Synergies between FNT developments and advanced nuclear fission technologies

Fusion Engineering and Design 81 (2006) 1667–1674

Synergies between FNT developments and advanced nuclear fission technologies Edgar Bogusch a,∗ , Mihaela Ionescu-Bujor b , Alain Chevalier c , Bogdan Bielak d a

EFET/FRAMATOME ANP GmbH, Freyeslebenstr. 1, 91058 Erlangen, Germany b Forschungszentrum Karlsruhe GmbH, Germany c EFET/NNC Limited, Germany d EFET/FRAMATOME ANP SAS, Germany

Received 17 February 2005; received in revised form 1 August 2005; accepted 1 August 2005 Available online 10 January 2006

Abstract The development of future fusion power plants will face similar development issues as future generation of fission plants in order to achieve the goals of technical feasibility and operability. Except for plasma physics, the major challenges for fusion reactors are in the areas of materials development for the heat source structures (plasma facing material for fusion and materials for the fission core) and design of cooling systems for high efficiencies. Helium cooling systems have been proposed for both future fusion plants and advanced fission reactors, in particular those for GEN IV programme. They offer the potential of high efficiencies in combination with advanced high-temperature resistant materials. Therefore, synergies for fusion and fission reactor development could be realized for development in these two areas. For example, the testing of major components of the helium cooling systems for the two power plant systems, as well as the non-nuclear testing of materials, could be possible within common test loops in which different test sections will have to be integrated. The identification of possible synergies in the relevant R&D programmes should be endorsed to minimise the development effort for future power plants. However, the assumed time schedules for the realisation of future fusion and fission reactors have to be taken into account. © 2005 Elsevier B.V. All rights reserved. Keywords: Fusion power plants; GEN IV reactor concepts; Helium loops; Advanced structural materials

1. Introduction

∗ Corresponding author. Tel.: +49 9131 189 5314; fax: +49 9131 189 5098. E-mail address: [email protected] (E. Bogusch).

0920-3796/$ – see front matter © 2005 Elsevier B.V. All rights reserved. doi:10.1016/j.fusengdes.2005.08.029

Following the construction and successful operation of ITER the development and construction of fusion power plants for electricity production and other applications is planned. In contrast to ITER, future fusion power plants will be cooled by liquid metal or helium to achieve higher

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Fig. 1. Time schedule for future fission/fusion reactor generation.

plant efficiencies, as described in the ongoing European Power Plant Conceptual Studies (PPCS). Based on existing know-how, relevant cooling systems will have to be designed and manufactured according to R&D programmes including testing of the main components and related plant cooling systems. Another main focus will be the development of high-temperature materials resistant to elevated temperatures up to or above 1000 ◦ C. Within the GEN IV programme among the future fission reactors under development, two concepts, i.e. gas-cooled fast reactor (GFR) and very hightemperature reactor (VHTR) encounter in some areas quite similar development needs to those of future fusion reactors like the PPCS concepts. In addition, medium-term concepts like the pebble bed modular reactor (PBMR), the gas-turbine modular hightemperature reactor (GTMHR) and ITER also have some areas where development is still underway. Our analysis will concentrate on these similarities and this paper will comment possible synergies of the technological development related to the two reactor system key design issues, namely: cooling systems and materials. It should be mentioned that other areas for possible synergies could be waste management, decommissioning, safety approach, etc. Concerning cooling systems, the common issue for the reactor systems quoted above (GFR, VHTR, PBMR, GTMHR, PPCS B and C concepts) is the use of helium as coolant. In the past, nuclear fusion concept designs benefited from the developments made for fission reactors

in terms of materials (SS316LN for ITER) and cooling system design (water cooling systems for ITER and helium cooling systems design for DEMO). As the development of fusion continues following successful operation of ITER, the development of future fission power plants may take advantage of developments from fusion and vice versa in specific areas. Therefore, the development of future fusion and fission plants should be carried out with as close interaction as possible (see Fig. 1) giving mutual benefit for construction and operation of both power plant systems.

2. Advanced power plant concepts 2.1. GEN IV fission reactors In 2002, a “Technology Roadmap for Generation IV Nuclear Energy Systems” was published [1] describing the goals and the development needs for six advanced fission reactor concepts beyond the present fission reactors such as EPR and HTR. The six fission reactor concepts encompass a gas-cooled fast reactor system, a lead-cooled fast reactor system, a molten salt reactor system, a sodium-cooled fast reactor system, a supercritical-water-cooled reactor system and a very high-temperature helium-cooled reactor system. For the objective of this paper the gas-cooled fast reactor system (GFR) and the very high-temperature reactor system (VHTR) have been selected to be analysed with respect to the possible synergies with fusion reac-

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tor developments. In addition, synergies have already been defined for VHTR and GFR concepts which will share common material, component and cooling system development programmes. Commercial operation of these fission reactor plants is expected to start 2030 (see Fig. 1). The reference GFR is a 600 MWth /288 MWe Hecooled reactor operating at an outlet temperature of 850 ◦ C using a direct Brayton cycle gas turbine for high thermal efficiency. The GFR is primarily designed for electricity production and actinide management with the potential for hydrogen production. The reference VHTR has a 600 MWth heliumcooled core and is primarily aimed at high-temperature process heat applications but cogeneration for electricity production is also envisaged. The VHTR has coolant outlet temperatures above 1000 ◦ C. The VHTR requires significant development in fuel performance and high-temperature materials. However, it may benefit from earlier gas-cooled thermal reactor developments. The material R&D needs include hightemperature alloys, fiber-reinforced ceramics or composite materials. 2.2. Future fusion reactor concepts The European Power Plant Conceptual Study [2] studied four power plant models for commercial fusion power plants (see Table 1) starting operation around 2040. The models cover a range from near-term devel-

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opments (Models A and B) including limited technology (and plasma physics) extrapolations to advanced technologies (Models C and D). Consequently they differ significantly in terms of performance, design and materials selection whilst having the same electrical output of 1500 MW. In contrast to fission reactors, fusion reactors have two main heat sources, namely the breeding blanket and the divertor. Both components may operate at different temperature levels and have different material requirements. Models C and D will be considered further as being most comparable with GEN IV fission reactor concepts considered here due to their helium cooling systems. Model C has a lithium–lead blanket which is cooled by helium having an outlet design temperature of 480 ◦ C. The structural material is EUROFER. The divertor uses EUROFER as structural material protected by tungsten plasma facing plates. The helium outlet temperature of the divertor of Model C is 600 ◦ C. Model D has a liquid breeder blanket which uses the breeder material simultaneously as the primary coolant. The secondary coolant loop uses helium at design temperatures of about 1000 ◦ C. Therefore, process heat applications could be possible as for the VHTR. The proposed structural material for Model D is mainly SiC/SiC composite. A summary of features comparing PPCS models with GEN IV reactors is given in Table 2.

Table 1 Comparison of PPCS reactor models Parameters

Model A

Model B

Model C

Model D

Fusion thermal power (MWth ) Electric power output (MWe ) Plant efficiency (%)

5000 1500 31

3600 1500 37

3410 1500 42

2530 1500 60

Materials Breeder Blanket structure Divertor structure Divertor armour

Pb–17 Li EUROFER CuCrZr W

Li4 Si 04 EUROFER W alloy W

Pb–17 Li EUROFER W alloy W

Pb–17 Li SiC/SiC SiC/SiC W

Coolant Blanket Blanket coolant temperature (◦ C), inlet/outlet Divertor Divertor coolant temperature (◦ C)

Water 285/325 Water 140/167

Helium 300/500 Helium 540/720

LiPb/Helium LiPb: 480/700, helium: 300/480 Helium 540/720

LiPb 700/1100 LiPb 600/990

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Table 2 PPCS technologies similar to Generation IV fission reactors Gas-cooled fast reactor system—relevant synergies with Model B. He coolant, He loops and Turbomachinery technologies (490–850 ◦ C He temperature ranges) Very high-temperature reactor—relevant synergies with Models B and C. High temperature, He loops and heat exchangers (1000 ◦ C) for thermochemical H2 production Supercritical-water cooled reactor—Model A. Model A may be further optimised for high pressure water cooled blankets and divertor operating above thermodynamic critical point of water (374 ◦ C, 22.1 MPa) Lead-cooled fast reactor—Model D. Material technology, heat transport, energy conversion and chemistry for lead (PbLi/PbBi) coolant systems

3. Development needs and possible synergies 3.1. Materials Future fusion power plants represent a great challenge to structural materials development and operational behavior. Operating temperatures, neutron exposure levels and thermo-mechanical stresses are comparable or even greater than those for proposed GEN IV fission reactors. While pulsed thermal loads may play a minor role, the different neutron energy levels will have to be taken into account. The EU fusion materials development programme is focused in the short term in developing the reduced activation ferritic martensitic (RAFM) steel EUROFER, the structural material for the EU ITER test blanket modules (TBMs) of the helium cooled lithium lead (HCLL) and the helium cooled pebble bed (HCPB) blanket concepts [3,4]. Fusion operated materials like EUROFER must fulfil all blanket-related design and operational requirements and must be fully code qualified for low- and high-temperature applications in all mechanical properties. The material development for future fusion power plants considers the replacement of EUROFER with advanced oxide dispersion strengthened ferritic steels (ODS) made by dispersing nano-sized particles of oxide (yttrium oxide being the first choice), and thereafter with the development of fiber-reinforced silicon carbide (SiC/SiC) which would represent a real alternative [4].

Divertors are the other fusion components with high material development needs. Gas cooling is proposed as in the blankets. A coolant temperature of 600–650 ◦ C implies operating temperatures in the structure of 650/700 ◦ C and up to 1200/1300 ◦ C in the high heat flux removal elements. Tungsten in combination with a supporting structure of advanced ferritic (ODS) steel has been selected since it is expected that it will meet such requirements. Advanced ODS steels and SiC material could be tested under operational conditions in ‘second generation’ breeding blankets in DEMO. The use of low activation materials that can operate at higher temperatures than RAFM steels may lead to further improvements in power plant efficiency/economics and in waste management. Therefore, long term EU R&D is focused on ODS steels. Following EUROFER type ferritic–martensitic ODS steel will be developed, that, because of its similar thermo-physical properties, can be combined with conventional EUROFER. Ferritic ODS steels, investigated in a second step, could form the structural material of a gas (or liquid-metal) cooled divertor aiming at an operational window in excess of 700 ◦ C. In addition, SiC/SiC composites should be developed for a temperature window in the range of 600–1100 ◦ C. A limited but steadily increasing effort is devoted to characterize and improve tungsten alloys. The development of tungsten as a structural divertor material is extremely appealing. Because of its excellent thermal properties (high conductivity and low expansion coefficient), the upper operating temperature could be as high as 1200/1300 ◦ C. However, a lower operational window in the order of 650–750 ◦ C could be required to cope with ODS steel as structural material for the cassette bodies [4]. Tests are indispensable in order to validate the computational models and codes for the DEMO design (neutronics, tritium production, inventory and recovery; MHD effects; thermo-hydraulics and thermomechanics; electromagnetics) and to qualify them in a fusion environment as well as for investigating potential showstoppers for DEMO application. GEN IV fission reactors have the need to develop materials with superior resistance to fast-neutron fluence under very high-temperature conditions [1]. The main challenge are the in-vessel structural materials which have to withstand fast-neutron dam-

Table 3 Candidate materials for GEN IV reactor concepts operating in different temperature ranges of primary cooling systems [1] Temperature range

GEN IV reactor system

Corresponding fusion reactor concept

Candidate structural materials

Limiting phenomena

Low (<350 ◦ C)

Integral primary system reactors Simplified BWRs (boiling water reactor) Evolutionary pressure tube reactors High conversion LWRs (light water reactor)

Power Plant Conceptual Studies (PPCS) fusion reactor—concept A

Ferritic pressure vessel steels

Radiation embrittlement (toughness, DBTT)

Fe-base austenitic stainless steels

SCC, IASCC, high dose embrittlement

Ni-base stainless alloys and superalloys Zirconium alloys

IGSCC, IG corrosion, weld metal SCC, IASCC

Intermediate (∼350–600 ◦ C)

Supercritical LWRs thermal and fast Sodium-cooled LMRs

Ferritic pressure vessel steels

Lead/lead–bismuth cooled LMRs (liquid metal reactor)

Ni-base austenitic alloys and superalloys

Fe-base austenitic stainless steel

Ferritic–Martensitic alloys

Intermediate to high (∼600–900 ◦ C)

High (>900 ◦ C)

Lead–lead–bismuth cooled LMRs

Power Plant Conceptual Studies (PPCS) fusion reactor—concepts B and C

Iron and nickel-based superalloys

Molten salt fuelled reactors

Ferritic/martensitic alloys

Prismatic gas-cooled reactors

Refractory metal alloys

Pebble bed gas-cooled reactors

Ceramic composites and coatings

Gas-cooled fast reactors Very high-temperature gas-cooled reactors

Graphite Presently, only tungsten and molybdenum systems as metal based systems and SiC/SiC as ceramic composite system are believed to have the potential to operate in this temperature range

Creep strength, swelling and embrittlement, corrosion, IASCC He embrittlement creep strength, swelling and embrittlement, corrosion, IGSCC, IASCC Radiation embrittlement (toughness, DBTT) corrosion, IASCC, hydrogen cracking, corrosion in led-based coolants and molten salts Creep behavior, toughness, He embrittlement

Creep behavior, toughness, radiation-induced embrittlement, corrosion, in lead-based coolants and molten salts dispertion stability in ODS alloys Creep behavior, toughness, radiation-induced embrittlement, corrosion, oxidation, impurity pick-up Creep behavior, radiation and environmental effects on interfaces, toughness, corrosion, in lead-based coolants and molten salts Creep strength, swelling toughness, thermal conductivity Concerning metal-based systems, similar issues, as above, for intermediate to high temperature reactor systems (in particular temperature creep and creep rupture). Concerning SiC/SiC, differential swelling at the fiber–matrix interface under irradiation and helium production during nuclear transmutation.

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Molten salt cooled reactors Gas fuelled reactors

Power Plant Conceptual Studies (PPCS fusion reactor—concept D

Radiation embrittlement (toughness DBTT), IGSCC, IASCC, hydrogen embrittlement Radiation embrittlement (toughness, DBTT)

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Ferritic/martensitic alloys

Corrosion, hydryding

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age and high temperatures, up to 1600 ◦ C in accident conditions. Ceramic materials are reference options for in-core materials while metal alloys are proposed for out-of-core structures. Other structural materials such as Ni–Cr–W super alloys that will be required in particular for the VHTR. In addition, materials need to be developed for temperatures exceeding 1000 ◦ C in case of process heat application. The major programme goals for fusion and fission are the same. They comprise: • fabrication and welding/joining; • physical, neutronic, thermal, tensile, creep, fatigue, and toughness properties and their degradation and the reactor relevant range for neutron flux and dose; • microstructure and phase stability under irradiation; • irradiation creep, in-pile creep and swelling properties; • compatibility with helium. R&D is recommended to study and quantify mechanical performance and dimensional stability properties in particular under irradiated conditions. For the range of service conditions expected in Generation IV systems, including possible accident scenarios, the proposed materials must meet design objectives in the following areas: • dimensional stability, including void swelling, thermal creep, irradiation creep, stress relaxation, and growth; • strength, ductility, and toughness; • resistance to creep rupture, fatigue cracking, and helium embrittlement; • neutronic properties for core internals; • physical and chemical compatibility with the coolant; • thermal properties during anticipated and off-normal operations; • interactions with other materials in the systems. For each design objective, the fundamental microstructural features that establish performance (such as dislocation microstructure, void microstructure, precipitate microstructure, and radiation-induced segregation) must be understood to allow for further performance improvements. The formation and behavior of these features depend on materials temperature and neutron flux and spectrum. An additional objective is to limit impacts of neutron acti-

vation of components, which can complicate maintenance, handling, and disposal of irradiated components, through careful selection of material constituents. Candidate alloys for the 300–600 ◦ C temperature range include austenitic iron- and nickelbase alloys, ferritic/martensitic alloys and oxidedispersion strengthened ferritic and austenitic alloys. The primary materials candidates for 600–900 ◦ C range are those with good strength and creep resistance at high temperatures, such as oxide-dispersion strengthened ferritic–martensitic steels, precipitatestrengthened iron- or nickel-base superalloys, coated materials, or refractory alloys of molybdenum, niobium, and tantalum. Materials issues for applications at temperatures exceeding 900 ◦ C become increasingly severe. The potential limitations of metallic alloys at higher temperature motivate consideration of ceramic materials. The extreme temperatures also present challenges for conducting experiments in existing irradiation facilities. To realize the goal of fission reactor core-outlet temperatures higher than 1000 ◦ C, new metallic alloys for reactor pressure vessels have to be developed. At these core-outlet temperatures, the reactor pressure vessel temperature is likely to exceed 450 ◦ C. Further development of Ni–Cr–W superalloys and other promising metallic alloys will be required for the VHTR. The irradiation behavior of these superalloys at the service conditions expected in the VHTR will need to be characterized. Such work is expected to take 8–12 years and can be performed at facilities available worldwide. As it can be derived from the presentations above, similar structural materials are proposed for conceptual designs of GEN IV and fusion reactors (Table 3) such as complex ferritic–martensitic steels, superalloys, tungsten alloys and SiC/SiC composites [5]. Taking into account the different neutron spectra developments could be harmonised to limit the development effort. In addition, GFR type reactors offer the possibility of materials testing for fusion to establish confidence to high neutron damage doses. 3.2. Helium heat transfer systems The relevant power plants (GFR and PPCS Model C) use high-temperature helium cooling systems with direct Brayton cycle (5–8 MPa and up to 850 ◦ C (Model C 700 ◦ C) outlet temperature). For the VHTR and

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Model D the helium temperature will be elevated to 1000 ◦ C and above for possible process heat applications. However, synergies between the concepts can still be preserved. The design of the helium heat transfer system requires particular designs to meet the loads resulting from high temperatures which in turn are required to achieve high efficiencies. Special techniques have also to be developed for the auxiliary systems. Consequent testing of loop components is required to prove the designs. For that, loops to test different components for the different reactor types could be constructed for partial or global qualification. One particular challenge will be the development of direct cycle helium turbines for temperatures of 800 ◦ C and above (Model D) and other key components such as recuperators and high-temperature heat exchangers. The possible synergies described in the GEN IV R&D plan for GFR and VHTR could be expanded to PPCS Models C and D. The development would rely on the helium technology development, in particular on generic topics such as tribology, tightness, thermal insulation, purification, etc. The major results of these developments could be used for all systems. However, the possibly higher tritium content in the fusion reactor loops compared to fission reactor loops has to be taken into account as a fusion specific issue. In order to qualify various components and technologies, several helium test facilities are planned to be constructed or upgraded. For VHTR technology, CEA plans the construction of a pilot plant for verification of plant control systems, operability, and materials performance. A part of this pilot plant is a helium loop (HELITE) of 1 MW, coupled to a non-nuclear heat source. HELITE has been designed and will be commissioned in 2006. In parallel, purification systems and specific instrumentation have been developed and are planned to be tested in HELITE during period 2006–2007. The operation of HELITE has been envisaged between 2006 and 2010 for performing thermomechanical and thermo-hydraulic tests. In the period 2009–2010, after the HELITE program will be finished, CEA intends to build a 20 MW helium facility (HELLO) to test large components and to simulate normal and abnormal operating transients of gas-cooled systems with direct conversion. Larger helium test facilities are planned for operation in 2019. In the period 1994–1996, ENEA built the HEFUS3 facility, which has been used to carry out the thermo-

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mechanical tests on module subassemblies of the European Helium Cooled Blanket designs selected for DEMO, which are to be tested in the ITER reactor. In 2004, it has become apparent that two helium facilities would be needed in order to test the two European blanket concepts for the DEMO reactor: the helium cooled pebble bed (HCPB) and the helium cooled lithium lead (HCLL). Therefore, the European Fusion Program includes (i) the construction of a new facility (HELOKA) at FZK, Germany, for testing the HCPB concept and its cooling system and (ii) the upgrade of the HEFUS3 facility, for testing the HCLL concept. The temperature and pressure profiles in HELOKA are consistent with those expected during ITER operations and the loop components and the assembly technologies (procedures) will ensure a leak-tightness of 10−6 l mbar/s for the whole system. Under normal operating conditions, the loop will be able to operate during extended periods of time with an operating cycle adapted to the ITER operation schedule. In addition abnormal operating conditions (LOFA, LOCA) will be simulated and examined in order to define the proper control scenarios for such situations. HELOKA is planned to be commissioned in 2007. Both facilities (HELOKA and HEFUS3) will be able to test 1:1 mockups of the test blanket modules, before their installation in ITER. Even though the above-mentioned facilities have different goals, they have many technological similarities. The main common characteristics are as follows: (1) they are high-pressure loops (HELITE and HELLO—7 MPa, HEFUS3—8 MPa, HELOKA— 10 MPa); (2) they are “8-shape” loops, using a recuperator to transfer a part of the heat from the hot leg of the loop into the cold leg; (3) the temperatures in the hot leg are above 500 ◦ C (HELITE—1000 ◦ C, HELLO—850 ◦ C, HEFUS3—530 ◦ C, HELOKA—700 ◦ C). Therefore, the technical solutions concerning materials to be used for pipes and components will be similar and the experience gained at one of these loops could be used also at the others. Currently, no materials are capable of withstanding the pressure of 10 MPa at temperatures above 550 ◦ C for an extended lifetime. There are two possible solutions:

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(1) monitoring the creep of the materials and replace regularly the equipment (pipes, etc.) or (2) using heat-insulating materials inside the tube to reduce the temperature in the pipe material (more expensive). The main differences between the loops are related to the respective helium mass flow rates: HELITE—0.4 kg/s, HELLO—10.4 kg/s, HEFUS3— 0.35 kg/s and will be upgraded to 1.2 kg/s, HELOKA—1.8 kg/s and will be upgraded to 5.5 kg/s. These differences imply different choices concerning major components (e.g. heat exchangers, pumps). Concerning circulators, a possible solution could be the use of several identical circulators in parallel, in order to achieve higher mass flow rates. Here, again, the experience gained could be shared among the researchers at the various sites. A major step for the realisation of synergies could be the amendment of the HELOKA loop for HTR/VHTR helium cooling systems components development as presently under discussion.

4. Conclusions The development of future fusion power plants and of GEN IV fission reactors offer the possibility to take benefit of possible synergies in their R&D programmes. Among others, the development of materials and the design of cooling systems using the same medium (e.g. helium) are of particular interest. Specific low activation steels for structures under high irradiation and temperatures developed for fusion may be of interest also for advanced fission plants. Already in the EXTREMAT programme synergies in the hightemperature materials field will be identified consider-

ing both fission and fusion. In turn the experience from technologies developed for helium-cooled fission reactors may be of benefit for fusion. In particular helium test loops will be required for both systems. Sharing of these loops taking into account the different operating conditions may reduce the development effort. Looking at the time schedules for the development of fusion and advanced fission plants, synergies in all relevant technology areas should be observed with potential savings in time and cost. Acknowledgements The authors would like to thank B. Friedrich, Framatome ANP GmbH, and E. Diegele, EFDA, for their valuable input and comments, in particular for the materials aspects. References [1] A Technology Roadmap for Generation IV Nuclear Energy Systems, GIF-002-00, Issued by the U.S. DOE Nuclear Energy Research Advisory Committee and the Generation IV International Forum, December 2002. [2] A Conceptual Study of Commercial Fusion Power Plants, Final Report of the European Fusion Power Plant Conceptual Study, EFDA-S-RE-Draft, April 2005. [3] E. Diegele, R. Andreani, et al., European Fusion Materials Research Program—Recent Results and Future Strategy, Proceedings of the 16th ANS TOFE, Madison, WI, September 14–16, 2004, in press. [4] R. Andreani, E. Diegele, R. Laesser, B. van der Schaaf, The European Integrated Material and Technology Programme in Fusion, J. Nucl. Mater. 329–333 (2004) 20–30. [5] Workshop on Advanced Computational Materials Science, Application to Fusion and Generation IV Fission Reactors, ORNL/TM-2004/132, Washington, DC, March 31 to April 2, 2004.