and is appropriate as the focus for safety and environmental assessments. The pulsed operation tokamaks are presented in Table 2.2. They are based on the direct extrapolation of currently attained physics data. An Inductively Operated Day-Long Tokamak (IDLT) Reactor [2.5] has been proposed. It is a pulsed tokamak fusion power plant with only inductive current drive but with steady state electric power output. The operation time and the dwell time of the tokamak are taken to be 13 h and 300 s, respectively. Thermal power input to the turbine generator during the dwell time of the tokamak discharge is supplied by burning hydrogen gas, which is produced by electrolysis cells throughout the power plant operation. Table 2.2 summarizes typical parameters of three inductively operated, pulsed tokamak reactor concepts. The DEMO-P design [2.3] has a pulse duration of about one hour, and neutral beam injection, with energy up to 3 MeV, or an ICRF fast wave system is to be used for heating and current drive. The Conventional Reactor Concept [2.6] is a large, high current tokamak reactor concept based on the present state of the art. It is assumed to operate at a current of 25 MA with a poloidal cross-section close to that of ITER EDA, but with the major radius increased in order to allow sufficient flux capability. A maximum field at the superconductor of 12 T is also assumed. Both difficulty and cost (for external current drive) are involved in operating a tokamak reactor in steady state. Therefore, the PULSAR reactor study is currently under way in the USA to investigate the commercial attractiveness of pulsed tokamak reactors in comparison with steady state tokamak reactors. Particular attention is being given to the thermal fatigue issue for the fusion power core, especially for the first wall and divertor, operation of the balance of plant in steady state, and delivering constant electrical power to the grid by using a suitable energy storage. The PULSAR study is scheduled to be completed by early NUCLEAR FUSION, Vo1.34. N0.5 (1994)

7 TABLE 2.3. COMPARISON OF STELLARATOR AND TOKAMAK POWER REACTORS HSR SEAFP [2.4] CT ARIES-IV [2.2] Major radius, R (m) Average plasma radius, ap (m) Aspect ratio, R/ap Plasma volume (m3) Magnetic field on axis (T) Maximum magnetic field (T) Volume average 0 (%) Neutron wall loading (MW/m*) Fusion power core mass (t) lo o 22 OOO >44 OOO Stellarators Stellarators would have significant operational advantages as ignited steady state reactors because they do not require a net plasma current (and the continuous power recirculated to the plasma to drive it). The magnetic field is created by currents only in external coils, resulting in inherently steady state, disruption free magnetic configurations with relaxed constraints on the plasma parameters and profiles and a wide range of magnetic configurations available for optimization and control. The larger aspect ratio of stellarator reactors allows access from the inboard side for easier maintenance and a lower neutron wall loading, which increases the first wall and blanket lifetime. Two complementary approaches to stellarator reactors are being pursued: (1) a conservative extrapolation of the higher aspect ratio, modular coil HELIAS configuration to reactor size and study of its critical features [2.7]; and (2) extensive scoping and trade-off studies using more advanced materials assumptions for a lower aspect ratio compact torsatron with helical windings [2.8]. Table 2.3 compares the main device parameters, including the average plasma radius (ap) and the mass of the fusion power core, for the two types of stellarator reactor (the Helical Stellarator Reactor (HSR) and the six field period Compact Torsatron (CT6)), the US second stability tokamak reactor ARIES-IV [2.2], and the reference reactor for the SEAFP study The technology assumptions employed in the HSR and SEAFP reactors are very different from those used in the CT6 and ARIES-IV reactors. It should also be noted that a heliotron reactor study is in progress [2.9]. The HELIAS (Helical Advanced Stellarator) configuration was obtained by simultaneously optimizing a set of physics properties that are desirable for a stellarator reactor, including good confinement of both the thermal plasma and the fusion produced 01 particles. The HELIAS configuration is produced by 50 moderate sized non-planar coils arranged in a torus that has the same shape as the non-axisymmetric plasma; the average coil radius a, is 3.9 m, smaller than that in a tokamak reactor. The HSR follows a conservative engineering approach that employs currently available materials for the structure and magnets (for instance, the low maximum field on the coils allows the use of NbTi instead of Nb3Sn for the superconducting magnets). The modular design allows easier disassembly for repair and maintenance of the reactor. The blanket and shield are assumed to be the same as those used in tokamak reactors. Divertor plates inside the modular coils are used for particle and power handling, as in a tokamak. The CT configuration was obtained by maximizing the average radius of the last closed magnetic surface, subject to constraints that maximized the p limit. The configuration is produced by two large open helical windings on a circular cross-section torus (a, = 4 m). The CT6 reactor follows the aggressive approach of the ARIES and PULSAR tokamak reactor studies by using thicker, advanced, low activation materials for the blankets (Li20/SiC/Be/He) and shields (SiC/Pb) and higher current density, higher field, Nb,Sn magnets. The CT6 reactor has a large access space between its helical windings for blankets and maintenance. The diverted flux pattern exits between the helical windings, allowing an external divertor chamber with lower NUCLEAR FUSION, Vo1.34, No.5 (1994) 753

8 power density on the divertor plates. The presence of a near perpendicular loss region for CY particles prevents the accumulation of helium ash, which compensates for the lost CY particle heating. The CT6 case is similar in many of its parameters (and its cost of electricity) to ARIES-IV. TABLE 2.4. RFP REACTOR PARAMETERS Net electric power (MW) Fusion power (MW) Gross thermal power (MW) Neutron wall loading (MW/m*) Major radius (m) Minor plasma radius (m) Plasma elongation 4 Magnetic field on axis (T) Magnetic field maximum (T) Plasma current (MA) Breeder Structure Coolant Outlet temperature ("C) Plasma heating Current drive Pulse length Recirculating power fraction Cost of electricity (mills/kw.h)a REPUTER-I TITAN-I [2.10] [2.11] lo o Li20 Sic He 300 Ohmic f-8 pumping Continuous - 50 a Update (in 1992 US $) of published projection Li V-3Ti-1Si Li 700 Ohmic f-8 pumping Continuous Reversed field pinch The RFP is a third toroidal magnetic confinement approach, characterized by its dominant poloidal magnetic field, high /3 value and ohmic heating to ignition. The acceptability of resistive, low field toroidal field coils in the RFP allows thin radiation shields and high system power density. Two recent RFP power plant design studies, REPUTER-1 [2.10] and TITAN [2.11], summarized in Table 2.4, explored the features of this concept. Development of the RFP requires demonstration of favourable scaling of confinement time with increasing plasma current (now < 1 MA) in such experiments as the RFX at Padua. Ongoing activity relating to the RFP, as well as other alternative fusion concepts, has been significantly reduced since the last meeting in 1986 [2.12] DT inertial confinement fusion reactors Six inertial confinement fusion power reactor studies with DT fuel cycles were presented and are summarized below. Table 2.5 summarizes some of the key parameters for the six designs Laser drivers The Sombrero design [2.13] utilizes a KrF laser driver with e beam pumped amplifiers and angular multiplexing for pulse compression. Sixty beams deliver a total of 3.4 MJ to a direct drive target at 6.7 Hz. The laser system achieves an overall system efficiency of 7.5% and has a total energy consumption of 304 MW. The Sombrero chamber features a C-C composite first wall and blanket structure with a granular TABLE 2.5. MAJOR DEVICE PARAMETERS FOR INERTIAL FUSION REACTOR DESIGNS Sombrero Prometheus-L KOYO Osiris Prometheus-H HYLIFE-11 [2.13] [2.14] [2.15] [2.13] [2.14] [2.16] Net electric power (MW) lo lo Target type Direct drive Direct drive Direct drive Direct drive Indirect drive Indirect drive Driver energy (MJ) Repetition rate (Hz) Chamber radius (m) Breeder Li,O Li20 LiPb Flibe Li20 Flibe Primary coolant Li,O/He He, Pb LiPb Flibe He, Pb Flibe Structures C-C composite Sic composite Sic composite Graphite fabric Sic composite 304SS First wall protection 0.5 torr Xe Pb film LiPb Flibe Pb film Flibe jets 754 NUCLEAR FUSION, Vo1.34, No.5 (1994)

9 ions CONFERENCES AND SYMPOSIA Li20 breeding blanket. The Li20 granules flow through the blanket region of the chamber and serve as the primary coolant. The first wall is protected by 0.5 torr of Xe gas. Low pressure He is used to remove tritium from the breeding blanket and also to transport the LizO granules to and from the intermediate heat exchanger. The Prometheus-L reactor [2.14] also uses a KrF laser driver. The laser architecture is based on moderate energy (6 U), electric discharge excimer laser amplifier modules coupled with non-linear optical systems for beam combination and pulse compression. Sixty beams converge on the direct drive target with 4 MJ of driver energy. The cavity consists of separate first wall and blanket systems. The first wall panels are made of Sic composite with a porous first wall to allow the Pb protectant and first wall coolant to form a protective film on the surface facing the blasts. The blanket is a He cooled, solid breeder design that is similar to magnetic fusion blankets. Recently, a laser fusion reactor system called KOYO-1 [2.15] has been proposed by the Institute of Laser Engineering (ILE), Osaka University. In this system, four plasma chambers are driven by a 12 Hz laser diode pumped solid state laser. The chamber wall is protected by a liquid LiPb flow through a path made of Sic Heavy ion beam drivers The Osiris reactor [2.13] contains a heavy ion driver using linear induction accelerator technology. Twelve beams of 3.8 GeV Xe' ions deliver a total of 5 MJ to an indirect drive target at a pulse repetition rate of 4.6 Hz. The design is conservative in that it does not employ beam combination, beam separation or recirculation. The Osiris chamber features a flexible, porous, carbon fabric first wall and blanket that contain the molten salt Flibe, which serves as the tritium breeding material and primary coolant. A thin layer of Flibe coats the first wall for protection, and a cold spray at the cold leg temperature is injected at the bottom to aid in recondensation. Prometheus-H [2.14] uses a Pbz+ heavy ion driver. A single beam linac with storage beams was chosen, together with a novel channel transport technique for final focusing onto the indirect drive targets. Eighteen beamlets converge from two sides, are stripped of electrons at the entrance to the cavity and then form a narrow self-pinched channel to the target. The beamlets are held in storage rings and combined before final transport. A single cavity design was judged acceptable for both Prometheus-L and Prometheus-H. HYLIFE-I1 [2.16] uses a 5 MJ recirculation induction accelerator that delivers 10 GeV Hg + onto the targets six times per second. The driver characteristics were adopted from the HIFSA study [2.17]. The focus of the HYLIFE-I1 study was on the reactor design. The HYLIFE-I1 reaction chamber uses neutronically thick jets of liquid Flibe injected through nozzles and pulsed in such a way as to produce pockets where the targets are injected and microexplosions occur. Ordinary stainless steel can be employed because the thick liquid jets absorb the energy of the explosions, including the neutrons D-3He fusion reactors In a fusion power plant, the use of D-3He fuel would provide a significant reduction in neutron production and the resulting generation of radioactivity and damage in materials, but it would require a significantly more advanced level of plasma performance than would DT fuel. The reduced neutron environment would make possible a simpler heat recovering shield and the use of materials and coolants generally not applicable in the intense neutron fluxes associated with DT fuelled systems. Parameters for two conceptual D-3He tokamak power plants, the first stability Apollo-L4 [2.18] and the second stability ARIES-I , are summarized in Table 2.6, together with parameters for the FRC system Artemis [2.20]. The topic of advanced fusion fuels received little attention at this meeting (only ARIES-I11 was discussed), but D-3He was the theme of a recent separate international symposium [2.21], at which the papers presented variously concluded that: (1) sufficient terrestrial 3He exists for an R&D programme up to a commercial power plant; (2) the original estimate of the total lunar 3He resource at - 1 million tonnes remains valid; (3) the mining of lunar 3He should require only modest extrapolations of terrestrial mining techniques; (4) interesting D-3He physics can be explored in next generation tokamak experiments; and (5) the successful development of an advanced fusion configuration, such as an inertial electrostatic fusion device or an FRC, would lead to an attractive D-3He fusion power plant. It should be noted that another D-3He tokamak system study has been done [2.22]. NUCLEAR FUSION, Vo1.34, No.5 (1994) 755

10 TABLE 2.6. REPRESENTATIVE D-3He REACTOR PARAMETERS ApollO-L4 [2.18] ARIES-111 [2.19] first stability second stability tokamak tokamak Artemis [2.20] FRC Net electric power (MW) Fusion power (MW) Neutron wall loading (MW/m*) Major radius (m) Minor plasma radius (m) Plasma elongation Plasma current (MA) Toroidal p Magnetic field on axis (T) Magnetic field maximum (T) Breeder Structure Coolant Outlet temperature ("C) Current drivdheating power Current drive method Pulse length (s) Recirculating power fraction Cost of electricity (mills/kw.h)b lo None HT-9M Water Bootstrap, synchrotron radiation, and compact torus or helicity injection Continuous lo None HT-9M Organic coolant Bootstrap and neutral beam Continuous lo a None Stainless steel To be determined To be determined To be determined Not applicable a Burn chamber radius. Update (in 1992 US $) of published projection Fusion-fission hybrids To utilize fusion energy in the near term, fusionfission hybrid reactors have been considered, mainly in China, since the last meeting in Yalta. Hybrid reactors for the production of fission fuel [2.23] and for the transmutation of high level waste by small tokamak reactors [2.24] have been considered and were presented at the meeting. It was stated that the use of fusion reactors for the production of fuel for fission reactors would be a good way of making them competitive with fission reactors, and that a hybrid reactor would be an appropriate step towards achieving a pure fusion reactor. There are different viewpoints regarding hybrid reactors in various countries Summary Fusion power reactor studies for all the types of confinement schemes presented or discussed at the meeting have been briefly introduced. Tokamak power reactors have been studied in the most detail, but it is encouraging to learn that stellarator reactor studies are also being carried out. Other confinement schemes are in a more conceptual stage. It is hoped that the plasma physics of those concepts with attractive reactor potential will be studied in more detail. Acknowledgements The author thanks Drs M. Abdou, M. Hasan, Lijian Qiu, J. Lyon, R. Miller, J. Santarius, Daikai Sze, M. Tillack and C. Wong for their contribution to the writing of this summary. References to Section 2 [2.1] SEKI, Y., et al., Concept Study of the Steady State Tokamak Reactor (SSTR), Rep. JAERI-M , Japan Atomic Energy Research Inst., Ibaraki (1991). 756 NUCLEAR FUSION, Vo1.34. No.5 (1994)

11 12.51 L2.91 L2.101 [2.11] NAJMABADI, F., et al., The ARIES tokamak reactor studies, Proc. 5th Tech. Comm. Mtg and Workshop on Fusion Reactor Design and Technology, Los Angeles, 1993, Fusion Eng. Des. (1994) (in press). SOKOLOV, Yu.A., Status of the activities on fusion reactor design and technology in Russia, Proc. 5th Tech. Comm. Mtg and Workshop on Fusion Reactor Design and Technology, Los Angeles, 1993, Fusion Eng. Des. (1994) (in press). SEAFP TEAM, Progress on the European Safety and Environmental Assessment of Fusion Power (SEAFP), Proc. 5th Tech. Comm. Mtg and Workshop on Fusion Reactor Design and Technology, Los Angeles, 1993, Fusion Eng. Des. (1994) (in press). INOUE, N., et al., Feasibility study for an inductively operated day-long tokamak reactor, Plasma Physics and Controlled Nuclear Fusion Research 1992 (Proc. 14th Int. Conf. Wiirzburg, 1992), Vol. 3, IAEA, Vienna (1993) 347. TUBBING, B.J.D., et al., On the operation cycle of tokamak fusion reactors, Proc. 5th Tech. Comm. Mtg and Workshop on Fusion Reactor Design and Technology, Los Angeles, 1993, Fusion Eng. Des. (1994) (in press). GRIEGER, G., et al., Stellarator reactor design studies and related technology activities in the European Community, Proc. 5th Tech. Comm. Mtg and Workshop on Fusion Reactor Design and Technology, Los Angeles, 1993, Fusion Eng. Des. (1994) (in press). LYON, J.F., et al., Status of the U.S. stellarator reactor study, Proc. 5th Tech. Comm. Mtg and Workshop on Fusion Reactor Design and Technology, Los Angeles, 1993, Fusion Eng. Des. (1994) (in press). MOTOJIMA, O., SAGARA, A., Natl Inst. for Fusion Science, Nagoya, personal communication, KONDO, S., et al., A conceptual design study of a reversed field pinch fusion reactor, Fusion Eng. Des. 9 (1989) 359. NAJMABADI, F., et al., The TITAN Reversed-Field Pinch Fusion Reactor Study, Rep. UCLA-PPG-1200, UCLA (1990). L2.121 [2.13] [2.14] [2.15] [2.16] [2.17] [2.18] [2.19] t2.201 [2.21] [2.22] [2.23] [2.24] ABDOU, M.A., et al., Nucl. Fusion 26 (1986) MEIER, W.R., et al., Osiris and SOMBRERO Inertial Confinement Fusion Power Plant Designs, Rep. DOE/ER/ , USDOE, Washington, DC (1992). WAGANER, L.M., et al., Inertial Fusion Energy Reactor Design Studies: Prometheus-L and Prometheus-H, Rep. DOE/ER-54101, USDOE, Washington, DC (1992). MIMA, K., et al., Inst. of Laser Engineering, Osaka, personal communication. MOIR, R.W., et al., HYLIFE-11: a molten-salt inertial fusion energy power plant design, final report, Fusion Technol. (1994). DUDZIAK, D.J., et al., US. heavy-ion fusion systems assessment project overview, Fusion Technol. 13 (1988) 207. EL-GUEBALY, L.A., MAYNARD, C.W., Overview of Apollo studies and economic assessment of several proposed variations, Proc. 2nd Symp. on Helium-3 and Fusion Power, 1993, Univ. of Wisconsin, Madison (1993). NAJMABADI, F., et al., The ARIES-I11 Tokamak Reactor Study Final Report, Rep. UCLA-PPG-1384, Univ. of California, Los Angeles (1992). MOMOTA, H., et al., Conceptual design of the D-3He reactor Artemis, Fusion Technol. 21 (1992) Proceedings of the Second Wisconsin Symposium on Helium-3 and Fusion Power, Univ. of Wisconsin, Madison (1993). SHIMOTOHNO, H., et al., Design study of D-3He tokamak power reactor, paper presented at IEEE Symp. on Fusion Engineering, Hyannis, MA, HUANG, J.H., Fusion reactor design and technology program in China, Proc. 5th Tech. Comm. Mtg and Workshop on Fusion Reactor Design and Technology, Los Angeles, 1993, Fusion Eng. Des. (1994) (in press). QIU, L.J., et al., The fusion-fission hybrid reactor conceptual design for transmutation of high level nuclear waste, Proc. 5th Tech. Comm. Mtg and Workshop on Fusion Reactor Design and Technology, Los Angeles, 1993, Fusion Eng. Des. (1994) (in press). 3. PLASMA PHYSICS AND PLASMA ENGINEERING: STATUS AND R&D NEEDS Yu.A. Sokolov The continuing progress in plasma experiments and theory, and in plasma engineering, has made the design of a fusion reactor more credible, with the tokamak being the leading candidate because of its experimental database. The steady improvements in plasma parameters and operation and the successful demonstrations of the new plasma engineering techniques make the fusion reactor design increasingly attractive. In view of the importance of all these topics, this meeting could not provide a forum for the exhaustive presentation of all the plasma physics and engineering progress that has been made; consequently only a limited discussion of these topics took place. A summary of the latest experimental achievements on the route to developing an attractive commercial tokamak fusion reactor, based on the results of DIII-D, was presented by T.C. Simonen and the DIII-D Team. Several promising avenues of tokamak improvement were indicated: stability, confinement, divertor power NUCLEAR FUSION, Vo1.34. No.5 (1994) 757

12 and particle control, steady state RF current drive, etc. Continuous progress has also taken place in achieving some of these goals simultaneously. Advanced tokamak regimes with both improved confinement and improved stability, with improved confinement and impurity removal, and high P operation without disruptions are among the examples. The fusion performance of DIII-D has been doubled every two years, reaching about nrez = (4-5) X 10 m-3.s.kev. The DIII-D device has successfully demonstrated the VH confinement mode (H = 3.6), operation of the plasma core in the of 44% and second stability zone with a central average of about 11 %, H mode operation without impurity accumulation, a bootstrap current fraction of up to 80%, divertor power dispersal with gas puffing, and divertor pumping during the ELMy H mode, resulting in a reduction of density and an increase of temperature. Advanced plasma physics is studied utilizing new engineering techniques which may allow active control of the plasma shape, current density, pressure profiles, plasma rotation and density profile, as well as edge control. A high bootstrap current fraction and localized current drive, disruption control with local heating and current drive, and sustainment of stable profiles are among the means used or planned for DIII-D. Plasma confinement in a fusion reactor must be compatible with MHD stability, which depends on plasma current and pressure profiles, plasma shape and the location of the conducting structure. These quantities have to be optimized to meet the design philosophy of a fusion reactor (e.g. a pulsed or a steady state type of reactor). The analysis of MHD stability was presented by S.C. Jardin and co-workers in a paper [3.1] which summarizes results obtained during the ARIES reactor studies. For a steady state reactor, it is essential from an economics point of view to operate at high poloidal (&),in order to maximize the bootstrap current. For a first stability plasma, this leads to a trade-off between increasing E& to enhance the bootstrap current and reduce the external current drive power and increasing PIE to reduce the toroidal field (E is the inverse aspect ratio). This constraint is somewhat relaxed in the second stability regime, but the accessibility of this regime and stable operation with the suppression of the external kink modes establish other conditions for the current density profile and for the location of the conducting walls (at a distance of about 1.3 times the minor radius). The ARIES-I design, optimized to operate in the first stability regime, has a P value of - 1.9% and a bootstrap current fraction ZBs/Zp for aspect ratio A = 4.5. It requires auxiliary means of current drive. At sufficiently peaked current and pressure profiles, and - elevated central safety factor qo > 2, the ARIES-II/IV configuration is able to reach the second stability region with a moderate P of 3.5% but with ZBs/Zp 1. For ARIES-111, which burns D-3He fuel, the P value has to exceed 20%, which is possible deep into the second stability regime and for only a narrow class of plasma profiles. The strong requirements on P lead to a configuration with ZBs/Zp = 1.3. Bootstrap overdrive has to be compensated for with reversed external current drive. A first stability regime pulsed reactor does not have to operate at high Pp, so it may be able to operate at higher than a corresponding steady state reactor. Auxiliary means require modification of the current profile and preclude pulsed reactors from operating in the second stability regime. Steady state reactors that are optimized to operate at ZBs/Zp = 1, such as ARIES- II/IV, can do this with sufficiently peaked current and pressure profiles and elevated central safety factor qo > 2, and with conducting walls located close to the plasma. With the same reactivity, the ARIES-II/IV design has much less recirculating power, leading to a more economic reactor. Steady state and pulsed operating cycles of tokamak fusion reactors were analysed in the talk of B.J.D. Tubbing and the JET Team. Steady state operation appears to be limited to an advanced concept with low plasma current and advanced physics: higher confinement, a large bootstrap current fraction and a relatively high Troyon factor. The conventional pulsed concept with inductive current drive usually has high plasma current, and its application for a fusion reactor requires thermal energy storage and the augmentation of systems. The issues of fatigue induced by cycling of mechanical and thermal loads also have to be resolved. The advances in physics that may lead to attractive steady state concepts may also lead to attractive pulsed concepts. The development of the tokamak operating regimes at low plasma current would open the way for fusion reactor concepts with a smaller size and a smaller total fusion power that will be better suited for power plants. It was proposed to decrease the burn-off time by operating the tokamak in AC mode. In this mode, plasma current and transformer current are symmetric around zero in successive cycles, and there is no need for recharging of the central solenoid. In JET, AC operation was demonstrated in limiter discharges at 2 MA plasma current. Conventional and advanced reactor concepts operating in pulsed AC 758 NUCLEAR FUSION. Vo1.34. No.5 (1994)

13 mode were proposed. The first has a fusion power of about 7.5 GW and the second 3 GW. The energy deficit for the advanced concept, compared with the conventional, is relatively small GJ versus 2000 GJ. The advanced concept leads to attractive pulsed operation (the low current reduces the inductive flux requirements, the large aspect ratio allows a larger primary coil and a large flux capability, bootstrap current reduces the required loop voltage, etc.). Significant progress has been achieved in the demonstration of advanced regimes on JET and other tokamaks. A high bootstrap current fraction at high 6, was experimentally demonstrated. High enhancement factors for the energy component H up to 3.6 have been observed in the regime with high safety factor and high Pp. Although the results obtained on the advanced regimes are encouraging, the database for these regimes is much smaller than that for conventional regimes. High confinement in bootstrap dominated regimes has to be demonstrated on a time-scale of the resistive diffusion time. The advanced fusion reactor concept usually implies a design with high aspect ratio (A > 4) as compared with current experiments that are taken as having A - 3. In addition, the compatibility of the regime with divertor operation needs to be shown. A comprehensive experimental programme is still needed for both pulsed and steady state operation, and it is premature to make a choice of mode of operation - inductive versus non-inductive - for a tokamak fusion reactor at the present time. Many different methods for plasma heating and current drive can be used in fusion reactors. Such methods as neutral beams and RF, lower hybrid and EC waves are widely used in current experiments. They are also under assessment for fusion reactor applications. Conventional (neutral beams and RF waves) and advanced (synchrotron radiation and helicity injection) techniques have been considered, taking into account the criteria of current drive and overall system efficiency, the engineering and physics database, and system integrability. Because of the general lack of a database and physics models for potentially more advanced current drive techniques, only marginally efficient conventional drivers are usually considered. Steady state plasma operating scenarios were reported for three versions of the ARIES reactor by T.K. Mau [3.2]. On the basis of efficiency, cost, database and reactor integrability considerations, RF waves (ICRH fast waves and lower hybrid waves) are selected as the reference drivers for ARIES-I and ARIES-II/IV. Neutral beam injection is a driver for ARIES-111, which burns D-3He fuel. High bootstrap current operation NUCLEAR FUSION. Vo1.34, No.5 (1994) was proposed in order to minimize the recirculating power and the cost of electricity. To maintain the steady state operation of ARIES-I with a bootstrap current fraction of 68%, 92 MW of fast wave power at 148 MHz is needed. The operation of ARIES-II/IV in the second stability regime leads to a higher bootstrap current fraction (about 95 %) and relaxed requirements on the current fraction sustained with auxiliary reactor systems. In addition, precise local profile control is needed to maintain stability. The total power is 70.2 MW, with 20.5 MW for central current seeding with fast waves and 49.7 MW for current profile control with lower hybrid waves. Because of the higher operational requirements for the ARIES-I11 reactor, with a temperature of 50 kev and density of about 3 X 1020 m-3, neutral beam injection with energy in the 1-6 MeV range is the only viable choice for the current driver. Attaining an acceptable cost of electricity dictates the use of negative ion sources, RF quadrupole accelerators, drift tube linacs and plasma neutralizers, for which extensive R&D is needed. The ARIES studies showed that the extrapolation of the available physics and engineering database for noninductive current drive technology to a steady state reactor is possible. A number of issues have been identified, which include the need for further tokamak experiments, upgrading the theoretical and modelling tools, and technological improvements of the efficiency of the components of the heating and current drive system. A review of the possible reactor applications of EC waves was presented by R. Prater [3.3]. The experimental and theoretical status of the different physics applications of EC waves were discussed. In a fusion reactor, the EC system can be used for preionization and startup, bulk heating to ignition, current drive for steady state operation, and current profile control (for improvement of confinement and /3 limit and for suppression of plasma MHD activity). Reliable experimental demonstration of EC assisted startup will reduce constraints on superconducting coil voltages, electrical resistance of the vacuum vessel, the area of the poloidal magnetic field null and the fill pressure. Effective bulk heating, with well understood physics of wave absorption and propagation of EC waves, can be obtained easily over a broad range of plasma conditions compatible with advanced confinement modes. In spite of marginal efficiency for driving bulk current, the advanced mode of tokamak operation may be achieved with EC current drive control of the current profile. These advanced tokamak regimes can lead to a steady state reactor with reduced size, plasma current and cost. 759

14 The interface of the EC system with the plasma is simple: antennas can be located far from the plasma, very high power density (up to lo3 MW/m2) results in small size antennas, highly efficient transmission lines can be used and EC sources can be placed far from the reactor in a hands on location. Designing a reliable and robust window is still a problem, but a possible solution was demonstrated. The EC system has reduced the constraints on reactor design (i.e. 50 MW heating power can be introduced from a single port, with reduced impact on neutron shielding; there is no need for in-vessel remote maintenance, and hands on maintenance is possible in the source area). The advantages of a simple interface can result in a low cost for the EC system in fusion reactors because of a low indirect cost connected with remote handling and repair, structural supports and neutron shielding, port usage, downtime, the effects on other systems in the case of failure, etc. Arguments against the EC system were also identified: source development is incomplete, power levels in present tokamak experiments are limited and antennas for advanced applications would need to be developed. In the discussion that followed, the potential advantages of the EC system for stellarator reactors were emphasized. The complicated magnetic configuration of a stellarator requires a simplified interface of the heating system with the plasma, and the EC system satisfies this requirement. Stellarators have some intrinsic advantages over the tokamak concept (e.g. steady state operation without need to sustain plasma current, and inherently disruption free magnetic configuration). On the other hand, plasma physics issues and technological problems limit their viability as a fusion device (e.g. the magnetic surfaces, the MHD and neoclassical properties, ripple induced losses of thermal particles, and compensating electric field effects). Improvement in the outlook for stellarator reactors depends on establishing the physics basis needed for reactor extrapolation and on optimizing the magnetic configuration. The papers presented at the meeting were not strictly directed to reporting the current physics status of the stellarator programme, although the status of stellarator reactor studies was reported by J.F. Lyon et al. [3.4] and by G. Grieger [3.5]. Different approaches exist for optimizing the stellarator concept for fusion reactors, and present day stellarators can develop much of the physics basis needed for stellarator optimization. The ATF device was designed to study 0 optimization and confinement improvement through access to the second stability regime, and the Compact Helical System was designed to study stellarator behaviour at low aspect ratio. Stellarators on the scale of DIII-D that have superconducting coils and can demonstrate steady state operation and effective control of particles and impurities with divertors are under consideration. The main goals of the next generation Large Helical Device, with combined shear-magnetic well stabilization, are to demonstrate steady state plasma operation and to achieve reactor relevant plasma parameters with high 0. The HELIAS (helical advanced stellarator) configuration, reported by G. Grieger C3.51, is directed towards developing an advanced stellarator concept by trying to satisfy simultaneously the plasma physics needs of a fusion reactor. For optimization of HELIAS configurations, the following set of criteria has been used: high quality of vacuum field magnetic surface, good properties of finite 0 equilibrium and MHD stability, small neoclassical transport and bootstrap current, collisionless CY particle containment and modular coil feasibility. These criteria were satisfied simultaneously in the Wendelstein 7-X design. The optimized configuration exhibits favourable features with respect to further important aspects of toroidal confinement physics (anomalous transport, and particle and energy exhaust at the edge). References to Section 3 [3.1] JARDIN, S.C., et al., MHD stability regimes for steadystate and pulsed reactors, Proc. 5th Tech. CO. Mtg and Workshop on Fusion Reactor Design and Technology, Los Angeles, 1993, Fusion Eng. Des. (1994) (in press). [3.2] MAU, T.K., Current drive studies for the ARIES steadystate tokamak reactor, Proc. 5th Tech. Comm. Mtg and Workshop on Fusion Reactor Design and Technology, Los Angeles, 1993, Fusion Eng. Des. (1994) (in press). [3.3] PRATER, R., Simplification of reactor systems through use of EC heating, Proc. 5th Tech. Comm. Mtg and Workshop on Fusion Reactor Design and Technology, Los Angeles, 1993, Fusion Eng. Des. (1994) (in press). [3.4] LYON, J.F., et al., Status of the U.S. stellarator reactor study, Proc. 5th Tech. Comm. Mtg and Workshop on Fusion Reactor Design and Technology, Los Angeles, 1993, Fusion Eng. Des. (1994) (in press). [3.5] GRIEGER, G., Stellarator reactor design studies and related technology activities in the European Community, Proc. 5th Tech. Comm. Mtg and Workshop on Fusion Reactor Design and Technology, Los Angeles, 1993, Fusion Eng. Des. (1994) (in press). 760 NUCLEAR FUSION, Vo1.34, NOS (1994)

15 4. FUSION NUCLEAR TECHNOLOGY AND MATERIALS: STATUS AND R&D NEEDS M.S. Tillack 4.1. Introduction The Fourth IAEA Technical Committee Meeting and Workshop in 1986 marked the first time that fusion nuclear technology was included as a separate and distinct topic for discussion, indicating the increased importance and effort given to this expanding discipline [4.1]. Since then, important discoveries and successes have given us increased insight into the behaviour of fusion nuclear components (e.g. first walls and blankets, tritium fuel systems, radiation shielding and plasma facing components) for MFE reactors. These advances have significant implications for the feasibility of fusion reactors; however, much more work is needed to develop components with reliability and safety features that will prove acceptable for ITER, DEMO and future power reactors. In the following sections, the status and remaining R&D needs of fusion nuclear technology and materials are summarized. TABLE 4.1. MAIN MATERIALS OPTIONS FOR SOLID BREEDER POWER REACTOR BLANKETS Coolant Breeder Structure Multiplier Configurations He, water Li,O, LiA102, Li4Si04, LizZr03 Ferritic steel, Sic Be BIT, BOT, pin type, layered type 4.2. Solid breeder blanket status and R&D needs Several materials and design configuration options still exist for solid breeder blankets. The past seven years have been marked by an increasing level of detail in the designs and analyses, rather than a continued widening of the options considered. Table 4.1 summarizes the major materials candidates for the coolant, breeder and structural material. In Japan, layered pebble bed concepts are under development for both ITER and other power reactor applications [4.2], and both water and helium coolant are considered. In Europe, both pin type breeder-in-tube (BIT) and layered breeder-out-of-tube (BOT) options are still under investigation [4.3]. In the USA and Russia, solid breeder blanket design efforts are less focused, and a number of design options still exist. The major classes of issues for solid breeder blankets include: (1) breeder and multiplier tritium inventory, recovery and containment; (2) breedermultiplier-structure mechanical interactions; (3) thermal control; (4) purge flow; (5) structural behaviour, failure modes and reliability; (6) corrosion and mass transfer; and (7) off-normal and accident conditions. Significant results, both theoretical and experimental, have been obtained in many of these areas. There is currently considerable confidence that tritium release rates and inventories will be acceptable. Further R&D is needed to resolve burnup and lifetime effects. Previously, basic thermophysical and mechanical properties were still an issue; however, recent works have moved beyond basic properties and have begun to investigate the effects of irradiation in breeders and Be, and effects of element interactions in submodule configurations. Research and development results have tended to favour pebble beds over sintered block for the breeder and multiplier. Partly because of the greater perceived need for Be and the extensive use of it in ITER (including the first wall armour), Be R&D programmes have flourished in recent years. Studies of fabrication, bonding techniques, tritium release, radiation damage, and thermal and mechanical performance have been undertaken. Most of the remaining concerns for this class of blanket relate to attractiveness (i.e. reliability and performance limits) rather than to feasibility Liquid metal blanket status and R&D needs The R&D issues of liquid metal blankets are to a large degree different from the ones involved in solid breeder blankets. The crucial point for liquid metal breeder blankets, especially for self-cooled concepts, is the large influence of the strong magnetic field on the liquid metal flow. Magnetohydrodynamic effects can cause large pressure drop, influence flow profiles and distribution, and degrade heat transfer. Therefore, an extensive research programme is required to investigate the relevant MHD issues (insulator coatings, flow distribution, heat transfer, etc.), both theoretical and experimental. Other areas requiring R&D work are tritium control (extraction for Li and permeation for LiPb are both discussed in Section 4.6), purification of the liquid metal breeder (corrosion products, Po and Bi for LiPb blankets, and the chemistry of self-healing NUCLEAR FUSION, Vo1.34. NOS (1994) 761

16 coatings), and the behaviour of the liquid metal during transient electromagnetics (i.e. plasma disruptions) Magnetohydrodynamics It is generally agreed now that self-cooled liquid metal blankets need electrical insulators in order to decouple the flowing liquid metal from the load carrying wall. The preferred solution is an electrically insulating coating at the duct wall surface. The development of such coatings is a crucial issue, involving fabrication technology, testing in flowing liquid metal and irradiation tests. The main issues of insulator development are the selection of candidate materials, fabrication technology, imperfections (MHD issues, generation rate, healing, etc.) and radiation induced degradation (damaging irradiation, ionizing irradiation, electric field, etc.) Purijication of the liquid metal breeder Impurities leading to induced activation are a safety and maintenance issue. One kind of impurity is corrosion products, which should be removed continuously in order to avoid the plugging of loop components and to minimize activation. Another kind of impurity is certain trace elements in the liquid metal that will become activated in the blanket. These impurities are of no concern in Li. In the LiPb alloy, the most crucial element is the a emitter 210Po, which is a transmutation product of Bi. Bismuth is both an original impurity occurring naturally in Pb and a transmutation product due to neutron irradiation of Pb. On-line methods have to be developed for the removal of Po and/or Bi from the LiPb alloy Transient electromagnetics Large electrical currents can be induced in the breeding blankets in the case of plasma disruptions that, in connection with the magnetic field, can lead to large forces and stresses in the blanket segments. Liquid metal breeders have the potential to enhance the current flow because of their high electrical conductivity. For the quantitative description of these effects, new methods have to be developed and verified Neutronics status and R&D needs Neutronics activities r during the past decade have helped to reduce uncertainties in predicting the various critical neutronics parameters (e.g. tritium production rate, bulk heating and activation). The US Department of Energy and Japan Atomic Energy Research Institute (USDOE/JAERI) Collaborative Program on Fusion Neutronics is an example of a dedicated research programme on design oriented integral experiments conducted at the Fusion Neutronics Source (FNS) facility at JAERI [4.7]. Tritium production rates (TPRs), induced activity, nuclear heating and in-system spectrum measurements have been performed on LizO test assemblies placed in various geometrical arrangements. The programme consisted of three phases. Other design oriented fusion integral experiments have been conducted at the LOTUS facility at the Ecole polytechnique f6d6rale, Lausanne, Switzerland [4.8], and at Osaka University in Japan [4.9]. In some cases, large discrepancies have been observed between analytical predictions and the experiments themselves. For example, a wide range of the calculatedto-experimental (C/E) values were observed in all the experiments performed in the USDOE/JAERI Collaboration for local TPR (as well as line integrated TPR, a quantity closer to tritium breeding ratio in a blanket). First hand estimates for design margins have been obtained from the observed C/E values for the line integrated TPR that could be as much as -30% more than predicted by calculations. A substantial amount of data was also generated during the programme on radioactivity/decay heat and nuclear heating in ITERrelevant materials. Wide divergence between the computations and the measurements for a number of materials was found, showing the need for updating decay data and cross-sections of the associated libraries used in activation calculations. The JENDL activation file and the REAC3 (175-g) give the best results among the libraries tested. As for heat deposition, nuclear heating rates in several different materials (Li2C03, graphite, Ti, Ni, Zr, Nb, MO, Sn, Pb and W) have also been measured by the microcalorimetric technique. For all these materials, almost all of the C/E values lie in a band extending from 0.5 to 2.0. With regard to shielding capability, a number of bulk shielding integral experiments are currently in progress or planned at the FNS r4.101, the Frascati Neutron Generator [4.11], and the Technical University at Dresden [4.12] using a 14 MeV neutron source. The objective of these experiments performed on various shield thicknesses is to quantify design margins for the protection of superconducting coils and to give realistic estimates for peaking factors due to radiation streaming and the existence of opening paths. 762 NUCLEAR FUSION. Vo1.34, NOS (1994)

17 4.5. Plasma facing component status and R&D needs Most of the work on plasma facing components (PFCs) has been undertaken in conjunction with plasma confinement experiments (e.g. at JET, TFTR, Tore Supra, TEXTOR). Because of this, the emphasis has been on providing components that perform satisfactorily at zero neutron fluence over short periods of time and very low availability. The technology for long pulse operation at reactor relevant power densities does not exist. Fusion nuclear technology R&D for PFCs is still in its infancy and is driven primarily by next generation devices that include significant nuclear phenomena. Since PFCs still have critical feasibility issues, a much expanded R&D programme is acutely needed. Divertors tend to receive the most attention, as they are subjected to the most challenging array of conditions, including very high heat fluxes, high particle fluxes and intense neutron radiation. The most popular reactor divertor designs usually incorporate an actively cooled substrate and a separate plasma facing armour material. These are sometimes referred to as duplex structures. The armour may be bolted to the substrate, brazed or attached by some other high conductivity joint. Maintaining acceptable thermal stresses in a compliant joint, while still maintaining high thermal conductivity and lifetime, is a major challenge. The ability to rapidly replace the armour with in situ maintenance operations may be required if armour lifetimes cannot be increased. Several armour materials, including Be, various blends of graphite, high Z refractory metals (such as W), and composites of Be and C, are candidates. The choice of armour material depends primarily on plasma performance issues. The actively cooled substrate design functions primarily as a high heat flux heat exchanger. Different coolants are still under consideration, including water, He, liquid metals and even gassolid suspension flows. One option, which has received low level attention for many years, is the free surface liquid metal concept, in which a liquid metal is directly exposed to the edge plasma. Specific engineering issues and R&D needs for PFCs are: (a) Thermal hydraulics pe$ormance. In some cases, heat removal rates of 20 MW/m2 or higher are required for divertor plates. The capability to efficiently remove heat from the large surfaces of the impurity control system requires innovative design solutions. Thermomechanical behaviour. Divertor plates that are subjected to high heat fluxes and a high number of cycles also will be subjected to high thermal stresses, possibly exceeding the yield point of materials, Fracture mechanics and fatigue are important concerns. Plasma-su$ace interactions. Erosion and redeposition will alter the surface characteristics and cause macroscopic thinning of the plasma facing material. The behaviour of these modified surfaces under high heat flux and stresses is unknown. Tritium permeation and retention. Tritium transport is important for safety reasons, and also as an element in the overall tritium fuel cycle. Large amounts of tritium may become trapped in PFCs or transported into the coolant systems. Radiation damage effects. Over a long period of time, the effects of surface erosion and neutron radiation will strongly influence the divertor lifetime. Response to both normal and off-normal transients. In particular, surface erosion and electromagnetic forces during disruptions could severely reduce the operating life of PFCs. Advanced divertor designs. Alternative designs, such as liquid metal droplet divertors, have been suggested. Such approaches would eliminate the normal divertor plate. Many of the engineering concerns with standard plate designs would be eliminated, but new concerns will arise and the alternative concepts will need to be tested. Besides the divertor, other plasma interactive components also need to be carefully considered. For example, RF launcher arrays are large (- 1.5 m X 2 m), complex components very close to the plasma that require active cooling. Nuclear effects can have a significant impact on their performance, including electromagnetic spectrum degradation, reduced plasma coupling and thermomechanical problems Tritium processing systems status and R&D needs Tritium processing systems are in a more advanced stage of development than are other areas of fusion nuclear technology research. There are two fundamental reasons for this: (a) Tritium processing research can be based on the research of weapons programmes in the EC, Russia and the USA. NUCLEAR FUSION, Vo1.34. N0.5 (1994) 763

18 (b) Most of the work on tritium research does not require 14 MeV neutrons. All of the main parties in fusion research have major research programmes on tritium processing, for example the Tritium Process Laboratory in Japan, the Tritium Laboratory Karlsruhe in Germany and ETHEL (European Tritium Handling Experimental Laboratory) in Ispra. Also, considerable tritium expertise exists in Canada from the CANDU programme and in France from its weapons programme. Major activities must also exist in China and Russia from their weapons programmes, but the level of effort devoted to fusion is not certain. The method for the separation of hydrogen isotopes is cryogenic distillation, which has been well demonstrated, and computer codes are available to calculate the size, cost and tritium inventory of the unit under different operating conditions. The key problem is to minimize the tritium inventory of the system, while at the same time to maintain the system s operating stability. There are different problems associated with tritium recovery from the different types of blanket: (1) Solid breeders. On the basis of the information generated, it can be concluded that tritium can be recovered from solid breeders with a reasonable tritium inventory if the proper temperature of the solid breeder and the proper chemical conditions of the purge gas can be maintained. The purge channel is usually a small channel about 10 m long and is embedded into the solid breeder region. With radiation and thermomechanical effects, the dimensions of the solid breeder will certainly change. It is difficult to foresee whether, under those conditions, the purge channel will remain open for an extended period of time. (2) Liquid Li. The tritium solubility in the Li is very high. Thus, it is very difficult to remove the Li to a tritium concentration of - 1 appm. The molten salt tritium recovery method has been the only process to demonstrate the recovery of tritium to this concentration. However, salt will dissolve in the Li, which causes concern over corrosion effects of the salt on the structural material. Cold trapping may be an effective method of removing the salt from Li, but it has not been demonstrated. Recently, distillation has been proposed to recover tritium from Li, but the distillation system requires large power (- 100 MW) and high temperature operation ( l10ooc). There has been no experimental demonstration that distillation can be used to recover tritium on a large scale. (3) Liquid LiFb. The problem associated with tritium recovery from LiPb is very different from that for Li. While the key concern of tritium recovery from Li is to reduce the tritium inventory, the key problem associated with tritium recovery from LiPb is to reduce the tritium partial pressure. A permeation window is an effective way to recover tritium from LiPb. However, the resulting tritium partial pressure is still rather high. This partial pressure is acceptable if the reactor temperature is low and an effective tritium diffusion barrier can be established. The key goals for the tritium system R&D are: (1) to ensure that tritium is available at the composition required, (2) to minimize the tritium inventory, and (3) to minimize the tritium partial pressure for permeation. All of these have to be accomplished with an acceptable cost of the system. The key R&D directions for tritium systems are: For the tritium processing system: Development of control techniques that will ensure stable operation of the cryogenic distillation systems with minimum tritium inventory and with both steady state and pulsed input from the plasma. For solid breeders: Expansion of the R&D from pure chemistry to an engineering system to ensure the tritium can be recovered with the proper purge rate and chemistry, as well as purge configuration. For liquid Li: Demonstration of tritium recovery, including tritium inventory and the effect of the tritium recovery process on the reliability of the blanket. For liquid LiFb: Demonstration of the tritium recovery process with a residual tritium partial pressure over the LiPb that is acceptable for the blanket condition typical of a power reactor. Materials development status and R&D needs 4.7. I. Status of materials development for fusion The fusion radiation environment causes swelling, irradiation creep and degradation of thermophysical properties. Coolant, tritium and breeder material compatibility issues include corrosion and mass transfer, hydrogen embrittlement and degradation of thermomechanical properties. A database of structural materials response to various loads is essential to determine allowable tensile, fatigue, crack growth, creep and fracture toughness characteristics. Elevated temperature operation of PFCs pushes the limits of 764 NUCLEAR FUSION. Vo1.34. NOS (1994)

19 materials performance driven by time dependent deformations (elevated temperature and irradiation enhanced creep). The toroidal geometry of tokamak fusion power plants requires complex component structures, which raises issues of fabrication, welding and maintainability. The first step in new materials development is establishing the feasibility of candidate materials. Until 14 MeV neutron test facilities become available, the first step may not be fully realized. The second step entails microstructure tailoring to minimize the effects of radiation damage. Lastly, structural materials design qualifications have to be developed for licensing large scale production of materials. Currently, design codes such as the American Society of Mechanical Engineers (ASME) pressure vessel codes are used to evaluate materials performance. These design codes cover materials, construction, method of fabrication, inspection and safety devices. For a material to be approved by ASME, it must first be approved by the American Society for Testing and Materials (ASTM). The data required for approving a base metal and welded joints include: ultimate strength, yield strength, reduction in area, elongation, strain fatigue, creep strength, notch toughness, stress rupture strength, brittle fracture characteristics, neutron irradiation effects, and service experience, if available, preferably over a range of temperatures. The extensive databases needed to develop design codes must be generated for fusion power plants. In order to provide code validated licensable materials, this programme should evolve in several steps. For example, once candidate materials have been identified, it takes from 15 to 25 years to develop a metallic structural material according to the following logic: (a) Demonstration of feasibility 5 years (b) Development of fusion specific materials 10 years (c) Development of damage correlations 10 years (d) Development of design methodology (codes) 10 years Steps (b) and (c) can overlap; however, step (d) requires results from step (c) to develop component design methodologies similar to the ASME boiler pressure vessel codes. Radiation databases generally exclude other environmental conditions, such as coolant interactions, welds, mechanical joints, effects of damage gradients on large structures, fatigue and thermal cycling. These are generally developed during the (c) and (d) steps of materials development. Fusion materials R&D involves several international collaborations [4.13], including the ITER programme, several bilateral agreements, the International Energy Agency fusion materials agreement and informal collaborations. The ITER programme dominates intemational fusion materials development programmes (concentrating on structural materials and special function ceramic materials) for blanket and shield, PFCs, heating and current drive, and diagnostics. First wall and blanket related efforts concentrate on establishing properties of irradiated austenitic stainless steels ( C and up to 20 dpa) to study environmentally assisted crack growth, and rewelding irradiated austenitic steels. In the area of advanced blanket module R&D efforts, irradiation tests on ferritic/martensitic steels and V alloys are planned. For divertor applications, Cu alloys strengthened by oxide dispersion are being irradiated and Nb alloys are being assessed. Ceramics for diagnostic and plasma heating systems are studied by in situ measurements of the effects of ionizing and displacing radiation on electrical and optical properties. In summary, commercially available materials will not be able to satisfy the long term performance requirements of fusion power plants. The stringent plasma operating conditions impose near zero failure tolerances on large components inside the plasma chamber. An aggressive materials R&D programme with a much wider scope is required to achieve successful DEMO operation by the year A 14 MeV neutron source with adequate testing volumes would have to be in place by the year Integrated blanket testing in a 14 MeV neutron environment is essential for component development R&D needs for structural materials and ceramics Structural materials Major candidate structural materials include austenitic steels (e.g. 316SS), ferritidmartensitic steels (e.g. modified HT-9), V alloys and Sic-Sic composites. While the austenitic steels have limited DEMO relevance because of high temperature thermomechanical property limitations, these materials have the largest database for ITER use at fluences up to 10 dpa. The ferritic alloys must demonstrate adequate ductile to brittle transition temperatures after irradiation, and magnetic permeability issues need to be resolved. Furthermore, although ferritics have a significant fabrication database, welding techniques for irradiated materials, which contain high He concentrations, need NUCLEAR FUSION, Vo1.34. No.5 (1994) 765

20 to be developed. Vanadium alloys show promise; however, there has been only limited industrial experience, with few commercial suppliers. Also, tritium solubility in V alloys is an issue. Sic-Sic composites have the lowest known activation and show promising high temperature properties. However, radiation damage and performance feasibility have not been demonstrated, and cost and manufacturing issues are unresolved. Non-structural ceramics Issues include induced electrical conductivity, dielectric breakdown, radiation induced electrical degradation, and optical absorption and luminescence. There are relatively few data available on neutron damage effects, which may possibly cause major problems for ITER. Breeder materials Concerns with breeder materials include thermomechanical behaviour and chemical compatibility. While neutron interaction with Li ceramic breeder materials has been well documented, module performance still has to be demonstrated. For flowing selfcooled Li or LiPb, neutron interaction is not among the most critical issues; however, neutron stable electrical insulating coatings inside coolant tubes have to be developed. Divertor/PFCs Among the structural issues are degradation of thermal conductivity, embrittlement, tritium retention, duplex materials and structural integrity. The primary materials under consideration are Be, graphite, high Z refractory metals, Cu, Ni and Sic-Sic composites. The issues are too numerous to list here. Superconducting magnets Critical current reduction and insulation failure are the most crucial issues for superconducting coils. Superconductors, such as NbTi and NbsSn, have low radiation tolerances, and ceramic insulators cannot yet be specified with confidence Safety status and R&D needs Since the last IAEA Technical Committee Meeting and Workshop, planning within the world fusion community has become focused on the requirements of ITER. The needed technical data in the field of safety can be divided roughly into' four categories: 4.8. I. Tritium transport A wide range of experiments have been performed on the solubility and permeation of tritium in unirradiated materials. However, few materials samples have been irradiated in fluxes like those that will exist in fusion reactors. The tritium transport through these radiation damaged materials may be significantly different than that through the unirradiated material. Codes for the modelling of tritium migration pathways are fairly well developed. TMAP4, which models the transport of tritium through several in-reactor and external pathways, has been verified with respect to the correctness of its coding and validated by comparison with experimental data [4.14]. At the Kernforschungszentrum Karlsruhe, W. Raskob has developed UFOTRI [4.15] for modelling the atmospheric and biological transport of tritium. Barriers to hinder tritium permeation could be of great benefit, particularly in preventing tritium migration through heat exchanger surfaces. The development of durable barriers is a major R&D need Activation products For ITER, the release of activated first wall and blanket materials is the major pathway for radiation exposure of the public. Past experiments have shown that the material released from an alloy, such as stainless steel, is quite different from the original constituents of the alloy. Some elements, such as Mn, are released preferentially owing to the formation of volatile oxides. Experiments have been performed in which heated samples of candidate first wall materials are exposed to air or steam [4.16]. The oxidation and release of the constituent elements and their release rates are used to estimate the off-site doses following a reactor accident. To date, all of the experiments have been performed using unirradiated materials. The activation products which would accumulate in the irradiated material have been simulated by the inclusion of stable isotopes in the sample coupons. Better modelling of activation product behaviour will require either irradiated samples or powder metallurgy techniques to obtain the required sample constituents. Radiation damage will also undoubtedly alter the mechanisms for release of activation products from the samples. The transport of vapours and aerosols out of a damaged reactor will be reduced through the plate-out 766 NUCLEAR FUSION. Vo1.34. No.5 (1994)

21 of vapours on cold surfaces and through aerosol settling. Data on the rates of plate-out and settling will allow safety analysts to more confidently bound the doses to the general public in the case of a fusion reactor accident Dose codes Codes have been developed to model the dose at the site boundary and to the surrounding population from the release of tritium and activation products. The MACCS code [4.17], originally developed to calculate doses due to fission product releases, has been expanded to include all significant activation product isotopes [4.18]. Transport through the food-chain is modelled by the COMIDA code [4.19]. The EC has developed the COSYMA code for accident consequence analysis [4.20]. These codes calculate doses on the basis of the effective dose equivalent, in keeping with international standards Regulatory framework Most members of the worldwide fusion community would agree that direct application of the regulatory framework now used for fission reactors will not reflect the inherently different characteristics of fusion. However, in the absence of positive action on the part of the fusion community, those fission regulations will be applied by default. The ITER facility will be the first for which these issues of regulatory framework must be addressed. Since there already exists a regulatory framework in most countries for the construction and operation of experimental nuclear facilities, that framework will serve as the starting point for the licensing of ITER, and deviations from that framework will be sought when they appear to be justified. In the USA, DOE order 5480.FUS, for the regulation of fusion facilities, is being developed Summary The past several years have seen significant progress on several aspects of fusion nuclear technology. The R&D to date has focused on low cost, high leverage items that can resolve critical issues and help in concept screening. Many of the remaining tasks will require much greater effort and resources to develop attractive components for power reactors. As we proceed into the era of ITER, fusion nuclear technology is expected to play an increasingly important role in the world fusion programmes. Fusion development pathways have become a major topic of concern relative to fusion nuclear technology. Neutron test facilities are mandatory for fusion nuclear technology R&D, and will be the major driver of cost and timing for the future of fusion development. While ITER is a highly desirable step, it is generally agreed that one or more complementary facilities will be required prior to DEMO, The need for a fusion materials test facility has been clearly documented in the past. A new emphasis is being placed on complementary component testing in VNS facilities. Acknowledgements to Section 4 The author thanks Drs S. Herring, S. Malang, S. Sharafat, Daikai Sze and M. Youssef for their essential contributions to this summary. References to Section 4 Fusion Reactor Design and Technology 1986 (Proc. 4th Tech. Comm. Mtg and Workshop Yalta, 1986), 2 vols, IAEA, Vienna (1987). SEKI, Y., Fusion reactor design and technology programme in Japan, Proc. 5th Tech. Comm. Mtg and Workshop on Fusion Reactor Design and Technology, Los Angeles, 1993, Fusion Eng. Des. (1994) (in press). PROUST, E., et al., Breeding blanket for DEMO, Proc. 17th Symp. on Fusion Technology, Rome, 1992, North-Holland, Amsterdam (1993) GANESAN, S., MUIR, D.W., FENDL Multigroup Libraries, IAEA-NDS-129, IAEA, Vienna (1992). YOUSSEF, M., et al., Fusion integral experiments and analysis and the determination of design safety factors, Part I. Methodology, Fusion Technol. (in press). MAEKAWA, H., OYAMA, Y., Experiment on angular flux spectra from lead slabs bombarded by D-T neutrons, Fusion Eng. Des. 18 (1991) MAEKAWA, H., ABDOU, M.A., Overview of latest experiments under the JAERI/USDOE Collaborative Program on Fusion Neutronics, Fusion Eng. Des. 18 (1991) HALDY, P.A., et al., Experimental program at the LOTUS facility, Fusion Technol. 10 (1986) 962. SUMITA, K., et al., Status of OKTAVIAN I and proposal for OKTAVIAN 11, Nucl. Sci. Eng. 106 (1990) MAEKAWA, F., et al., Status of shielding experiments for ITER, Part 2. Analysis, Proc. Int. Mtg on Fusion Neutronics, Los Angeles, 1992, Rep. UCLA-FNT-60, ENG-93-15, UCLA (1992) 195. PILLON, M., et al., A plan of fusion neutron benchmark experiments using the Frascati Neutron Generator (FNG), Fusion Eng. Des. 18 (1991) ELFRUTH, T., et al., First intermediate report: specification and optimization of experimental set-up and procedure, Proc. Int. Mtg on Fusion Neutronics, Los Angeles, 1992, Rep. UCLA-FNT-60, ENG-93-15, UCLA (1992) 298. NUCLEAR FUSION, Vo1.34. No.5 (1994) 767