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Trends in nuclear systems design and analysis
Fusion Engineering and Design 18 (1991) 187-194 North-Holland
187
Trends in nuclear systems design and analysis G. Casini
Commission of the European Communities, Joint Research Centre/Ispra Establishment, Institute for Systems Engineering and Design, 21020 Ispra (VA), Italy
The papers presented at the poster session "Nuclear Systems Design and Analysis" are reviewed and trends discussed. The attention is focussed on the following areas: (i) Reactor Studies and (ii) Neutronic Tests and Analysis. Present studies on D-T power reactor conceptual designs should be encouraged. Recent conceptual studies on D-3He systems seem adequate to identify the critical issues. The economic feasibility of 3He supply from the moon should be assessed. Contributions on Neutronics are found to be of particular relevance and represent an important step towards the implementation of data base for nuclear design of plasma facing and breeding blanket components.
1. Introduction
This lecture is an introduction to the poster session VII of the ISFNT-2 Conference. The papers presented in this Session can be subdivided according to four subjects: - Fusion Power Reactor Studies ( D - T and D - 3 H e ) - Design Studies in Support to Next Step Machines (ITER, NET) - Neutronic Tests and Analyses - Others. This review will be focussed on the first three subjects. On these bases, trends for the development of Reactor Design and Neutron Analysis activities will be discussed.
2. Power
reactor
studies
The largest number of papers presented is dealing with Commercial Power Reactor Studies. The papers refer basically to two types of fuel cycle, namely D - T and D-3He. All contributions regard tokamak type of reactors.
2.1. D - T power plants Four reactor concepts are discussed, namely SSTR, ARIES-I, T A R T R and CFAR. SSTR [1, 2 and 3] is a Japanese conceptual design of a steady state tokamak reactor. The SSTR design is based on the present day physics data base (in particular the recent observation
in JT-60 of a large boostrap current), with a small extension of technologies which can be realized in about ten years from now. Main relevant features discussed in the papers at the conference are: - a low resistance vacuum vessel integrated with the shield in order to form a double thin wall structure - the reduction of the peak surface heat flux on the divertor plates by gaspuffing and Fe addition. With this hypothesis, the authors show a divertor design, water cooled, which calls for a limited extrapolation of the present cooling technology. ARIES-1 [4] is an American conceptual design of a steady state tokamak power reactor plant. At the conference the progress on the engineering aspects of the reactor core are outlined, namely the use of low activation silicon carbide (SIC) composites as structural material and advanced design solutions for the internal components satisfying a high level of safety and the criteria for shallow land burial. The present knowledge in fabrication and mechanical properties of SiC-composites developed in various industrial applications appears encouraging. The effects of neutron irradiation are poorly investigated. The main concern is related to the high helium generation rate which could create problems in the mechanical behaviour of the material. The design proposed for first wall, blanket and divertor structures is based on the experience in manufacturing large integrated pieces in the aerospace industry. The use of lithium zirconate as breeding material for blanket and tungsten as protection material for divertor creates problems from induced radioactivity point of view. In order to achieve the required low activation
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level for standard operation, accident and waste disposal conditions, the authors propose isotopic tailoring, which however for the blanket seems not sufficient. The question of the economical feasibility of the isotopic tailoring is not discussed. T A R T R [5,6] is an English concept of a tokamak power reactor with tight aspect ratio (l.5 and less) consistent with the retention of a central toroidal field conductor, of copper type, without inboard breeding blanket and divertor. Two aspects of this concept have been analysed and discussed, the neutronic properties and the toroidal field coil design. Tritium breeding requirements are satisfied without the need of an in board blanket. The main difficulties appear those related to the radiation effects on the conductor, in particular the resistivity increase, even in presence of a thin radiation shielding. The provision of a toroidal field system looks feasible but the use of a single turn coil leads to very high power supply current requirements. The possibility of using multiturn conductors is bounded to the resistance of insulators under irradiation. In any case the concept looks attractive for its small size and best use of neutrons. C F A R [7,8] is a conceptual study of a power reactor station with an advanced blanket using the compact Rankine cycle, as originally proposed by B.G. Logan. In this concept synchrotron radiation is used both for MHD-electricity conversion and current drive. Two specific studies are reported on high performance non-equilibrium disk-type M H D generator, microwave superheater, neutronics, tritium production and losses. The interest of mixing lithium to the superheated mercury in the blanket is discussed. First corrosion tests are also mentioned. The results of the study look promising, even if, in particular concerning the tritium breeding blanket design and tritium losses, the analysis is not yet sufficiently detailed. To conclude the review of the D - T power reactor studies one has to mention an interesting study of the impact of ports on the breeding performance of different types of blankets in the European D E M O N E T configuration [9]. Starting from 3D Monte Carlo analyses it is shown that it is possible to obtain an estimation of the blanket port effect on the basis of the reduced blanket coverage. Typically, the tritium breeding decrement is in the range of 7-10% in case of 10 outboard blanket ports, depending on type of breeding blanket (solid or liquid). It is worth to mention a paper not dealing with a specific power reactor design but with the strategy for fusion development [10]. The author discusses the interest of what he calls a "fusion neutron reactor",
namely a reactor which exploits the fact that fusion is neutron rich as compared to fission. Such a reactor could bc of interest as a tool to produce fuel for fission reactor plants and to burn long term radioactive elements (actinides) of fission wastes, as well as to test thc behaviour under neutron irradiation of materials which are of interest for fusion power reactors. The advantage would be that of not requiring to reach ignition: however, to be effective, such a type of reactor should have to operate in steady state conditions. The qucstions related to the complexity of such a system as compared to the present Next Step (ITER, NET), are not answered; neither thosc rclated to the economical interest of producing fucl for fission plants by this way as compared to other techniques (uranium cnrichmcnt, extraction from spent fuel, use of accelerators, etc.). 2.2. D - ~ H e power plants
In the area of advanced fuel cycle studies the attention is focused on D - 3 H e . A concept of a tokamak commercial reactor is proposed by the University of Tokyo [ll]. The basic physics assumption is that it operates in the second stability regime of plasma. This would reach /3-values of about 20% so one could operate at low magnetic field (12 T) and relatively low plasma current (9 MA). In these conditions also the other relevant parameters, namcly the synchrotron radiation and total wall loading would be acceptable. A second contribution concerns the interest of using an organic coolant and reduced-activation-steel in the in-vessel components of ARIES-III [12]. The organic coolant can operate at high temperature (425°C) and low pressure (2 MPa), then an efficient power conversion cycle is obtained. Radiolytic decomposition rate of organic liquids, which was the main potential drawback in case of fission reactor plants, should be acceptable here due to the low neutron production ratc. 2.3. Trends
The two conceptual designs SSTR and ARIES-1 correspond to different philosophies in assessing the features of commercial fusion power plants. Indeed, as already mentioned, the Japanese effort has been that of evaluating what are the prospectives of fusion power reactors in the hypothesis of a moderate extrapolation of the present knowledge both on plasma physics and on nuclear technology. On the other side, ARIES-1 considers very advanced solutions for the magnets and structural materials of in-vessel components. However both rely on a large boostrap current (70-80%) and on
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high aspect ratio (4.0-4.5) which leeds to relatively low plasma currents (10-12 MA) and high magnetic fields (16-20 T). The importance of these assumptions is shown when comparing the main parameters of these two machines with those of EEF [13], the European power reactor design whose main parameters, defined in 1989, were based on the existing physics data base without major extrapolations (table 1). For structural materials both SSTR and EEF consider martensitic (ferritic) steels. Their behaviour under neutron irradiation up to the fluences expected in power reactors remains an open problem. In the EEF study it was considered, as an alternative to martensitic steels, vanadium-base alloy. Such alternatives should be kept also in the next design studies. One has to note that the use of composites, as proposed in ARIES-l, introduces a large flexibility in the design which has been proved usefull in other technological branches. The main problem seems to be the behaviour of SiC and fibres under neutron irradiation at the level of fluence expected in power reactors. The solution proposed in ARIES-1 of isotopic tailoring the divertor materials seems less realistic. However in the next years the knowledge on the the physics of the plasma edge will be improved, so one can expect new ways to spread-out
the particle and heat fluxes on the collector plates of the divertor. This would facilitate the choice of materials more favourable from radioactive point of view than tungesten. The possibility of getting the high magnetic fields proposed in ARIES-1 is also still open. However the recent development of new conductors in some laboratories makes this assumption reasonable. The TARTR study belongs to a different category as compared to the previous ones. Indeed it calls for radical innovations which still require confirmation in particular on plasma physics and on the behaviour of materials under neutron irradiation. However, in our opinion, this concept merits further investigation due to its potential capability for easy extrapolation of the present technology and low capital cost. In the area of advanced power reactor systems it is interesting to note that for some time conceptual studies have focused on D - 3 H e fuel cycle and tokamak configurations. Recent designs as APOLLO (1988), ARIES-III in USA and now the Japanese one are usefull to identify the main critical issues and the areas where to address the research effort. The trend is now to consider plasma operation in the second stability regime. This adds another uncertainty to the several ones, both on physics and technology, which character-
Table 1 Major parameters of commercial D-T tokamak reactor Type Configuration Plasma major radius (m) Plasma half width (m) Plasma elongation Aspect ratio Plasma current (MA) Toroidal filed on axis (T) Peak filed on TF coil (T) Mean electron density (102°/m3) Plasma temperature (keV) Current drive method Boostrap current fraction (%) Current drive power (MW) Peak neutron wall loading (MW/m 2) Coolant Breeder Structure Thermal power (MW) Net electrical output (MW)
EEF (1989)
SSTR
ARIES-1
DN 5.31 1.40 2.0 3.78 16.6 6.2 14.9 1.45 20 NBI 50 91 4.1 water liquid (Pb-17Li) ferritic steel 3780 1200
SN 7.0 1.7 1.85 4.1 12.0 9 16.5 1.4 17 NBI 75 60 5.0 water ceramic (Li20) ferritic steel 3710 1080
DN 6.75 1.5 1.6 4.5 10.2 11.3 21.0 1.5 20 ICRF fast wave 68 100 3.5 He ceramic (Li2ZrO3) SiC composite 2544 1000
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ize this type of power reactor concept. Further specific studies on single aspects of the design are not recommended. On the contrary it seems mandatory to make an evaluation, even if inevitabliy approximate, of the expected costs of the fuel supply from lunar soil, based on the operations and processes identified by NASA in 1988. 3. D e s i g n a n a l y s i s in s u p p o r t to t h e N e x t S t e p
The papers presented are dealing with the neutron effects on various components around the plasma, namely: - Neutral Beam Injectors, - Pumping Ducts, - Windows for Electron Cyclotron Resonance Heating (ECRH), - Magnets. The 2D-analysis of the I T E R neutral beam (NB) system shows that the beamline components will be exposed to an important neutron flux, 10%10 ~ n / c m 2 [14]. On the other side, the induced radioactivity in the NB-room is sufficiently low to consider the corresponding waste of "very low" level. Similar results are obtained for the I T E R plasma pumping ducts, analysed by the 3D Monte Carlo code TRIPOLI-2 [15]. With an appropriate bending of the ducts, an attenuation of the neutron flux up 10 ~' on the concrete wall in front of them can be obtained. In another contribution [16], calculations show that degradation of the dielectric and thermal conductivity properties of ceramic (AI203) foreseen for gyrotron windows of E C R H are unacceptable even at moderate neutron fluence (1019 n/cm2). This is a further confirmation of the difficulty one has to expect in the design of plasma facing components which involve insulating material. The paper on neutron effects on magnets [17] shows that in the case of austenitic steels as structural materials, the possibility of avoiding remote handling after shut-down would require an optimization of the inboard shielding materials and an increase of about 10 cm of the shielding thickness which is in contrast with other design and economic aspects. A last paper is devoted to a sensitivity analysis of the effect of taking energy self-shielding into account in radiation shielding calculations for NET [18]. As response parameter, the total nuclear heating in the inboard coil is taken. Variations of up to 13% are found when self shielding is included and up to 10% due to uncertainties in the angular distribution of the secondary neutrons.
4. N e u t r o n i c tests a n d a n a l y s e s
4.1. Integral experiments in blanket configurations All contributions in this area are related to the J A E R I / U S D O E collaborative programme on fusion neutronics. This programme is based on the utilizatkm of the Japanese fusion neutron source (FNS) for integral experiments of increasing complexity on configurations simulating a blanket with LifO as breeder material. Up-to-now three phases of the collaborative programme have been realized. At the conference the results of the Phase-Ill are presented together with a review of the overall set of experimental results and analysis. Three papers [19,2(/,23] are reporting an overview and detailed experimental results and comparisons with calculation of Phase III programme. In these measurements a line source was simulated by cyclic movement of the annular test assembly relative to a stationary point source that was located axially at the center of the inner cavity. The test assembly was including a 1.5 cm thick stainless steel first wall, a 20 cm thick L i f o zone and a 20 cm thick LizCO 3 zone. The simulation of the line source has been achieved through two modes of operation, namely stepwise and continuous mode of operation. Measurements included tritium production rate from ~'Li and 7Li, in-system spectrum and various foils activation. Different codes have been considered for the analysis, namely DOT 5.1 and D O T 3.5, Monte Carlo codes MCNP and MO RSE in conjunction with E N D F / B - V and J E N D L data files. Tritium production rate from 7Li is generally predicted within 10-15% while for "Li it is within 10-20Ci. Tritium production rate from natural Li is well predicted (5-10~/,). The agreement with foil activation measurements is within 10-20%. Two other papers [22,23] make the point on the capability now achieved in USA and Japan to predict tritium production and other reaction rates in fusion systems on the basis of the experience acquired during the three Phases of the collaborative neutronics programme. The total number of experiments was near to 20, with different incident neutron source conditions, point and line source, and experimental arrangements of the test assembly, ranging from a simple, one-breeder zone to a more prototypical blanket that included stainless steel first wall, beryllium and coolant channels. In all cases the breeder material was Li,O. In the USA analysis the calculated-to-experimental values for the various parametcrs are treated statistically. The prediction uncertainty of tritium production from ¢'Li is ranging between 9% to 5%, based on
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deterministic and Montecarlo results respectively. For 7Li the uncertainties are higher (10.4% and 7.5% respectively). It is intersting to note that, if the results of Phase-1 which involved simplest geometrical and material characteristics are excluded, the prediction uncertainty is reduced. In the JAERI analysis an evaluation of the experimental accuracy obtained in the various Phases of the test programme is made. It appaears a continuous progress in the definition of the experimental configurations and in the measurement techniques. This fact explains why, when the Phase-I experimental results were removed in the U S A analysis, the calculated to experimental ratios were improved. The calculation accuracy for reaction rates including tritium production in the overall set of experiments is found to be less than 5-10%. In the first Phase a systematic discrepancy on the tritium production cross section of 7Li was found which could be explained by using an upgraded data library for the analysis. Also in case of configurations involving beryllium a systematic effect was observed. It is concluded that there are still uncertainties in the back scattering and secondary energy spectrum in the low energy range of Be cross sections. Another paper related to the U S D O E / J A E R I collaborative programme [24] concerns direct nuclear heating measurements and analysis. Microcalorimetric technique has been employed for on-line measurements. Small probes of several materials which are candidates for fusion applications have been irradiated in close vicinity of the rotating target. Temperaturechange and heat deposition rates as low as 30 /xK/s and 35 ~tW/g respectively, have been measured. These experiments are analysed by MCNP code and by comparing various heating coefficients and kerma factor libraries. Large deviations from the experimental data and discrepancies when using different libraries are found. The same applies to the comparison between measured and calculated spatial profiles of heat deposition. A last contribution [25] in the frame of J A E R I / U S D O E collaboration is dealing with the analysis of measurements of activation of several elements (20) of interest for fusion. The paper deals with the comparison with the code T H I D A which is including an updated activation cross section library. As compared to previous analyses made by the same authors, the accuracy of the code prediction is increased in the high energy range ( > 1 MeV), whereas below this threshold large discrapancies still exist. One serious problem raised by the authors is that of inadequate prediction in the products of (n, y) reactions.
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4.2. Integral experiments on single material
The question of Be-nuclear data is analysed in two papers [26,27] which present measurements of neutron transport and multiplication in integral experiments on Be-spherical shells. The comparison between E N D F / B VI (USA) and EEF-I (EC) based on neutron leakage spectra measurements shows remarkable differences in spectra and neutron multiplication. There are indications that the 14 MeV (n, 2n) cross section of Be in E N D F / B - V I is too low and the secondary energy distribution of the EFF-1 neutron emission cross-section at low secondary energies is too high. Three papers from JAERI report on neutron spectrum and flux spactial distribution in integral experiments using test assemblies of tungsten, lead and iron. In the first one [28] Tungsten was found to give the largest neutron attenuation among various materials as Fe, Be, C and concrete. The comparison with codes ( D O T 3.5/JENDL-3) enabled to identify a possible understimation of the cross section of inelastic or (n, 2n) reactions in the data library. In case of lead [29], the angular neutron flux spectra from slabs were measured by the time-of-flight method. The agreement between calculations (JENDL-3) and experiments is in general good in the energy range investigated (0.1-15 MeV). Some minor changes are proposed in the Pb (n, 2n) cross section libraries. In the test with iron [29], neutron spectra in a large cylindrical assembly, using proton recoil gas proportional counters are measured. The comparison with codes (MCNP/JENDL-3) gave in general good agreement, but the discrepancies at higher energies ( > 100 keV) become larger, up to 30%, at positions far from the source. The novelty of the experiment lies on the fact that it investigates low energy range ( < 1 MeV) typical of radiation shielding and magnet conditions, for which no data were previously available. In another paper y-heating measurements by a proportional counter have been made in a low-Z assembly ( M g / C ) irradiated with 14.8 MeV neutrons [30]. The neutron-induced signal was separated from the gamma-induced signal by exploiting the signal size-time rate differences inherent to radiations of different Linear Energy Transfer (LET), which are observable in a proportional counter. The experiment was modelled using the one-dimensional radiation transport code A N I S N / P C and ENDFB-V. To complete the panorama one has to mention the announcement of a plan of neutron shielding experiments using the Frascati Neutron generator (FNG)
[32].
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4.3. T r e n d s
5. Conclusions
The amount of results on neutronic tests presented at the Conference is impressive and the trend towards the establishment of a complete set of data for neutronic design studies looks appropriate. The largest set of experimental data has been obtained at the Japanese Fusion Neutron Source (FNS), in the frame of the J A E R I / U S D O E Collaboration Programme on Fusion Blankets. Among them, one has to first highlight the importance of the Nuclear Heating and Induced Radioactivity measurements because they represent, the unique systematic experimental approach to the implementation of data which are vital in the design and safety of fusion reactors. In particular one has to stress the advanced microcalorimetric technique used for online nuclear heating rate measurements. As one would expect, due to the novelty of the experimental data, the first comparisons with the available codes show a large spread in the calculated versus measured heat deposition values. A further effort, even involving other Countries out of Japan and USA, is highly recommended. A similar approach is being pursued by JAERI and UCLA in the area of activation calculations. The analysis presented is involving the most important elements responsible for neutron activation in fusion materials. Here again even if remarkable improvement with the existing data base has been already obtained, a continuation of the effort remains mandatory. Moving to breeding blanket tests, the capability of predicting tritium production from 6Li and 7Li looks now rather satisfactory, at least when Monte Carlo Codes are used in the analysis. The accuracy is around 5%, a figure which fits well with the usual safety margins taken in neutronic design evaluations. In case of beryllium, the way of measuring in a clean singlematerial geometry neutron leakage spectra, leakage multiplication factors and absorption rates, as proposed by KfK, looks as the best way to check the cross-reaction data of this important element. The discrepancies still existing in the most recent nuclear data libraries, as E N D F B / V I and E E F / I , show that a definitive clarification is not yet achieved. On the other hand, the question of how to exploit at the best the complementarity of the homogeneous experiments with the heterogenous ones (as in the J A E R I / U S D O E case) merits further investigation. For other materials the way of setting-up clean-geometry, single-medium integral experiments of the type presented at the Conference should be pursued.
The progress and prospectives of activity, in the areas here reviewed can be summarized as follows: - D-T
Power Reactor Studies
Conceptual design studies of commercial power stations of Tokamak type are underway in Japan and USA. Similar work is expected in the next period in Europe. Even if the hypotheses taken in physics and technology are often quite different, these studies represent a valid forum to identify the prospects of D - T fusion. The present tendency towards high aspect ratios, low plasma current and high magnetic field should be exploited. The investigation of the alternative solution of a reactor with tight aspect ratio should also be pursued. - D-3He Power Reactor Studies
The recent effort in the evaluation of the main aspects of power plants of tokamak type based on this fuel cycle seems adequate to identify the critical design issues. An assessment of the economic implications of 3He supply from the moon is needed. - N e u t r o n i c Tests a n d A n a l y s i s
The experimental information from neutronic integral tests on materials used in plasma facing components, breeding blankets, shielding and magnet materials has remarkably grown in the last period. The recommandation would be that of pursuing along the way of experiments which are as clean as possible, concerning homogeneity and geometry, and that of using Monte Carlo Codes for their analysis. In general, neutronics seems one of the technology areas where the effort is being pursued in the most satisfactory way.
References
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[28] Y. Ikeda, K. Oishi, C. Konno, T. Nakamura, Neutronics experiment and analysis on a tungsten slab assembly bombarded with D - T neutrons, ISFNT-2, Karlsruhe 1991, Fusion Engrg. Des. 18 (1991) 3(19-315, in these Proceedings, Part C. [29] H. Maekawa and Y. Oyama, Experiment on angular neutron flux spectra from lead slabs bombarded by D - T neutrons, 1SFNT-2, Karlsruhe 1991, Fusion Engrg. Des. 18 (1991) 287-291, in these Proceedings, Part C. [30] C. Konno, Y. Ikeda, K. Kosako, Y. Oyama, H. Maekawa, T. Nakamura, E.F. Bennet, Measurement and analysis of
low energy neutron spectrum in a large cylindrical iron assembly bombarded by D--T neutrons, 1SFNT-2, Karlsruhe 1991, Fusion Engrg. Des. 18 (1991) 297-3(13, in these Proceedings, Part C. [31] H.K. Chiu, E.F, Bennet, B.J. Micklich, A technique tk)r measuring gamma heating in a low fluence mixed radialion field, presented at ISFNT-2, Karlsruhe 1991. [32] M. Pillon, A. Santamarina, A plan of fusion neutron benchmark experiments using the Frascati Neutron Generator (FNG), ISFNT-2, Karlsruhe 1991, Fusion Engrg. Des. 18 (1991) 293-296, in these Proceedings, Part (7".
187
Trends in nuclear systems design and analysis G. Casini
Commission of the European Communities, Joint Research Centre/Ispra Establishment, Institute for Systems Engineering and Design, 21020 Ispra (VA), Italy
The papers presented at the poster session "Nuclear Systems Design and Analysis" are reviewed and trends discussed. The attention is focussed on the following areas: (i) Reactor Studies and (ii) Neutronic Tests and Analysis. Present studies on D-T power reactor conceptual designs should be encouraged. Recent conceptual studies on D-3He systems seem adequate to identify the critical issues. The economic feasibility of 3He supply from the moon should be assessed. Contributions on Neutronics are found to be of particular relevance and represent an important step towards the implementation of data base for nuclear design of plasma facing and breeding blanket components.
1. Introduction
This lecture is an introduction to the poster session VII of the ISFNT-2 Conference. The papers presented in this Session can be subdivided according to four subjects: - Fusion Power Reactor Studies ( D - T and D - 3 H e ) - Design Studies in Support to Next Step Machines (ITER, NET) - Neutronic Tests and Analyses - Others. This review will be focussed on the first three subjects. On these bases, trends for the development of Reactor Design and Neutron Analysis activities will be discussed.
2. Power
reactor
studies
The largest number of papers presented is dealing with Commercial Power Reactor Studies. The papers refer basically to two types of fuel cycle, namely D - T and D-3He. All contributions regard tokamak type of reactors.
2.1. D - T power plants Four reactor concepts are discussed, namely SSTR, ARIES-I, T A R T R and CFAR. SSTR [1, 2 and 3] is a Japanese conceptual design of a steady state tokamak reactor. The SSTR design is based on the present day physics data base (in particular the recent observation
in JT-60 of a large boostrap current), with a small extension of technologies which can be realized in about ten years from now. Main relevant features discussed in the papers at the conference are: - a low resistance vacuum vessel integrated with the shield in order to form a double thin wall structure - the reduction of the peak surface heat flux on the divertor plates by gaspuffing and Fe addition. With this hypothesis, the authors show a divertor design, water cooled, which calls for a limited extrapolation of the present cooling technology. ARIES-1 [4] is an American conceptual design of a steady state tokamak power reactor plant. At the conference the progress on the engineering aspects of the reactor core are outlined, namely the use of low activation silicon carbide (SIC) composites as structural material and advanced design solutions for the internal components satisfying a high level of safety and the criteria for shallow land burial. The present knowledge in fabrication and mechanical properties of SiC-composites developed in various industrial applications appears encouraging. The effects of neutron irradiation are poorly investigated. The main concern is related to the high helium generation rate which could create problems in the mechanical behaviour of the material. The design proposed for first wall, blanket and divertor structures is based on the experience in manufacturing large integrated pieces in the aerospace industry. The use of lithium zirconate as breeding material for blanket and tungsten as protection material for divertor creates problems from induced radioactivity point of view. In order to achieve the required low activation
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level for standard operation, accident and waste disposal conditions, the authors propose isotopic tailoring, which however for the blanket seems not sufficient. The question of the economical feasibility of the isotopic tailoring is not discussed. T A R T R [5,6] is an English concept of a tokamak power reactor with tight aspect ratio (l.5 and less) consistent with the retention of a central toroidal field conductor, of copper type, without inboard breeding blanket and divertor. Two aspects of this concept have been analysed and discussed, the neutronic properties and the toroidal field coil design. Tritium breeding requirements are satisfied without the need of an in board blanket. The main difficulties appear those related to the radiation effects on the conductor, in particular the resistivity increase, even in presence of a thin radiation shielding. The provision of a toroidal field system looks feasible but the use of a single turn coil leads to very high power supply current requirements. The possibility of using multiturn conductors is bounded to the resistance of insulators under irradiation. In any case the concept looks attractive for its small size and best use of neutrons. C F A R [7,8] is a conceptual study of a power reactor station with an advanced blanket using the compact Rankine cycle, as originally proposed by B.G. Logan. In this concept synchrotron radiation is used both for MHD-electricity conversion and current drive. Two specific studies are reported on high performance non-equilibrium disk-type M H D generator, microwave superheater, neutronics, tritium production and losses. The interest of mixing lithium to the superheated mercury in the blanket is discussed. First corrosion tests are also mentioned. The results of the study look promising, even if, in particular concerning the tritium breeding blanket design and tritium losses, the analysis is not yet sufficiently detailed. To conclude the review of the D - T power reactor studies one has to mention an interesting study of the impact of ports on the breeding performance of different types of blankets in the European D E M O N E T configuration [9]. Starting from 3D Monte Carlo analyses it is shown that it is possible to obtain an estimation of the blanket port effect on the basis of the reduced blanket coverage. Typically, the tritium breeding decrement is in the range of 7-10% in case of 10 outboard blanket ports, depending on type of breeding blanket (solid or liquid). It is worth to mention a paper not dealing with a specific power reactor design but with the strategy for fusion development [10]. The author discusses the interest of what he calls a "fusion neutron reactor",
namely a reactor which exploits the fact that fusion is neutron rich as compared to fission. Such a reactor could bc of interest as a tool to produce fuel for fission reactor plants and to burn long term radioactive elements (actinides) of fission wastes, as well as to test thc behaviour under neutron irradiation of materials which are of interest for fusion power reactors. The advantage would be that of not requiring to reach ignition: however, to be effective, such a type of reactor should have to operate in steady state conditions. The qucstions related to the complexity of such a system as compared to the present Next Step (ITER, NET), are not answered; neither thosc rclated to the economical interest of producing fucl for fission plants by this way as compared to other techniques (uranium cnrichmcnt, extraction from spent fuel, use of accelerators, etc.). 2.2. D - ~ H e power plants
In the area of advanced fuel cycle studies the attention is focused on D - 3 H e . A concept of a tokamak commercial reactor is proposed by the University of Tokyo [ll]. The basic physics assumption is that it operates in the second stability regime of plasma. This would reach /3-values of about 20% so one could operate at low magnetic field (12 T) and relatively low plasma current (9 MA). In these conditions also the other relevant parameters, namcly the synchrotron radiation and total wall loading would be acceptable. A second contribution concerns the interest of using an organic coolant and reduced-activation-steel in the in-vessel components of ARIES-III [12]. The organic coolant can operate at high temperature (425°C) and low pressure (2 MPa), then an efficient power conversion cycle is obtained. Radiolytic decomposition rate of organic liquids, which was the main potential drawback in case of fission reactor plants, should be acceptable here due to the low neutron production ratc. 2.3. Trends
The two conceptual designs SSTR and ARIES-1 correspond to different philosophies in assessing the features of commercial fusion power plants. Indeed, as already mentioned, the Japanese effort has been that of evaluating what are the prospectives of fusion power reactors in the hypothesis of a moderate extrapolation of the present knowledge both on plasma physics and on nuclear technology. On the other side, ARIES-1 considers very advanced solutions for the magnets and structural materials of in-vessel components. However both rely on a large boostrap current (70-80%) and on
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high aspect ratio (4.0-4.5) which leeds to relatively low plasma currents (10-12 MA) and high magnetic fields (16-20 T). The importance of these assumptions is shown when comparing the main parameters of these two machines with those of EEF [13], the European power reactor design whose main parameters, defined in 1989, were based on the existing physics data base without major extrapolations (table 1). For structural materials both SSTR and EEF consider martensitic (ferritic) steels. Their behaviour under neutron irradiation up to the fluences expected in power reactors remains an open problem. In the EEF study it was considered, as an alternative to martensitic steels, vanadium-base alloy. Such alternatives should be kept also in the next design studies. One has to note that the use of composites, as proposed in ARIES-l, introduces a large flexibility in the design which has been proved usefull in other technological branches. The main problem seems to be the behaviour of SiC and fibres under neutron irradiation at the level of fluence expected in power reactors. The solution proposed in ARIES-1 of isotopic tailoring the divertor materials seems less realistic. However in the next years the knowledge on the the physics of the plasma edge will be improved, so one can expect new ways to spread-out
the particle and heat fluxes on the collector plates of the divertor. This would facilitate the choice of materials more favourable from radioactive point of view than tungesten. The possibility of getting the high magnetic fields proposed in ARIES-1 is also still open. However the recent development of new conductors in some laboratories makes this assumption reasonable. The TARTR study belongs to a different category as compared to the previous ones. Indeed it calls for radical innovations which still require confirmation in particular on plasma physics and on the behaviour of materials under neutron irradiation. However, in our opinion, this concept merits further investigation due to its potential capability for easy extrapolation of the present technology and low capital cost. In the area of advanced power reactor systems it is interesting to note that for some time conceptual studies have focused on D - 3 H e fuel cycle and tokamak configurations. Recent designs as APOLLO (1988), ARIES-III in USA and now the Japanese one are usefull to identify the main critical issues and the areas where to address the research effort. The trend is now to consider plasma operation in the second stability regime. This adds another uncertainty to the several ones, both on physics and technology, which character-
Table 1 Major parameters of commercial D-T tokamak reactor Type Configuration Plasma major radius (m) Plasma half width (m) Plasma elongation Aspect ratio Plasma current (MA) Toroidal filed on axis (T) Peak filed on TF coil (T) Mean electron density (102°/m3) Plasma temperature (keV) Current drive method Boostrap current fraction (%) Current drive power (MW) Peak neutron wall loading (MW/m 2) Coolant Breeder Structure Thermal power (MW) Net electrical output (MW)
EEF (1989)
SSTR
ARIES-1
DN 5.31 1.40 2.0 3.78 16.6 6.2 14.9 1.45 20 NBI 50 91 4.1 water liquid (Pb-17Li) ferritic steel 3780 1200
SN 7.0 1.7 1.85 4.1 12.0 9 16.5 1.4 17 NBI 75 60 5.0 water ceramic (Li20) ferritic steel 3710 1080
DN 6.75 1.5 1.6 4.5 10.2 11.3 21.0 1.5 20 ICRF fast wave 68 100 3.5 He ceramic (Li2ZrO3) SiC composite 2544 1000
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ize this type of power reactor concept. Further specific studies on single aspects of the design are not recommended. On the contrary it seems mandatory to make an evaluation, even if inevitabliy approximate, of the expected costs of the fuel supply from lunar soil, based on the operations and processes identified by NASA in 1988. 3. D e s i g n a n a l y s i s in s u p p o r t to t h e N e x t S t e p
The papers presented are dealing with the neutron effects on various components around the plasma, namely: - Neutral Beam Injectors, - Pumping Ducts, - Windows for Electron Cyclotron Resonance Heating (ECRH), - Magnets. The 2D-analysis of the I T E R neutral beam (NB) system shows that the beamline components will be exposed to an important neutron flux, 10%10 ~ n / c m 2 [14]. On the other side, the induced radioactivity in the NB-room is sufficiently low to consider the corresponding waste of "very low" level. Similar results are obtained for the I T E R plasma pumping ducts, analysed by the 3D Monte Carlo code TRIPOLI-2 [15]. With an appropriate bending of the ducts, an attenuation of the neutron flux up 10 ~' on the concrete wall in front of them can be obtained. In another contribution [16], calculations show that degradation of the dielectric and thermal conductivity properties of ceramic (AI203) foreseen for gyrotron windows of E C R H are unacceptable even at moderate neutron fluence (1019 n/cm2). This is a further confirmation of the difficulty one has to expect in the design of plasma facing components which involve insulating material. The paper on neutron effects on magnets [17] shows that in the case of austenitic steels as structural materials, the possibility of avoiding remote handling after shut-down would require an optimization of the inboard shielding materials and an increase of about 10 cm of the shielding thickness which is in contrast with other design and economic aspects. A last paper is devoted to a sensitivity analysis of the effect of taking energy self-shielding into account in radiation shielding calculations for NET [18]. As response parameter, the total nuclear heating in the inboard coil is taken. Variations of up to 13% are found when self shielding is included and up to 10% due to uncertainties in the angular distribution of the secondary neutrons.
4. N e u t r o n i c tests a n d a n a l y s e s
4.1. Integral experiments in blanket configurations All contributions in this area are related to the J A E R I / U S D O E collaborative programme on fusion neutronics. This programme is based on the utilizatkm of the Japanese fusion neutron source (FNS) for integral experiments of increasing complexity on configurations simulating a blanket with LifO as breeder material. Up-to-now three phases of the collaborative programme have been realized. At the conference the results of the Phase-Ill are presented together with a review of the overall set of experimental results and analysis. Three papers [19,2(/,23] are reporting an overview and detailed experimental results and comparisons with calculation of Phase III programme. In these measurements a line source was simulated by cyclic movement of the annular test assembly relative to a stationary point source that was located axially at the center of the inner cavity. The test assembly was including a 1.5 cm thick stainless steel first wall, a 20 cm thick L i f o zone and a 20 cm thick LizCO 3 zone. The simulation of the line source has been achieved through two modes of operation, namely stepwise and continuous mode of operation. Measurements included tritium production rate from ~'Li and 7Li, in-system spectrum and various foils activation. Different codes have been considered for the analysis, namely DOT 5.1 and D O T 3.5, Monte Carlo codes MCNP and MO RSE in conjunction with E N D F / B - V and J E N D L data files. Tritium production rate from 7Li is generally predicted within 10-15% while for "Li it is within 10-20Ci. Tritium production rate from natural Li is well predicted (5-10~/,). The agreement with foil activation measurements is within 10-20%. Two other papers [22,23] make the point on the capability now achieved in USA and Japan to predict tritium production and other reaction rates in fusion systems on the basis of the experience acquired during the three Phases of the collaborative neutronics programme. The total number of experiments was near to 20, with different incident neutron source conditions, point and line source, and experimental arrangements of the test assembly, ranging from a simple, one-breeder zone to a more prototypical blanket that included stainless steel first wall, beryllium and coolant channels. In all cases the breeder material was Li,O. In the USA analysis the calculated-to-experimental values for the various parametcrs are treated statistically. The prediction uncertainty of tritium production from ¢'Li is ranging between 9% to 5%, based on
G. Casini et aL / Trends in nuclear systems design and analysis
deterministic and Montecarlo results respectively. For 7Li the uncertainties are higher (10.4% and 7.5% respectively). It is intersting to note that, if the results of Phase-1 which involved simplest geometrical and material characteristics are excluded, the prediction uncertainty is reduced. In the JAERI analysis an evaluation of the experimental accuracy obtained in the various Phases of the test programme is made. It appaears a continuous progress in the definition of the experimental configurations and in the measurement techniques. This fact explains why, when the Phase-I experimental results were removed in the U S A analysis, the calculated to experimental ratios were improved. The calculation accuracy for reaction rates including tritium production in the overall set of experiments is found to be less than 5-10%. In the first Phase a systematic discrepancy on the tritium production cross section of 7Li was found which could be explained by using an upgraded data library for the analysis. Also in case of configurations involving beryllium a systematic effect was observed. It is concluded that there are still uncertainties in the back scattering and secondary energy spectrum in the low energy range of Be cross sections. Another paper related to the U S D O E / J A E R I collaborative programme [24] concerns direct nuclear heating measurements and analysis. Microcalorimetric technique has been employed for on-line measurements. Small probes of several materials which are candidates for fusion applications have been irradiated in close vicinity of the rotating target. Temperaturechange and heat deposition rates as low as 30 /xK/s and 35 ~tW/g respectively, have been measured. These experiments are analysed by MCNP code and by comparing various heating coefficients and kerma factor libraries. Large deviations from the experimental data and discrepancies when using different libraries are found. The same applies to the comparison between measured and calculated spatial profiles of heat deposition. A last contribution [25] in the frame of J A E R I / U S D O E collaboration is dealing with the analysis of measurements of activation of several elements (20) of interest for fusion. The paper deals with the comparison with the code T H I D A which is including an updated activation cross section library. As compared to previous analyses made by the same authors, the accuracy of the code prediction is increased in the high energy range ( > 1 MeV), whereas below this threshold large discrapancies still exist. One serious problem raised by the authors is that of inadequate prediction in the products of (n, y) reactions.
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4.2. Integral experiments on single material
The question of Be-nuclear data is analysed in two papers [26,27] which present measurements of neutron transport and multiplication in integral experiments on Be-spherical shells. The comparison between E N D F / B VI (USA) and EEF-I (EC) based on neutron leakage spectra measurements shows remarkable differences in spectra and neutron multiplication. There are indications that the 14 MeV (n, 2n) cross section of Be in E N D F / B - V I is too low and the secondary energy distribution of the EFF-1 neutron emission cross-section at low secondary energies is too high. Three papers from JAERI report on neutron spectrum and flux spactial distribution in integral experiments using test assemblies of tungsten, lead and iron. In the first one [28] Tungsten was found to give the largest neutron attenuation among various materials as Fe, Be, C and concrete. The comparison with codes ( D O T 3.5/JENDL-3) enabled to identify a possible understimation of the cross section of inelastic or (n, 2n) reactions in the data library. In case of lead [29], the angular neutron flux spectra from slabs were measured by the time-of-flight method. The agreement between calculations (JENDL-3) and experiments is in general good in the energy range investigated (0.1-15 MeV). Some minor changes are proposed in the Pb (n, 2n) cross section libraries. In the test with iron [29], neutron spectra in a large cylindrical assembly, using proton recoil gas proportional counters are measured. The comparison with codes (MCNP/JENDL-3) gave in general good agreement, but the discrepancies at higher energies ( > 100 keV) become larger, up to 30%, at positions far from the source. The novelty of the experiment lies on the fact that it investigates low energy range ( < 1 MeV) typical of radiation shielding and magnet conditions, for which no data were previously available. In another paper y-heating measurements by a proportional counter have been made in a low-Z assembly ( M g / C ) irradiated with 14.8 MeV neutrons [30]. The neutron-induced signal was separated from the gamma-induced signal by exploiting the signal size-time rate differences inherent to radiations of different Linear Energy Transfer (LET), which are observable in a proportional counter. The experiment was modelled using the one-dimensional radiation transport code A N I S N / P C and ENDFB-V. To complete the panorama one has to mention the announcement of a plan of neutron shielding experiments using the Frascati Neutron generator (FNG)
[32].
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4.3. T r e n d s
5. Conclusions
The amount of results on neutronic tests presented at the Conference is impressive and the trend towards the establishment of a complete set of data for neutronic design studies looks appropriate. The largest set of experimental data has been obtained at the Japanese Fusion Neutron Source (FNS), in the frame of the J A E R I / U S D O E Collaboration Programme on Fusion Blankets. Among them, one has to first highlight the importance of the Nuclear Heating and Induced Radioactivity measurements because they represent, the unique systematic experimental approach to the implementation of data which are vital in the design and safety of fusion reactors. In particular one has to stress the advanced microcalorimetric technique used for online nuclear heating rate measurements. As one would expect, due to the novelty of the experimental data, the first comparisons with the available codes show a large spread in the calculated versus measured heat deposition values. A further effort, even involving other Countries out of Japan and USA, is highly recommended. A similar approach is being pursued by JAERI and UCLA in the area of activation calculations. The analysis presented is involving the most important elements responsible for neutron activation in fusion materials. Here again even if remarkable improvement with the existing data base has been already obtained, a continuation of the effort remains mandatory. Moving to breeding blanket tests, the capability of predicting tritium production from 6Li and 7Li looks now rather satisfactory, at least when Monte Carlo Codes are used in the analysis. The accuracy is around 5%, a figure which fits well with the usual safety margins taken in neutronic design evaluations. In case of beryllium, the way of measuring in a clean singlematerial geometry neutron leakage spectra, leakage multiplication factors and absorption rates, as proposed by KfK, looks as the best way to check the cross-reaction data of this important element. The discrepancies still existing in the most recent nuclear data libraries, as E N D F B / V I and E E F / I , show that a definitive clarification is not yet achieved. On the other hand, the question of how to exploit at the best the complementarity of the homogeneous experiments with the heterogenous ones (as in the J A E R I / U S D O E case) merits further investigation. For other materials the way of setting-up clean-geometry, single-medium integral experiments of the type presented at the Conference should be pursued.
The progress and prospectives of activity, in the areas here reviewed can be summarized as follows: - D-T
Power Reactor Studies
Conceptual design studies of commercial power stations of Tokamak type are underway in Japan and USA. Similar work is expected in the next period in Europe. Even if the hypotheses taken in physics and technology are often quite different, these studies represent a valid forum to identify the prospects of D - T fusion. The present tendency towards high aspect ratios, low plasma current and high magnetic field should be exploited. The investigation of the alternative solution of a reactor with tight aspect ratio should also be pursued. - D-3He Power Reactor Studies
The recent effort in the evaluation of the main aspects of power plants of tokamak type based on this fuel cycle seems adequate to identify the critical design issues. An assessment of the economic implications of 3He supply from the moon is needed. - N e u t r o n i c Tests a n d A n a l y s i s
The experimental information from neutronic integral tests on materials used in plasma facing components, breeding blankets, shielding and magnet materials has remarkably grown in the last period. The recommandation would be that of pursuing along the way of experiments which are as clean as possible, concerning homogeneity and geometry, and that of using Monte Carlo Codes for their analysis. In general, neutronics seems one of the technology areas where the effort is being pursued in the most satisfactory way.
References
[1] M. Kikuchi, Y. Seki, A. Oikawa, T. Ando, Y. Ohara, Conceptual design of the Steady State Tokamak Reactor (SSTR), ISFNT-2, Karlsruhe 1991, Fusion Engrg. Des. 18 (1991) 195-202, in these Proceedings, Part C. [2] S. Mori, S. Yamazaki, J. Adachi, T. Kobayashi, S. Nishio, M. Kikuchi, M. Seki and T. Seki, Blanket and divertor design for the Steady State Tokamak Reactor (SSTR), ISFNT-2, Karlsruhe 1991, Fusion Engrg. Des. 18 (1991) 249-258, in these Proceedings, Part C. [3] Y. Suzuki, M. Yamada, M. Tomita, M. Kikuchi, S. Nishio, Y. Seki, A conceptual design design of a low resistance vacuum vessel for steady state Tokamak Reactor (SSTR),
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