IHLÖ\Aooo^
ASSOCIATIE EURATOM-FOM
FOM-ÏNSTITUUT VOOR PLASMAFYSICA
RUNHUIZEN - NIEUWEGEIN - NEDERLAND
ON FUSION AND FISSION BREEDER REACTORS; THE IIASA REPORT RR-77-8 REVIEWED AND UPDATED
by
B. Brandt, H.Th. Klippel and W. Schuurman
Rijnhuizen Report 81-129 ASSOCIATIE EURATOM-FOM February 1981
FOM-INSTITUUT VOOR PLASMAFYSICA
RIJNHUIZEN - NIEUWEGEIN - NEDERLAND
ON FUSION AND FISSION BREEDER REACTORS; THE IIASA REPORT RR-77-8 REVIEWED AND UPDATED
by
B. Brandt, H.Th. Klippel and W. Schuurman
Rijnhuizen Report 81 -129
The author H.Th. Klippel is employed by the Slichting Energieonderzoek Centrum Nederland (FX'N)
This work was performed as part of the research programme of the association apreement of Euratom and the "Stichting voor Fundamenteel Onderzoek der Materie" (FOM (with 'inancial support from the "Nederlandse Organisatie voor Zuiver-Weten schappelijk Onderzoek" (ZWO) and Euratom. CONTENTS page
INTRODUCTION 3
FUEL RESERVES 3 PRESENT STATUS, RESEARCH AND DEVELOPMENT, AND 6 REQUIREMENTS FOR COMMERCIALIZATION 3.1 Present status and development of fast breeder reactors 6 3.2 Present status and development of fusion reactors 10 3.3 Cost aspects 14 3.4 Summary 14 REFERENCE REACTOR DESIGNS 15 4.1 Designs of fast breeder reactors 15 4.2 Design studies for fusion reactors 18 4.3 Reference reactors 18
RADIOACTIVE INVENTORY OF REACTOR SYSTEMS 22 5.1 Problem exposition 22 5.2 The fast sodium-cooled breeder reactor 2 3 5.3 The fusion reactor 30 5.4 Comparison between fast breeder and fusion reactor 35 5.5 Conclusions 39
EMISSION OF RADIOACTIVITY AT NORMAL OPEi^ATION 40 6.1 The fission reactor 40 6.2 The fusion reactor 40 6.3 Conclusions 42 DANGER OF ACCIDENTS 42 7.1 The fast breeder reactor 42 7.2 The fusion reactor 4 4 7.3 Consequences of an improbable hypothetical accident 45 7.4 Summary 46 PROTECTION AGAINST ABUSE 47 8.1 Nuclear explosives 47 8.2 Radiological weapons 48 8.3 Safeguards 50 MATERIALS AND RADIATION DAMAGE 50 9.1 The fast breeder reactor 50 9.2 The fusion reactor 52 9.3 Conclusions 57 9.4 Consumption of materials 58
SUMMARY 62 REFERENCES 65 -1-
1. INTRODUCTION
In July 1977 the International Institute for Applied Systems Analysis at Laxenburg, Austria, published a detailed report (here after called I), in which fission and fusion reactors were compared. 2) Data of fission reactors were mainly taken from the SNR-300 ; as a prototype of the fusion reactor the UWMAK-I Tokamak Design Study of the University of Wisconsin 3) was taken. The D-D fusion reactor, the hybrid reactor,and the laser fusion reactor were briefly described in appendices.
The fact that the IIASA-report was composed of the work of several independent groups involved in fission and fusion research led certain points of view to not being tuned in tc each other, especially the treatment of the radioactive inventory of the reactor. In the following considerations we shall summarize the results of I and clarify them on certain points. We shall also consider more recent developments in the field of thermonuclear reactors. In the conclusions, a summarizing table will compare the two ways of energy production qualitatively and sometimes quantitatively. This report was written in Dutch for internal information in 1979. The authors have provided the present version upon repeated suggestions trom abroad.
2. FUEL RESERVES
The world fuel reserves for fission and fusion reactors, their energy content and consumption time are assembled in table 2.1.
a. The number for lithium was conservatively estimated in 1970 and is valid for a production price up to 0.06 $ per gram metallic lithium. For uranium the number was taken from a OECD/IAEA-report, December 1975. It concerns only uranium inventories producible at a maximum
price of 0.07 $ o+; 1975 per gram U,Og. b. The volume of seawater is 1.4 x 1018 m3 . The concentration of D in seawater is 33 g/m3 . The average concentration of Li in f.eawater is 170 mg/m3, that of uranium only 3.4 mg/m3. The price of this uranium and the technology of its winning are uncertain.
-3- c. For D the basic reactions are D + T * "He + n + 17.6 MeV and 6 Li + n •* "He + T + 4.8 MeV. For Li v/e assumed a tritium breeding ratio of 1.3 without enrich ment of 'Li. For uranium we used the statement in I that 5*1012 kg U in the oceans, if used in fast breeders, have an energy content of 2*10-:" J. d. The load factor is assumed to be 100%. e. The total energy consumption in the world in 1975 was 7*io13 kWh (10% electrical). f. The assumption was made that all energy was produced in reactors. g. By fuel costs are meant the costs of winning of the natural ele ment. h. In this row the total cost of the fuel inventory per kWh(e) produced are given. In the case of LiA&O, as breeding material the cost of the neutron-multiplying beryllium have been taken into account. Further included in the costs are those of the first loading of the blanket and of the reprocessing or replacement of breeding material and neutron amplifier once every two years (with depreciation taken into account). For uranium the costs are mainly those of the fabrication of fuel elements, interest costs for the plutonium inventory and the costs of transport, reprocessing and waste disposal. As a comparison: the corresponding costs of the fossile fuels hard coal and oil/gas are 23.2 mill/kWh(e) and 19.2 mill/kWh(e) respectively, the fuel costs of the LWR are 5.5 mill/kWMe) .
CONCLUSION
The fast breeder and the fusion reactor produce an energy in the order of 1 MW{th)day per gram fuel. The ore costs are relatively low so that also low grade ores can be used. The reserves are not accu rately known but will certainly be sufficient on the long run. Thus, the duration of a world economy based on nuclear energy is not limited by the fuel reserves, but possibly by the availability of the other materials (see chapter 9).
-4- Table 2.1.
Fusion reactor Fission reactor
deuterium natural lithium natural uranium a. estimated world reserves - 6xl09 kg 3.5xio9 kg {not in seawater) b. estimated world reserves 4.6*1016 kg 2.4*10llt kg 4.8*10IZ kg in seawater c. energy content per gram 100 MV?h(e) 4 MWh(e; 7 MWh(e) of natural element d. fuel quantity required per yeer for a 1 GW(e)- 90 kg 2300 kg 1300 kg reactor e. fuel quantity required fuel: 7*105 kg fuel: 1.8xl07 kg fuel: l.OxlO7 kg per year for total raw material: raw material: raw material: world energy consumption 8,5*109 kg seawater 4.5*10a kg pegmatite or 3.5*l0y kg Colorado sand 16x10 r/ kg seawater stone or 2 . 3X101 •' kg seawater f. number of years that without seawater: without seawater: fuel is sufficient for 6;-10io years 330 years, with seawater: 350 years, with seawater: world economy 1.3*1Q7 years 4.4*10'' years g. fuel costs (ore) 5xio-3 mill/kWh(e) 4xio-'* mill/kWh(e) 1.5xio"? mill/kWh(e) h. fuel cycle costs 6*10~3 iruil/kwh(e) Li: 0,6 mill/kwh(e) 4 mill/kWh(o)
LiA^o3: 6.4 mill/kWh(e) 3. PRESEMT STATUS, RESEARCH AND DEVELOPMENT, AND REQUIREMENTS FOR COMMERCIALIZATION 3.I Present status and development of fast breeder reactors
The principle of breeding with fast neutrons was already known when nuclear research began. Fermi and Zinn designed a fast breeder reactor as early as 1944. The first fast breeder experiment was Clementine, that became critical for the first time in 1946. Electric ity production was first demonstrated with EBR-I in 1952. According to views of that time uranium or plutonium metal was chosen as a fissile material. The small experiments till 1960, based on this principle, can be designated as the first generation or fast reactor experiments, the most important ones being Clementine (1946), EBR-I (1952), BR-5 (1958), DFR,and EFFBR. The majority of these are no longer operative. Their power did not exceed 60 MW(th). Uranium or plutonium as a fissile macerial made the core compact and the power density high. The coolant was either liquid sodium or mercuy. In relation to the long-term strategy, the achievement of a short doubling time was emphasized rather than a small inventory and low costs of the fuel cycle. In view of the limited power (- 100 HW(e) ) of the power stations of that time one had expected to reach this power level in a single step. After 1960, also based on progress in technology of thermal reactors, stress was laid on economic aspects of the fuel cycle, in particular on the realization of a high burn-up. In comparison with the thermal reactor, the fuel of the fast bleeder reactor has a high enrich ment factor, making the economic burn-up of the order of 10s MWday/ton, two or three times as high as in the LWR. The most suitable fuel for this is the ceramic mixture UO-/PuO„. The choice of this fuel ushered in the second generation of fast reactor experiments, with completely different physical aspects (a.o. softer spectrum) and technical aspects (a.o. lower power density). The most important members of this second generation of fast breeders are SEFOR (1969), BOR-60 (1969), Rapsodie (1970), KNK (1977), Joyo (1977) and FFTF (1978). Their power is of the order of 60 to 300 MW(th) without electricity production. Larger proto types with energy production of about 300 MW(e) are already operative (BN 350, Phenix, PFR) or are being built (see Fig. 3.1). The modified reactor-physical and technological characteris tics of these reactors with oxidic fuel led to much theoretical and experimental research around the determination of the Doppler coeffi cient and the sodium-void effect (see b). Parallel to this, the re? tor
-6- MW 64 66 68 70 72 74 76 78 80 82 84 I t I I I I I III III! 1 2 3 FRÖ SNR-300 300 Belgium I 11
the 1 SNR-2 1200 Netherlands /
1 2 3 France PHENIX 250 1 SUPER - 1 2 3 1200 PHENIX 1 1. J-.:, -..-;,.:• 1 SAONE-1 1500 / 1 SAONE- 2 1500 /
United 1 2 3 PFR 250 Kingdom I Ik.' ',J*:.-:.-.-.: .• 1 2 3 CFR-1 1300 y yy
1 2 3 USSR BN -350 350 I I,-. i-U!^. •::.-:. " 1 2 3 BN 600 600 I r:;;i"":::'..
1 2 3 USA CRBR 350 • yi i/r. ••••
1 2 3 Japan MONJU 300 / y::j
1 start of construction 2 criticality 3 full power operation
Fig. 3.1. Time schedules for different international fast breeder projects. safety and accident prevention were analyzed experimentally, both "in-pile" and "out-of-pile". Figure 3.2 gives a survey of the projects for the various physical and technological aspects.
-7- 2O00 19<-4 1950 1960 1970 I960 1990 (2010)
VERA BFS p ZPR - VI CLEMEN :[_ZPR- tX TINE BRM fast *—. r~rrr-\ /-t physics Of larg« reactor cores and special core geometries ) physics BR 2 FCA ZPPR 1955 SNEAK ZPR- lit MASURCA
TREAT 19t>9 - 1972 SAREF- ,-CABRI fast SEFOR reaccor ft] HSLFS - ETR \X] safety il dJJdJMOL - BR 2/-J
smalt sodium USA loops Japan £] 3J1 —[ sodium France Germany technology pump intern Netherlands heat exch. Belgium steam generators component corros.test loops . test loops RAPSODIE 1967
DFR 1963 — BIM - 350 SNR - 300 1973 EBR - II SUPER - 1965 ,PHENIX-74 [FENIX EBRH [jPFR -76 I CRBR 1951 /rrrryri power reactor cfi cQ -M develop • —^p i i ment BR- 5 BOR - 60 rT SN- BN -1600 |_ first large scale 1955 .. !SNR -2 commercial LMFBR 1969| | 600 SAONE RAPSODIE L KNK - II u—' MONÜU 1967 JOYO CFRH (XLJ-LHil l materials 1est of PuO,/ UOj R (- D for advanced fuei develop t fuel, steel and absorbers advanced absorbers and ment ppy'£3p steel (high neutron fiuences)
USSR France 1 fabrication fC^Ldecision on UK Germany plants for e o / iïrV PuC/UC Japan Belgium few tons PuO,/ U02 carbide| USA Netherl. PuOj/UO fuel fabric fuel fuel fuel per year plants to application fabrication 1 to serve I serve ,R + D fo plants metallic ^el pu0"/"uo^prototypes j—commercial RT-\ +T UD +I - * -v I 1 ;-hPu0 / UO, LI 2 .fabrication fuel fabrication r—' | £2 'fuel plants pilot plant France small pilot plants first type for EBR - II UK reproces - rpj up) to about 10 t // year P commercial sing r 300 MW(e) plants •preprocessing -' plant
F.i']. '>.2, Development strategie:-; for \.V=A_ •r.wAar ro^ictoru -8- The transition from compact cores with metallic fuel to larger cores with ceramic fuel had important consequences for the neutron spec trum, the Doppler coefficient and the sodium-void effect. In the cores with ceramic fuel the neutron spectrum is softer, making the intermediate part of this energy spectrum (the resonance region] important. In this region self-screening of the nuclides occurs Because the thermal expansion of the oxidic fuel is low, the Doppler coefficient becomes the most important parameter for inherent stability and therefore had to be determined accurately (^s was done in SEFOR). The loss of sodium in the core has a number of contrasting effects on the reactivity due to changes of the spectrum, of the total cross- section for neutron capture and of the neutron leakage. For the cores with metallic urarium all the effects combined gave a negative coeffi cient. In the ceramic reactors the Na-void coefficient may be positive, essentially affecting the reactor safety. In many places of the world theoretical and experimental investigations were devoted to the fore- mentioned aspects. It led to the development of extremely refined calcu- lational techniques and extensive libraries for reaction cross-sections, as well as the performance of varied experiments in fast critical fa cilities (SEFOR, ZPR, ZEBRA, SNEAK, MASURCA, STEK).
c- Të£b02i22i£§i_E?5Ë3£Sh The most important technological research concerns the coolant, with liquid sodium as a favourite choice. Helium is a gooo. alternative, but its development is progressing slowly. Before 1970 H^O-steam-cooled and D-O-steam-cooled reactor concepts were studied, but they were aban doned for technical reasons. The thermal breeder reactor with molten salt as a coolant is still in its first stage of development. Among other advantages such as a low cross-section for neutron capture, sodium has a high specific heat, a high thermal conductivity, and a high boiling point (900 C). This lenders possible a reactor without high-pressure pipes, with a high operation ter.perature, and thus a high efficiency. Disadvantages of sodium are its high melting point (100 C), the chemical reactions with water and air, its opacity, ana its becoming radioactive in the radiation field of neutrons. These properties require a specific reactor design, in particular a subdivision of the integral cooling circuit into a primary, radio active sodium circuit, a secondary, not radioactive intermediate rodium circuit, and a tertiary watersteam circuit in which the turbine-genera-
-9- tor system is situated. An accident in th« steui generator viil there fore not lead to the release ot radioactive material. In the sodium technology, during the last decennium much work was done on the development AnA testing of sodium pumps, heat exchangers and stean generators. In various 1:1 test facilities these sodium com ponents were tested with satisfactory results. Other technological research aspe ~ts concern the fabrication of fissile fuel elements And absorber elements, the behaviour of materials at high neutron fluencies (see chapter 9), and the development of a loading and reloading apparatus.
From the experiences with the first and second generation of reactor experiments and fros: the testing of sodium components it turned out that a satisfactory use of the scciua technology over a long period is possible, provided that auch care is bestowed on the fabrication ar.d on the selection of materials. The construction of 300 M%i(e} prototypes started in I"H5, soce of them are already in operation (see Fig. 3.1 J. This figure shows that the development in the USA is lagging behind that in Europe. In the Phenix reactor a burn-up of 70 GOO Wtóay/t was reached without any indi cation of failure of the fuel pins at a lead factor of at least 801. For a further description of the characteristics of some reactcrs, see chapter 4.
The next step toward.» commercial feasibility will Le the construc tion of large fast breeder reactors of an output in the range of 1200 to 1300 M»(e) or more. Super-phenix and BN6Q0, and the Jesigr, of JN3-2 and CFR-1 (see also Fig. 3.1) are the first steps in this direction. Only after I >8S will these dcsmr.stration reactors he fully operative. Their technical success will mainly depend on the quality jf the sodiur. components 3.2 Pre5'-'n' status anti dcv"..'lf)p-in'n> of i'u.sior _r" arrays a. .zc: **r.'•iïi£_!£25i2ili!ï Tht: fursi^r. reactor Ï? -or.siH.TH R'-i«:n* i f i~a I iy f-'-isib!»' ih<*r: ri-^ 'h'TE.ii outp.i* i'owor cxrocïis * he inra! !Kwr. Tho cor,f'ir.o:.i y 1 isni *h»*-, 1 his to fuifiii ,i "'T'fiin cor.'iit i' ", t"he z- --'••! I l«-r: ,:ri'<" ion. !'-T a D-T r«?a-rt'ir '•his cri^ri'ir. r»>afis r: 10* n~ ' s .r.d T • r> '<«•'.', in which, r., »:'.•'! T ,tr^ t'nc fiism-i densi'y, f:-,«> cin! i :-^7",:-\t tim.- .\~ I r '.'," :>lasira •ciriM'r.i'uri! rnsiic-* ivoly. For *••?'< ïrsaks, • ho f,111 /,i* i or. r>f ;\;irh par.Ei'1'1'.; is »'Xi)Oi-'••r; i:> r;' !'»84, f'»r ]",••?' ; ii '" >r. f i n>'T,'v:,f fXD'T :-•;••:•*:; -I"/- with particle beams or lasers a similar timing is expected- The progress of tokamak research is shown in Fig. 3.3 and is summarized in table 3.1. The following is valid for the most advanced type of machine, the tokamak. other confinement systems, such as mirror machines and fast pinches are lagging somewhat behind, but a break-through is not excluded. Laser fusion research is waiting for the development of 100 kJ-lasers with a pulse duration less than 1 ns and an efficiency of at least 5%. peak ion-temperature T\ (10° K) ^ 10 100 1000 21 t 10 10 20 coming ALCATOR (1978)» experiments / \ * oTFTK / PDX ASDEX ' I V / / PLT (1977) ALCATOR (1977) FT / 10 19 4^ / •PLT (1978) \ / recent \ / experiments TFR / , ^PULSATORN / :s 10 18 T - 4 z i ATC SCYLLAC / • TOKAMAK CLM ST older A experiments A high 6 BELTPINCH H / + STELLARATOR Clio* 'ORMAK/ T;3 / 10 17 _L -L I I 1 M ' ' ' JLL j 1 ' ill' 1 10 100 peak ion-temperature T4 (KeV) Fig. 3.3. Lawson diagram with a survey of present-day and future values of nT and T. •11- Table 3.1. Plasma Parameters in Toroidal Devices Sustainment Year Tl (s) (Kelvin) (s/m*) Time (s) 5 1955 io- 105 1Q9 10^ 1960 10"^ 106 io10 3-10-3 1965 2«10"3 106 1011 2«10-2 1970 10-2 5-106 5-1Ü11 lO"1 7 1976 5^10"2 2'10 1013 10° (T-10, PLT) (TFR, ORMAK) (ALCATOR) (T-10, PLT) Needed for a 10° 108 10lu > 10 Reactor Particle beam accelerators, especially of light ions, presently seem to be more advanced. The budget for inertial confinement in the USA in 1978 amounted to 120 M$, for magnetic confinement this figure was 270 M$. Of the latter, 60% went to tokamaks, 20% to mirror machines, and 20% to the remaining subjects. In order to progress from scientific to engineering feasibility it must be demonstrated that a fusion reactor can generate net power on a reliable basis over long periods of time. This step will probably be as large as the demonstration of scientific feasibility itself. There are general problems, such as radiation damage, the realization of a breeding ratio larger than one, and the handling of the highly volatile tritium, and specific problems. Examples of the latter are the develop ment of superconducting coils for tokamaks and mirror machines, heating and fuelling in tokamaks, load-leveling schemes to deliver power during the periods between burn pulses, and the finding of fatigue-resistant structural materials for laser and pinch reactors. This requires the building of expensive, time-consuming experiments and test facilities. For the development of a demonstration reactor (DEMO) in which all technological problems have been overcome, a number of possible scenarios were described. The total costs until the realization of the DEMO are now estimated to be 15*109 $, equivalent to about three months of oil import of the European Community. The intermediate stages in the programme for the development of the DEMO in relation to the technological facilities are depicted in -12- Fig. 3.4. It is seen that the first experimental reactor (EPR) with a pow er of about 300MW(th) is not expected before 1990. The DEMO with a sig nificant but still uneconomic output power appears after the year 2000. Fig. 3.4. Development strategies for a commercial Tokamak reactor. -13- Figure 3.4 illustrates the projects for the different physical and tech nological aspects. In the case of serious problems one has to fall back on a less ambitious scenario. Recently a bill was adopted by the Congress stating that means should be found in order to get the EPR before 1990 and the DEMO before 2000. This might require a doubling of the annual spending within five years. The USA yield about 30% of the total world effort. In budget proposals of the US government the years 1980/1989 show a yearly amount of 4 x 108 $ for fusion by magnetic confinement. c. Cgmmercial_feasibility_ This could occur early in the 21st century, after some years of reliable operation of the DEMO reactor. However, much will depend on the development of prices and the environmental risks of the other fuels in electric power plants. 3.3 Cost aspects Recent cost estimates indicate that the cost of construction of a 1 GW(e) fast breeder reactor are about 2500 $/kW(e) (based on 1977 prices). These total cost are about twice those of a light water reactor of comparable power. The contribution of the fissile-fuel cost, however, is lower than for the light water reactor. At the fast breeder, accord ing to Ref. 6, 67% of the kWh-price is determined by capital deprecia tion, 22% by fu^l costs (predominantly reprocessing), and 11% by servi cing and maintenance. At the light water reactor these portions are 55, 35, and 10% respectively. Estimated cost of construction of a tokamak reactor are 2500-4000 $/kW(e), see Ref. 7, with an inflation correction of 11% per year taken into account. The price of 1 kW for a fusion reactor is mainly determined by depreciation (90%); the fuel costs are negligibly small (< 1%). On the basis of these cost evaluations, the kWh-price for a tokamak reactor becomes 80-120 milis/kWh, comparable to that of a fast breeder reactor (about 30 milis/kWh). 3.4 Summary The stages of the development of new energy production systems are those of scientific, engineering, and commercial feasibility. The present development of the fast breeder reactor is based on the mixed- oxide fuel UO2/Pu0- and on sodium as a coolant. A large research effort has been devoted to reactor-physical and technological aspects, and to reactor safety. Sufficient knowledge and experience has been obtained -14- by means of many experiments in small critical assemblies (scientific feasibility), in experiments both inside and outside test reactors and in medium-sized (~ 300 MW(e)) reactor systems. Technological problems are centred around the sodium components, particularly the steam gener ator. In the near future experience will increase when a large number of reactors of about 300 MW(e) (1980) and big demonstration reactors (1985) will become operative. Then, the engineering feasibility of fast breeder reactors will in principle have been proven. Commercial feasi bility is not expected before 1990. The scientific feasibility of the fusion reactor is expected to be demonstrated around 1984 by the newly designed machines (JET, TFTR, JT60, TM-10). Magnetic confinement (tokamak, mirror machine, high-0 pinch) is more advanced than inertial confinement, and also seems to reach commercial feasibility at an earlier date. In both systems serious technological and material problems will have to be overcome. Engineer ing feasibility should be proven around the year 2000. Only well after the year 2000 proven commercial feasibility may lead to a contribution to electricity generation in the developed countries of about 10%. 4. REFERENCE REACTOR DESIGNS 4.1 Designs of fast breeder reactors In the development of fast breeder reactors the line with sodium as a coolant is dominating. A limited research effort is devoted to gas- cooled and thermal reactors. Of the sodium-cooled fast breeder reactors (LMFBR) several prototypes are already in operation, under construction, or in the design phase. Some details are given in table 4,1. LMFBR's ex ist in two types, the so-called "pool-type" and the "loop-type", see Fig. 4.1. At the pool-type reactor, the core, the sodium pumps, and the Na-Na heat exchangers are all placed in the sodium-filled reactor vessel. On this principle the reactors PFR, CFR-1, Phenix, Super-phenix, and BN-600 are based. At the loop-type reactor, the sodium-cooled vessel contains the reactor core only. The sodium pumps and heat exchangers outside are connected by pipes to the reactor vessel. This is the case in the designs of BN-300, SNR-300, CRBR, and Monju. Both types have medium-sized reac tors in operation and each has its merits and disadvantages. The most fa vourable type will step forward only after ample experience with the large reactors of the near future. The elaborate description of the SNR-300, the RN-300, and the BN-600 in the IIASA-report illustrates the involvement of the authors in the projects mentioned, and it gives an insight into the state of af fairs concerning the design, the safety aspects, and the way of operation of both pool-type and loop-type reactors. -15- Table 4.ï. Fast Breeder Prototype and Demonstration Reactors UK FRANCE FRG SUPER PFR CFR-1 PHENIX PHENIX SNR 300 Reactor Power Thermal MW(th) 600 2900 563 2910 736 Electrical MW(e) 270(254) 1320(1250) 250 1200 312(282) Primary Circuit Pool Pool Pool Pool Loop Number of Loops (3) Diameter of Reac tor Vessel m 12.2 22.5 11.8 = 21 6.7 Coolant Na Na Na Na Na Coolant Tempera ture at Core inlet C 400 400 400 395 377 Core outlet C 552 562c 560 535 546 Core Dimensions Height cm 91 100 85 100 9£ Diameter cm 147 =290 139 -366 178 d Fuel U02/Pu02 U02/Pu02 'J02/Pu02 UO /PuO U02/Pu02 Cladding ss SS SS SS SS(1.4970) Pin Diameter mm 5.84 5.84 6.6 8.65 6,0 Number of Fuel Pins per Fuel Element 325 325 271 271 169 Maximum Rod Power W/cm 450 450 430 450 460 Maximum Clad Temperature (hot spot) C 700 700° 700 700 670 Maximum Fluence at Clad n/cm2 3«10?3 4.6'1023 2 to 3*1023 Burn-up MW(th) day/t 70,000 >70,000 SO,000 70,000 55,000 Breeding Ratio 1.2 = 1.2 1.16 1.17 1.0 - 1.2 Status In oper Planned, In oper Planned, Start of ation start of ation start of construc since construc since construc tion 1973 IS,5 tion 1978 1973 tion 1976 Start of operation 1982 Site Dounrf.ay Marcoule Creys- Kalkar (Scotland) (France) MaLville (FRG) Supplier TNPG TNPG CEA SNR Con sortium Operator UKAEA CEGB CEA/EdF EdF/ENEL/ SBK SBK a or 150 MW(e) 4 120,000 mVday fresh water b . , J. including one as reserve preliminary values, may be lower initial loading partly with UO 8 at first UO with SNR Consortium -16- Table 4.1. (continued) USSR USA JAPAN BN 350 BN 600 CRBR MONJU Reactor Power Thermal MW(th) 1000 1480 950 714 Electrical MW(e) 350a 600 360 300 Primary Circuit Loop Pool Loop Loop Number of Loops <6)b (3 or 4) (3) Diameter of Reac tor Vessel m 6.0 12.8 6.2 7.1 Coolant Na Na Na Na Coolant Tempera ture at Core inlet C 300 380 387 390 Core outlet C 500 550 540 540 Core Dimensions Height cm 106 75 91 90 Diameter cm 158 205 188 178 e Fuel uo2 U02/Pu02 U02/Pu02 U02/Pu02 Cladding ss SS SS 316 SS 20% c.w. Pin Diameter mm 6.1 6.9 5.84 6.5 Number of Fuel Pins per Fuel Element 169 127 271 169 Maximum Rod Power W/cm 470 530 475 457 Maximum Clad Temperature (hot spot) °C 660 700 640 700 Maxinum Fluence at Clad n/cm2 3'1023 Burn-up MW(th) day/t 50,000 90,000 80,000 Breeding Ratio 1.0 - 1.4 0.9 - 1.3 1.23 1.2 Status In oper Under con Planned, Planned, ation struction, start of start of since 1973 start of construc construc operation tion 1977 tion 1978 1977 Site Shevshenko Bjelojarsk John (Caspian (Ural) Sevier Sea) (Tenn.) Supplier PNC Operator State Ministry TVA Committee for Elec Common- for Atomic tric Energy wealth Energy Edison -17- intermediate heat exchanger intermediate heat exchanger pool reactor loop reactor Fig. 4.1. Alternative LMFBR design concepts - pool and loop. With present-day experience the development of commercial fast breeder reactors is expected to be finished before the year 2000. 4.2 Design stuoies for fusion reactors In contrast to the advanced stage of fast breeder reactors, where practical experience is available, commercial reactors are being desig ned, and detailed plans exist concerning the fuel cycle, design studies of fusion reactors have exposed a number of engineering problems, and work on their solution has been started. The most detailed designs take the tokamak systems as the basis for a reactor, in table 4.2 a survey is given of the principal characteristics of several designs of high- power reactors. At the IAEA-conference on Fusion Reactor Design 1977 (Ref. 8) a table was shown of "ignition-test" reactors and experimental reactors (see table 4.3). At this conference it was emphasized that for economic reasons the energy densities and wall loads of power reactors may have to be increased. Integral wall loads of 10 to 20 MW(th) year/m2 seem necessary for economic viability. Suitable wall materials are not yet available and their development will demand a large part of the fusion effort. A concerted effort of the USA, the USSR, Japan, and Europe is the pre-design, at present, of the experimental fusion reactor INTOR. 4.3 Reference reactors For technical and historical reasons the designs of the sodium- cooled fast breeder reactor and the lithium-cooled fusion reactor are the most detailed ones. For the comparative study in I the SNR-300 and the UWMAK-I are used as reference reactors, because the authors had easy access to existing reports and data. SNR-300 is a demonstration reactor, at present under construction at Kalkar; UWMAK-I is a relatively early design study that leaves some room for improvements, particularly a scalmg-dowi • A n 9) •18- u o •* in iC •* CN —• f-i T p- it in —4 ia rt CD o rr ^ v «S M-l XI p- P- p- p- p- r» p- p- r> r> r* oo r> p- p- P- p- p- p- p- O 3 CTi <7l en CJ1 cn cn CT> m cy. CT. CTt cjl CTi en en U ff) i0 u"> en en en cn m ro in m r- en en IC CD ID V 0) O O O O O O O O O O O O O O O •-t »-< »-t *—t •-§ »-t w—1 '-* f-* »^-t r-* *—i «—• ^4 o o ^4 u >! X X X X »-X * 1 X X 1 X X X X X X X 1-* •-* X >. M n r~ p- p- W CN O CN 4J 33 * c O id O H CN .H •A A CU •A CU -A 3) 1 K £ =5 Hl •3 >-l il J J = CU w Co o X O in O ia r~ ro. i o o m 1 ca e in o 1 u 0 u O LD m o ^ rt i£> O *r in m *"* o T o 10 CO o 0) •P ^-^ •-4 wt o m o in •* vO c"? tl o. T> c «J 0 0 O^J in ta )-l r-l in CJ iH in co vo t> o va o ri ex OO o UT VO O v£) i-I o 4-1 e •>. T-l • 3 M 3 o *-< «-( OJ -H • I wal l «-I VO rs t -eallo y -Mo-stee l -allo y -IZ r -31 6 colo y allo y 1 £ 04 -31 6 •H XI CO w CN) W X) c o < ••4 UI a A X) M W 1 XI fc, Z W co E-> o. s M S w tH tl z Z z u U w > z ^ O O o o 10 O O o o in O o m O o o o O O o «J ^ o o o O *r O O O O tN O O 0 1 1 1 1 M M h-Satur n Blasco n -19- DESIGN PARAMETERS Table 4.3 "Ignicion-Test-Reactors" and "Experimental-Power-Reactors". TNS1} ITR2> ITR-UP25 ORNL/W GA/ANL GA/ANL MTF3) Operation conditions power MW 100-2000 700 800 230-400 cycle time (s) 30 300 120 1200 burn time (s) 16 30 30-90 30-60 number of cycles (*106) 0.5 0.2 0.1 0.025 Dimensions large radius R(m) 5.0 3.8 4.2 6.3 small radius r(m) 1.25/2 1.1/3 1.2/3.25 1.4/2.1 plasma volume (m3) 250 260 340 367 reactor radius (m) 10 9.5 10.5 9.8 reactor height (m) 12 12 13.5 10.5 Plasma parameters 7t (keV) 5-10 13 13 6-10 n (xlO20 m-3) 0.6-2 1.9 1.8 1.2 zeff < 1.5 1-2 1-2 T (s) 1 2 2 4 Btor on axis (T) 4.3 4.0 3.9 4.7 q 3 > 2.5 > 2.5 frpol 2.7-3.4 Btor <%> 3-10 5-10 5-10 3.8-4.9 plasma (MA) 4 11.3 11.7 4.8 TF-coils number 12-20 12 12 conductor Cu/NbTi/Sn NbTi NbTi NbTi space (height/width) (m) max.field strength (T) 8 8(10) 8(10) 8 PF-coils conductor eq/ohmic Cu/Cu Cu/NbTi Cu/NbTi Cu/Cu position in-/outside coils in-/outside in-/outside in-/outside in-/outside Additional heating beam energy (keV) 150 150 150 120-150 injection power (MW) ~ 50 ~ 50 ~ 50 60 pulse time (s) 2-10 2-5 2-5 ,facuum vessel position first wall first wall first wall first wall material SS Inconel 625 Inconel 625 Inconel 625 First wall type and material RVS Inconel 62 5 Inconel 625 Inconel 625 (with low- Z coating) neutron wall load (MW/m2) 0.7 0.7 life time (years) Blanket structural material none none SS 316 none coolant - - He - breeding material - - none - thickness (m) - - - - Shield material SS, Pb W,B4C/SS,Pb W,B4C/SS,Pb thickness (m) 0.54 0.44/1.0 0.52/1.0 0.6 H coolant borated HO 2° H20 Tritium flow rate (g/h) ~ 1.7 1,7-2,0 inventory (kg) 0.19 0.2 0.25-1.4 consumption (kg/year) 1) Oak Ridge 2)General Atomic + ANL 3) JAERI, Japan 4) OSSR 5) ANL 6) Italy. -20- Table 4.3 (continued) EPR5) EPR3) T-204> J\NL fINTOR-I6) JAERI Operation conditions power MW 600 90 100 cycle time (s) 300 75 > 250 420(180) burn time (s) 15 60 240 300(1001 number of cycles (*106) 0.4 Dimensions large radius R(m) 5.0 4.7 9.0 6.75 small radius r(m) 1.61/1.75 1.34/2.2 2.25 1.5 plasma volume (m3) 337 900 300 reactor radius (m) 8.7 18 20 reactor height (m) 13.5 27 20 Plasma parameters Ti (kev) 5-10 8 20 7 n (xio20 m~3} 0.5 1.4 0.17 1.1 zeff 1.7 3 1.6 T (s) 2 2.5 16 2.9 Bf-Qj. on axis (T) 3.7 4.5 3.5 6 q 3 2 2.5 Bpol 1.5 2 2.2 Btor (%> 3 7 3 1.7 plasma (MA' 5.0 7.3 5 4 TF-coils number 16 16 24 16 conductor Cu/NbTi K'bTi NbTi Nb3Sn space (height/width) (IT.) 8.7/5.6 16.8/10.5 11.2/7.3 max.field strength (T) 8 9(10) 8 12 PF-coils conductor eq/ohmic Cu/Cu NbTi/NbTi Cu/Cu NbTi/NbTi position in-/outside coils outside outside inside outside Additiona] heating beam energy (keV) 80-160 180 100 200 injection power (MW) 40 12 30 pulse time (s) 2-12 4-6 9 320(120) Vacuum vessel position first wall outer wall reactor inner wall of of blanket vessel magn. shield material SS SS SS SS First wall type and material SS SS36 with SS (welded TZM (with low- Be-coating pipe) Z coating) neutron wall load (MW/m2) 1.3 0.07 0.17 life time (years) 5 2.5 3 5 Blanket structural material SS SS 316 SS 316 SS 316 coolant He H2O/steam He(30 atm.) He(10 arm.) breeding material Li none Li Li20 thickness (m) - 0.2 0.75 0.55/0.95 Shield material SS,B4C,Pb, SS,B4C,Pb concrete, W, A;. SS,B4C, Pb thickness (m) 0.6/2 0.5/1.0 0.7/0.7 0.4/0.8 coolant He H2O He borat'jd H^O Tritium flow rate (g/h) 126 8 63 inventory (kg) 1.3 2-3 0.5 consumption (kg/year) 16 4 2.5 1) Oak Ridge 2) Genaral Atomic + ANL 3) JAERI, Japan 4) USSR 5) ANL 6) Italy. -21- 5. RADIOACTIVE INVENTORY OF REACTOR SYSTEMS 5.1 Problem exposition Of the complicated phenomena that begin with the production of radionuclides, and end with the absorption of the radiation by the human body made possible by various leakage pathways out of the reactor, we shall discuss in this chapter only the production and the radiotoxicity of these nuclides. Discussion of other aspects is postponed. The approaches of the authors of I with regard to this problem {G. Kessler for the LMFBR and G.L. Kulcinski for the fusion reactor) diverge somewhat and are not consistent. Whereas Kulcinski lates the complete radioactive inventory within the whole reactor ..em af ter two years of operation, Kessler discusses mainly the radioactivity of the quantity of fissile material yearly passing through the repro cessing facility. A comparison of both systems on such a basis is mis leading. For a correct comparison both the complete radioactive inven tories of the two systems after a fixed time of operation and their yearly amount of radioactive material to be (re)processed (fissile material and waste) should be compared. The radioactivity (in Curie) is not a measure of the potential hazard to the population. This hazard depends on the composition and the half-lives and is expressed by the biological hazard potential of the nuclides concerned. For the biological hazard potential of the sep arate nuclides, the maximum permissible concentration (MPC) in air or water during continuous exposure is considered, as determined by the International Commission on Radiological Protection (ICRP). The defi nition of the biological hazard potential (BHP) reads: BHP = radioactivity (Ci/W) * MPC-value (Ci/m3) The value of the BHP can be expressed in m3 air/W or m3 water/W: it is the amount of air or water (per unit of generated power) , required to dilute the generated radioactivity to the maximum permissible concen tration of the radionuclide in air or drinking-water. For the determi nation of the BHP of the complete system the quantities and the MPC- values of all radionuclides in the system have to be known. The required volumes of dilutant fcr each radionuclide are to be added. The BHP in air is related to inhalation and is a realistic index in case of acci dents, whereas the BHP in water is to be used in case of radionuclides released into the cooling-water, and when long-term waste storage is considered. The ICRP has changed the terminology recently. In place of the MPC one now considers the Annual Limit of Intake (ALU and derives from it the DAC (Derived Air Concentra tion) . -22- The application of the BHP-index is useful for a global compari son of the inventories of radioactive substances in different reactor systems, without giving an opinion about the possible release from the system followed by dispersion into the environment. It is rightly point ed out in the IIASA-report that the BHP-values presented do not permit a quantitative conclusion, since even under abnormal conditions not nec essarily a large part of the radioactive inventory is set free. Further more, the biological absorption of isotopes of actinides and metals from a contaminated environment is not well-known. In a complete safety anal ysis the release pathways of the radionuclides and the probability of such a release must be considered in detail. It should be emphasized that the relevant knowledge is more advanced for a fast breeder reactor. In the following comparison of the fast breeder and the fusion reactor, the simplified view is taken that at a maximum hypothetical accident the whole radioactive inventory is released. The concept of the BHP merely considers the effect on individuals during their lifetime. Isotopes with a very long half-life car* traverse the biological system more than once (i.e. several generations), each time with decreasing activity. In order to take account of these effects one can use for each nuclide i the following integral value of the BHP: IBHPi = ƒ BHPi(t)dt = BHPi(t=0) t?/0.693 , 0 where t, is the half-life of isotope i. The unit of IBEiP is ir.'W - s. The IBHP is useful only in the case of long-term storage, thus one has to take the IBHP t water In view of disposition, reprocessing, and storage, the following timesca.Ies for decay of nuclides are distinguished: - short-term (up to a few weeks): important when dispersing nuclides in case of an accident; - intermediate (up to one year): to be used at maintenance and repair; - long-term ( •• 100 years): relevant for waste storage. The next sections give, in a schematic way, the radioactive in ventories of the fast breeder reactor and the fusion reactor, according to the IIASA-report. This is followed by a, to our opinion, correct way of comparison between the two systems. Wherever possible, values have been normalized to an electric power of I GW. 5.2 The fast sodium-cooled breeder reactor The data of tne reference reactor SNR-300 are shown in table r>. I . The locations with radioactive material are the breeder reactor itself, the reprocessinq plant, the location of waste storage, and the fuel- fabrication plant. The last two locations will be discussed whon, start ing at the reprocessing plant, the fuel cycle is treated. -23- Table b.l. Reference sodiumt-cooled fast breeder reactor. Electric power 1 Gï-'(e) ] Thermal pover 2.5 C«(th) Specific power 94 MW(th)/ton U • P« metal Core -harge 3.63 t PJ02 * 2.39 t Pu 2*. 17 t L'02 * 23.1 t V Charge oi axial blanket It t L'O * 15 .93 t L* Charge of radial blanket 43.1 t tX>2 * 38.23 t U MaxiBUK bum-up core charge 77500 M*tó/t (after 2 years of irradiation; j Composition of fissile material O.it :5^U, 51.7% :?4U, 0.06% :5'Pu. 4.8% :3*Pu, 1.25 ~ :t»u. It -^-Pu, 0.12% -^Pu- Neutron flux. 4.5*1015 n/wr Coolant voxume 1.25 »J »a/MK(th) Recfcargings of the core one half of the core, once per year Table 5.2. The radioactivity and the biological hazard potential of the breeder reactor after two years of operation. ! BiiP (air) BHP (water) 10*Ci/GW(e; Ci/W(th) k»YGW(e) k»VG¥(e) Structural material (steel) 200 0.08 10-lQ7 I-10' j Coolant (sodium) 200 0.08 5- 107 0.7*10* I Actmides in fissile 1 material 3 700 1.5 2.5 -1 a:" 2.5-10' rission products 1 10 000 4 2.5-10% 3 * 10r' 1 - The total activity is limited to a relatively small volume. - The activity i:. the structural material stems for 60% frcei the cladding »>n the fuel elements in t^e core (30 t, about 5 Ci/g) . The reactor vessel (1070 t, 5.5 Ci/t) contains only 0.0021 of tne activity, this is comparable to natural uranium cieta! (0.32 Ci/t). The most important radionuclides in the structural material Are 'Mr. (t^ = 2.? hours) and To (tt - 71.3 days), contributing to the saturation activity for 501 and 23* -espectivoly. The afterheat generated in the structural -•itorial amounts to 2.r> MW/GWU-), this reduces in 180 kWA;w(eJ 180 days -24- after shut-down of the reactor. 300 years after shut-down the activity has decreased by 6 orders of magnitude. Both the"pool-type" reactor (3000 m3 Na/GW(e)) and the"loop-type" reac tor (800 m3 Na/GW(e)) contain an activity of about 200 MCi/GW(e) in the coolant, predominantly 24Na {t, = 15 hours). At the moment of shut-down (after 2 years of burn-up) the activity of the fuel actinides is 140 MCi/ton U + Pu metal, 94% of which stems from 239U (t, = 23 minutes) and 239Np (t, = 2.4 days). Due to the short half-life of 239U and 2 3 9Np the activity of the actinides rapidly reaches the saturation level, so that the activity is almost indepen dent of the irradiation period. After shut-down of the reactor the activity decreases rapidly (1.4 MCi/t after a decay time of 30 days). The residual activity is then mainly determined by 241Pu and 21+2Cm. The activity given here is that of the core, the activity of the blanket is only 10% of the core activity. In the fast breeder xeactor there is no dominant nuclide in the fis sion products. No nuclide contributes more than 2% to the total acti vity of the fission products. At shut-down cc Jhe reactor after 2 y^ars of operation the activity of the fission products exceeds the activity of the actinides by a factor of three. Figure 5.1 gives a survey of the decay or the radioactive material;? in the reactor. tJ 10-5 10"* 107L^ - •--—~pf + —( J_ »-, "+*—p -4- L_4-_..]_ 10° 101 1102 103 I 104 I»5 106l 107i 108 10,9 1010 1011 10'2 1mm 1h fday Imonthlyr 30 yr t(sec) Kig. 5.1. Specific radioactivity of various radionuclides na n function of the time after discharge. -25- - Up to about 50 years after shut-down of the reactor the fission pro ducts are the main source of radioactivity of the reactor inventory. After 10 years the activity of the fission products is still about 10 MCi/GW(e). After 100 years the Pu-isotopes begin to dominate (1 MCi/GW(e)). The activity of the structural material has levelled off to 100 Ci/GW(e). In order to determine the annual mass flow out of the reactor, the following assumptions are made: . the average residence time of the fuel in the core and axial blanket is 2.2 years; . the average residence time of the fuel in the radial blanket is 7 years; . the Pu-yain by breeding amounts to about 100 kg/GW(e).yr; . fission products are formed at a rate of 2.5 kg/GW(e).day; . the power distribution is: 90% in the core + axial blanket, 10% in the radial blanket. With these data and with those of table 5.1 we find mass flows to and out of the reactor according to table 5.3. Table 5.3. Flow of Material per year in a 1000 MW(e) LMFBR (t/GW(e)-yr) Materi al Charged Material Disch arged By Region Total By Region Total Core Pu: 1.54 Core and axial U : 10.53 blankets Pu: 1.54 U+Pu : 18.45 Axial Np+Am+Cm : 0.03 blanket U : 7.24 U : 23.23 fission Pu : 1.64 products: 0.830 U : 22.18 Radial Np+Am+Cm: 0.033 blanket U : 5.46 Radial blanket Fission products: 0.922 U+Pu : 5.37 Np+Am+Cm : 0.00 3 fission products: 0.092 -26- The total mass flow out of the reactor (core + blankets) is 23.82 t Pu + u/GW(e).yr. Its average activity at the instant of shut down is about 260 MCi/t (75% is caused by fission products and 25% by actinides), 150 days after shut-down it is 7.7 MCi/t (91% caused by fission products and 8.5% by the faster decaying actinides). After a periodic discharging of the reactor some important points of time in the fuel cycle are: . 150 days after discharging: transport to the reprocessing facility; . 200 days after discharging: chemical reprocessing; . 365 days after discharging: re-use of the separated U + Pu in the fabrication of new fuel elements; . 1-3 years after discharging: re-use in the reactor; As a consequence of the chemical separation of the fission products and the actinides Np, Am, from the uranium and plutonium fuel, the radioactive inventory of che fuel is reduced to 10% (mainly 2l>1Pu with t, = 14 "ears)of what it was before reprocessing. The waste contains another 1% of the Pu-inventory. The radioactive nuclides should be carefully handled along two pathways: from the re processing facility to the fuel fabrication plant and from the repro cessing facility to the waste storage location. One half of the radioactive reactor inventory (in Ci) is transported annually through the reprocessing cycle. Figure 5.2 gives a survey of the LMPBR fuel cycle. toss ro HIGH LEVEL LIQUID WASTE 0 22711/ 0 015 I Pu CHEMICAL 4 0 071 Pu | I 62 t Pu REFROCCSSIMG —I STORAGE FOR I REUSE | I 641 Pu I 22 >9tU I 71 96 t U LMFBR StORAGE FOR REUSE ENERG» PRODUCTION HWId yt ( L F = T ) ~s— I I . 1 SS r P+u | I FABRiUriM OF I ! 54 I Pu Pud; / UO?- J 73 231 U f PElLEfS ( .J FASRIUriO* Sf 10? • PELLETS • U.PU -• OHL* U DEPLETED UF 6 LOSS TO SCRAP '-!> WASTE 0» SCRAP 23 45 t U ..» 0 222 r U ^ 0 016» Pu Fig. 5.2. Integrated LMFBR fuel cycle system. Yearly mass flows of uranium and plutonium for one GW(e). -27- The biological hazard potential of the nuclides in the fuel cycle in clusive of the reprocessing step is represented in table 5.4 and Fig. 5.3. 5 1Q I£ M04a: iACTINIDES m Q. Puli» M03K c m INHALATION HAZARD < AmZ. .M *te^ SrSO.CiU* 2 Cm W. l-io & en X-si?-0--l FISSION PRODUCTS .HO1-* —»FUEL ONLY (FABRICATION AND REUSE IN FAST REACTOR CORE ) FI'iSION PRODUCTS IACTINIDES INGESTION HAZARD -Pu(*t- Puinr — "T 1 1 200 400 600 800 1000 REPROCESSING TIME AFTER DISCHARGE (day) F.g. 5.3. BHP of an LMFBR as a function, of the time after discharge. -28- Table 5.4 The biological hazard potential of the fuel cycle and the waste outside the reactor (in km3/GW(e)). Time after shut-down Biological hazard reprocessing potential t=0 t=150 days t=200 days t=200 days t=365 days t=10 years t=100 years t=1000 years before after Inhalation (air) Fission products 1.32*109 0.33*109 0.30X109 fuel 0.3 xio9 waste 0.30X109 0.40xl08 0.42X107 0.32x103 Actinides 1.28X1010 1.25xl010 1.25X1010 1.15xl010fuel 1.14X1010 (used agair in reactor) 0.1 xl010waste 1.0 xio9 0.70xl09 0.40xio9 0.13*109 Ingestion (water) Fi s s ion p roduc t s 2.30*105 1.12X101* 1.12X104 fuel 1.1 xio1* waste 1.0 xio1* 0.35X101* 0.38X103 2X10-2 Actinides 1.02X104 2.42X102 2.30X102 155 fuel 150 75 waste 75 25 14 4.5 As a comparison: The BHP (water) of uranium ore is 0.75 kmJ/GW(e) (the 23 tons of uranium for a 1 GW(e)-plant are gained from 16450 tons of uranium ore). Accordingly, after 1000 years the BHP of the reactor activity is a factor of 6 higher than that of the mined uranium ore. - The BHP(air) is mainly determined by the actinides (Pu), whereas the BHP(water)is determined by the fission products. - The high-level waste is stored for a few years in steel tanks for cool ing. During this time the short-lived isotopes disintegrate, so that cooling requirements are much relaxed thereafter. The waste is then vitrified and the glass blocks are stored in air- and/or water-cooled buildings for approximately another 50 years. Then they are transported to the final storage (geologic formations such as bedded salt or salt domes). From this moment, the only way by which waste could reach the biosphere is the leaching of the waste by water, therefore for long-term storage the BHP(water) is the relevant parameter. - For the complete reactor inventory and the yearly waste flow from the reactor the IBHP(water) is given in table 5.5. Table 5.5. IBHP(water) for a 1 GW(e)-reactor with reprocessing facility. 1 Reactor IBHP(water) km3 s/GW(e) Waste reprocessing km3s/GW(e)/year facility Structural material 1011 Structural material 5*10i0 Fission products 2.5xl013 Fission products 1.3*1Q13 Actinides (without Actinides (Pu in Pu) 2.5*1012 cluded) 1.7*1012 Pu 5.7xl013 5.3 The fusion reactor The design of the reference reactor, the 1 GW(e) UWMAK-I tokamak fusion reactor, is less detailed than the fast breeder design of section 5.2. As a consequence, it is far from being optimizec. Table 5.6 shows the blanket composition. Table 5.6. Blanket composition of the fusion reactor UWMAK-I. Torus major radius: R = 13 m Thickness of the first wall: 0.4 cm First breeding section Torus minor radius: r .. = 5.5 m (95% Li, 5% SS) : 51 cm Neutron wall load : Pwal l= 1.25MW/m7- Reflector section (SS) : 15 cm Structural material: w,SSn Second breeding section (95% Li, 5% SS) : 5 cm Coolant: Liquid Li Supporting wall (SS) : 2 cm ...... _ .. . The two major sources of radioactivity in the fast breeder reactor, the actinides and the fission products, are not present in the fusion -30- reactor. The BHP is thus much smaller and this is one of the arguments for development of the fusion reactor. However, the structural material activated by neutrons constitutes an important problem. In particular, the activated volume is larger than in a fast breeder. Obviously, the amount of radioactivity will greatly depend on the choice of the reactor system. Some parameters affecting this amount are the type, the choice of the structural material, the breeding ma terial, and the coolant. From table 5.7 it is seen that the main sources of radioactivity are situated in the blanket. Other activated materials, such as the shielding of the magnets and the biological shield have a negligibly low activity. Table 5.7. Radioactivity in the blanket of a DT fusion reactor (per GW(e) and after two years of operation). ! mass activity BHP (air) BHP (water) (kg) (xlO6 Ci) (km3) (km3) Tritium max. 25 max. 250 lxio6 lxlO2 Structural material 4xl06 2000-13000 0.7-10*10e 1-70X10'1 Coolant lxio6 75 Tritium For the calculation of the hazard potential of tritium, one mostly assumes the most unfavourable form of tritium absorption, namely as tri- tiated water (T_0, H_0) and not as molecular tritium (T2, HT). The MPC- values for the population are: MPC(water) = 3xl0"3 Ci/m3 and MPC(air) = 2*10~7 Ci/m3 . &t a specific activity of tritium of 107 Ci/kg the biolo gical hazard potentials become: BHP(water) = 3.3*109 m3 water/kg T and BHP(air) = 50 000 km3 air/kg T, For the reference reactor the most important sources of tritium activity are given in table 5.8. Table b.8. Sources of tritium in a tokamak fusion reactor. Fuel injection system (collection and purification included) 2 kg/GW(th) Breeding material of the blanket and extraction system 0 05-2 kg/GW(th) Storage outside the reactor 5-14 kg/GW(th) 7-18 kg/GW(th) Tritium inventory = rounded off 10 kg/GW(th) == 25 kg/GW(e) -31- - A continuous power production of 1 GW(th) requires the daily fusion of 0.14 kg tritium and 0.09 kg deuterium. At a 1% burn-up of the injected fuel and a residence time of 4 hours in the total circulation system of tritium, about 2 kg of tritium is needed in the injection system (per GW(th)). - The tritium inventory in the breeding material of the blanket great ly depends on the breeding medium and the blanket design. For UWMAK-I with liquid lithium, a value of 2 kg T/GW(th) is mentioned; for designs with solid lithium compounds this number is two orders of mag nitude lower. - The undisturbed continuation of the power production in case of fail ure or repair of the tritium purification system demands a certain tritium reserve in the storage system. A reserve of 14 kg T to bridge a repair time of one day can be stored eesily and safely outside the reactor system. The hazard potential of this amount is much less than that of the tritium in the high-temperature blanket where the proba bility of some malfunction is much larger. It is sometimes argued that the tritium reserve may be taken much lower because the interruptions mentioned will stop the power generation anyhow. A£tiy^te^_str^c^u£al_material The activity of the structural material, although dependent on the choice of material, is of the order of 1 Ci/W(th), see table 5.9. Table 5.9. Activity of the structural material in the blanket of a fu sion reactor (after 2 years of operation, per GW(th)). 1 Specific activity of the first wall at time t Structural Activity at after shut-down of the reactor material shut-down (Ci/cm3) (10& Ci/GW(th)) t = 0 t = 1 day t = 1 year t = 100 years SS 1062 100 68 29 5*10~3 TZM(=Mo) 4120 125 83 0.04 7xl0-3 V-Ti 1261 27 6.6 0.31 < 10"37 Nb-Zr 5155 158 94 6xl0-'( 10" 5 Al-alloy 884 44 8.7 0.3 10"5 Ti-alloy 332 25 19 0.93 8xl0"7 natural uranium 6*10"° Ci/cni3 -32- The compositions of the alternative materials in table 5.9 are: SS = 63% Fe, 19% cr, 12% Ni, 2% Mn, 2% Si. TZM = 98.9% Mo, 1% Ti, 0.1% Zr. V-Ti = 84% V, 16% Ti. Nb-Zr = 99.25% Nb, 0.75% Zr. Al-alloy = 94.5% Al, 3.02% Mg, 2.15% Cu, spurs of Si, Cr, Mn, Fe, Zn. Ti-alloy = 90% Ti, 6% Al, 4% V. - All materials have advantages and disadvantages. Al has a low activity but also a low melting-point, a high (n, a) cross-section, and it can not be used in contact with liquid lithium. There is presently no well- established industry (or mining capacity) that could supply the neces sary tonnage of refractory metals Nb, Mo, V, required for a well- established fusion economy. Their irradiation behaviour is insuffi ciently known. Joining techniques for Mo-alloys are difficult to im plement. Stainless steel has a mature industry to manufacture million ton quantities under strict quality-assurance standards. Its behaviour under irradiation is known best. However, it has problems with lithium compatibility. - The accumulation of the activity depends on the time of operation. After one day of operation all materials have reached more than 50% of the saturation activity, after 10 days this percentage is 90 (except SS). After 2 years of operation the radioactivity hardly changes any more . - The distinction between three timescales for the decay of radioactivity after shut-cown, made for the fast breeder reactor (see section 5.1.), is also valid here. The afterheat is significantly smaller than in a fast breeder reactor and spreads out over a larger volume. At the instant of shut-down of the reactor, the activities of the unfavourable (maximally activated) materials and the favourable (minimally accivated) materials differ a factor of 6. One day after shut-down this factor will be 50, after one year more than 1000. After 100 years, the specific activity (in Ci per gram) of the material of the first wall, is often even lower (e.g. vanadium) or of the same order as the specific activity of uranium. SS and TZM are not suitable for re-use after long-term storage, these materials would have to be stored and monitored for longer periods of time. - The biological hazard potential of the structural materials after shut down of the reactor is summarized in table 5.10. -33- Table 5.10. BHP of the structural material of the fusion reactor after shut-down of the reactor. ...... _. . . — 3 3 3 km / BHP (water) (km /kW(th)) km /cm BHP (air) (km3/kW(th)) r cm mat. t=U t=l I t=100 mat. t=l t=l t=100 t=i t=0 t=ioo ! day year 1 years t=0 day year years years 3 ss 156 150 SO 0.1 2 0.01 8*10~3 lO" 4*10-7 5.4 TZM 391 300 0.15 10"3 2.9 0.05 0.03 10"5 lO"7 1.0 V-Ti 27 20 1.5 co-io 0.2 4xl0-3 2xlO-3 4xl0_t* < 10"25 2xlO-30 rib-zr 39 30 0.01 4xl0_l* 0.36 0.26 0.20 5*10~7 lO"7 0.6 Al-alloy 7 3 20 3 lO'3 0.55 0.01 3*10~3 2X10"5 2xl0-8 0.2 Ti-alioy 91 81 7 4xl0-5 Natural uranium 1.3'-10~3 0.2 - The values of the BHP (air), to be used in case of accidental release, show a variation of a factor 15 for the materials considered, and are roughly constant during the first day after shut-down. - The values of the BHP (water), to be used at long-term storage,will have dropped considerably after 100 years. The activity in the V-20Ti alloy then has dropped to insignificant levels and the activities of the Al- alloy and Nb- Zr have decayed to values comparable to that of natural uranium. - The resulting IBHP (water) per cm3 material of the first wall is a maximum for Al (46X101* km3s/cm3) and a minimum for V (2xio'' km3s/cm3), but is still one to two orders of magnitude higher than that of natural uranium {2*103 km3s/cm3}. For stainless steel this value is 23*10'* km3s/cm3 . A£tivated_coolant - The reference reactor contains about 10b kg of liquid lithium per 2.5 GW(th) (=lGW(e)). The induced activity is mainly caused by impurities (about 30 Ci/kW(th), mainly F and Cl) and to a lesser extent (about 1 Ci/kW(th)) by corrosion products (1000 kg per year per GW(e)). - The most serious danger of the radioactive inventory lies in the activity of the liquid litnium, because in the event of a rupture of the primary coolant loop and subsequent ignition of the lithium the formed radioactive oxides could be widely dispersed. - Cooling by lithium is not necessary, several reactor studies take cooling by He, using solid lithium compounds for breeding. The induced activity of 'He (t^ = 10~?1 s) becomes 500 Ci/kW(th). The above-men tioned fire hazard can be avoided by proper engineering. -34- 5.4 Comparison between fast breeder and fusicn reactor The following comparison between the fast breeder reactor and the fusion reactor with regard to the radioactive inventory deviates from that in I. in the latter report the long-term behaviour of the invento ry of the fusion reactor is almost exclusively compared to the yearly waste production of the fast breeder. The waste from the reprocessing facility is supposed to be free of Pu, this Pu is thought to be .re cycled into the reactor. When considering the short-term safety, one should take into account for the fission reactor both the complete reactor inventory and the discharged fuel elements that are not reprocessed, whereas for the fusion reactor data of different parts have to be specified. To find the consequences of the long-term storage of waste a comparison of the values of the BHP (water) connected with the yearly waste flow is of primary importance. The IIASA-report does not calculate the yearly flow of radioactive material from the fusion reactor. An estimate can be made as follows. With a tritium breeding ratio of 1.25 and a tritium consumption of 0.14 kg/GW(th).day the yearly amount of tritium produced in the blanket is 63 kg. 51 kg of this is reinjected into the reactor and 12 kg is periodically removed from the system to be used in other fusion reactors. Per GW(e) the annual tritium flow out of the reactor is then 30 kg. Radiation damage sets a limit to the permissible lifetime of various reactor components, in particular the first wall. The lifetime of the first wall and the first 20 cm of the blanket in the reference reactor is assumed to be 2 years (some other design groups accept a lifetime of 5 years at a comparable wall load) . 50% of the radioacti vity will build up in this front part of the blanket. The more remote parts of the blanket (with a lower radiation load) are assumed to be exchanged once per 10 to 15 years. Thus, per 2 years, 75% of the radio active inventory of the structural materials (iu Ci) is taken out of the reactor. This part consists exclusively of non-volatile nuclides and T. The low power density in the blanket of the fusion reactor causes a considerable flow of high-grade materials, that amounts to about 400 tons/year «GU(e) or about 50 m3/year • G'W(e) . The annual flow of activated corrosion products is about 1 ton/year. For both types of reactor, table 5.11 gives the radioactive in ventory and the BHP, normalized to 1 GW(e) and 2 years of operation. Figures 5.4, 5.5 and 5.6 show the activity and the BHP as functions of the time after snut-down. -35- Table 5.11. Reactor inventory in Ci per GW(e) after 2 years of operation at moment of shut-down. Fast breeder reactor Fusion reactor activity BHP (air) BHP (water) activity BHP (air) BHP (water) 3 Ci/GW(e) km /GW(e) km3/GW(e) Ci/GW(e) km3/GW(e) kjt)3/GW(e) Structural material 0.2<109 1 xio8 1 xio1* Structural material 2.13X109 0.7-lOxlO8 1-70XKJ1* (if with SS) (2.5 xjo9) (4 xio8) (2.5 xio1») Sodium coolant 0.2x109 0.5x108 0.7xl04 Lithium coolant 0.1 xlO9 Actinides 3.7xl09 2.5x10*° 2.0X101* Tritium 0.25xl09 O.lxlO7 1 *K)2 Fission products 10 xio9 0.2xl010 5.0xl05 Total 14 xio9 2.7*1010 5.5X105 f min. 2.5 xio9 0,7xi0B 1 xiQ1» Totali max. 14 xiO9 1 xio9 7 xio5 [ with SS 3 xio9 4 xio8 2.5 xio1* Tab-1 e 5.12. Activity and BHP of the annual waste flow at moment of shut-down (per GW(e)). Fast bxv.eder reactor Fusion reactor activity BHP (air) BHP (water) activity BHP (air) BHP (water) Ci/GW(e) km3/GW(e) kra3/GW(e) Ci/GW(e) km3/GW(e) km3/GW(e) Structural material O.lxlO9 5 xio7 5 xio3 Structural material I xio9 1.6X108 lxio'* (400 t/year SS) Corrosion products Corrosion products 2.5xl06 Actinides 1.8xio9 1.3xl010 1 xlO4 Tritium O.lxlO9 3 xlO6 30 Fission products 5 xlO9 O.lxlO10 2.3XJ05 Total 7 xio9 1.4X1010 2.4X105 Total (with SS) l.lxiO9 1.6X1Q8 lxlO'* » to* reactor inventory LMTOR - blanket inventory' fusion reactor iSS) 3 io: \ Z io -i 10"- 10-3 n-« io-5 \ 10lJJ y.- 10* v 10'' 0 2 < 6 8 13 12 1« log time after shutdown (s) •• Fig. 5.4. Comparison of the radioactive invencorifc.i of a fusion reactor and a breeder reactor (both with. SS as the structural material). 10 = discharged fissile fuel of LHFBR reactot inventory LM?BH u j< 10- reprocessing I o ^ k X 4 10' -O ri e e annual waste o c blanket inventory \ -t o of the fusion reactor , 0 V O -> ïï 10 -t V \\ annual waste* io-^ after reprocessing io-3 io-4 ly 10-1 y. \ I0ft y io-5 0 2 4 6 8 10 12 14 log time after shutdown (s) » t-'ig. 5.5. Comparison of the biological hazard potentials for inhalation of a fusion reactor and -i fast breeder reactor. -37- I , i*-^ 5 * •7 S 19 19-4 - I3"*r - -3 «s- f I ..-4 i t io-+ 9 2 4 * 1 !« 12 14 loq *>•»» after tóut*o#a 's'. —»• :-i ?. J.» for irirüci:. j-w*t.«tr -Ï* Ï f*sij;\ r«.- t If the structural arterial of the fusion reactor is wc-il-choser. (SS, V or AL), its radioactivity i.. Ci r.i" be made a factor of 6 lover than that in the fast breeder reactoi . The sr.ain sources of activity are the fissior. pi. jucts in the fast breeder and the activated structural r&- teriai in the fjsijr reactor. Appr-,/.ïsiateiy one year after sh-.it-do*#n the number cf Curie", of ts«o ir.ver.tori s of both -ystep-s are ej.il. After 10 years, howevev, the ictivity oi the tusiv. 'eac^.or inve--.'.-.ry drops mach faster than the act-iwty of the lon - ift- If one chooses V or Ti as the structural material, the long-terra activity can be reduced still more. Recently, Kulcinski et a!. pub lished calculations that jusLify the re-use of activated structural ma terials after an acceptable decay time. At these calculations the BHP for inhalation is compared with an analogous chemical toxicity index, the so-called Industrial Hazard Index (IHI). If the value of '-he BHP(air) becomes lower than the IHI, reprocessing even in the conventional metal- reprocessing facilities can be justified. It was found that V, Cr and Sn satisfy this criterion within 10 years, chiefly as a consequence of the low specific radioactivity. The radiation level of these materials after 10 years becomes very low. It is smaller than 10~8 mrem/hour fcr 1 cm3 of material at a distance of 30 cm, i.e. a radiation six orders of magnitude lower than the natural background radiation. The rather low MPC-value and the residual activity of some iso topes prevent Al, Nb and Mo to reach the lower limit. 5.5 Conclusions Depending on the time after shut-down, the fusion reactor has a BHP one to four orders of magnitude lower than the fast breeder reactor. The biological hazard potential of the breeder reactor is inherently determined by the fission process, namely the BHP(air) by the actinides (Pu), and the BHP(water) by the fission products (l29I). The BHP of the fusion reactor is not coupled to the fusion process, but is mainly de termined by the choice of the structural material. Development of materials with good mechanical characteristics, a low induced radioactivity, and a low BHP could make the advantage of the fusion reactor still greater. A possible disadvantage of the fusion reactor is the relatively large amount of highly-active, high-grade materials, that yearly leave the fusion reactor (50 m3/yr as compared to 10 m3/yr for the breeder reactor). The fast breeder also has a yearly flow of 1400 m3 of mc ium- and low-level solid and liquid waste. Besides being chemically more stable, the waste of the fusion reactor is less dangerous because of the low BHP. Even waste storage in the vicinity of the fusion reactor plant is not to be ruled out (see also chapter 7). -39- 6. EMISSION OF RADIOACTIVITY AT NORMAL OPERATION 6.1 The fission reactor Table 6.1. gives a summary of the main radioactivity releases from a reactor power plant (fuel fabrication and reprocessing included) during normal operation. Table 6.1. Released Radioactivity During Normal Operation of an LMFÜR Power Plant. Release Rate (Ci/GW(e).yr> Into the Atmosphere Into Water H3 350 350 Kr85 0.4 (300*) - Xel33 0.03 (4200*) - 1131 0.01 0.01 Conservative number for reactors being presently built. The EPA (U.S. Environmental Protection Agency) has proposed standards for the release of radioactive nuclides from reactors. These standards are still under discussion but the values lie close to those ir. table 6.1. The confinement factor is defined as the ratio between the rate of flow of radioactive material through the fuel cycle and the release rate. The confinement factors for plutonium a-emitters, for 12-'I, and for 85Kr must be 2xio9, 10u and 10 respectively. According to I, these factors have been achieved wherever necessary, and have in some cases been exceeded. Present-day safety regulations determine the complete fuel cycle already in its design phase. The authors of I do not discuss the desirability to complete the safety analysis with an investigation of the possible pathways, along which radioactive isotopes can reach the population (thereby replacing Curies by mrem/yr). This would be an extensive work in which the mutual influence of a large number of local parameters is difficult to calcu late. A more global treatment would be the replacement of Curios by the time-integrated BIIP. For this IEnP the total number of initially excited nuclides and their specific decay reactions are taken as a starting-point (see chapter 5). 6 . 2 The fusion reactor The most volatile part of the radioactive inventory of the fusion reactor is the tritium. This emits relatively harmless (;:-radiation , but the considerable inventory makes it a non-negligible problem. If HT occurs in the air, the concentration should not exceed the Ml'C-value of 4-<10~' Ci/m''. ItTO is more readily absorbed by the human body and thus -40- the permissible concentration in air is smaller (a factor 200}. To he on the safe side it is assumed that all - "itium occurs in the form of HTO. Equilibrium considerations do not contradict this, but in view of the relatively slow conversion cf HT to HTO it is possible that the hazards to human beings near the sources of tritium are overestimated by as much as a factor of 200 by this assumption. The tritium inventory of a PWR of 1 GW(e) is about 2*10^ Ci. The yearly tritium production of this PWR is about 1.5*10lf Ci. 40% of this is released through the cooling-water into the environment. Tne remaining part is released during the reprocessing. Thus, the fractional daily loss of the tritium inventory in the reactor is 10~3. This is perfectly permissible in a PWR. However, in a fusion reactor w?.th a tritium inventory >107 Ci and with tritium as a fuel cumponent and not as just one of the many decay products, a better tri'ium control must be assured. We can find the necessary conditions as follows: The permissible release of HTO in water is found with a large safety margin by considering an individual that takes all his or her drinking-water from the cooling-water of the fusion reactor. This limits the annual permissible release to 105 Ci (0.07 kg HTO). Since the amount of tritium in the fusion reactor is about 2.5 kg or 2.5*107 Ci (not counting the tritium supply outside the reactor), the maximum permissible annual discharge of HTO is 4><10~3. This means a daily leakage of the inventory of 10-5, so the degree of control has to be two orders of magnitude tighter than that in a PWR. For emission of HTO from a 30 m stack the required control factor is also one part in 10s per day. In the case of HT-emission and/or a higher stack, the situation would become more favourable. If we would again relate the confinement to the annual (circula ting) mass flow, a "burn-up" of 1% and an annual tritium consumption of 120 kg/GW(e) would lead to a tritium flow rate of 12.000 kg/yr. This demands a confinement factor of 1.2x10". It seems technologically possible to realize such a confine ment factor. Since the task is eased by a smaller tritium inventory, some reactor designs try to reduce the inventory to about 1 kg. The confinement can be further improved by separation of hot and cold tritium, by cooling of exterior walls and by permeation barriers such as copper surfaces on steel. There is a strong economic incentive to minimize the escape of neutrons from the blanket region. It is also essential to protect the superconducting magnets from the neutron fluxes. Nevertheless, the great penetrating power of the neutrons and the complex geometry of -41- the fusion reactor interior (injection and pumping ports) will make a complete shielding of the immediate area of the reactor difficult. Sufficient shielding should be provided within the reactor building so that the neutron dose outside is reduced to any desired level.Then neutron exposure will be an occupational» not a public hazard. Another problem is posed by the activation of the reactor compo nents. An exchange of structural parts is not considered here. It is assumed that during normal operation activated material is not able to escape. It is not yet clear if gaseous isotopes will occur or if gases from vacuum pumps require additional safety measures. Metals could oxidize and the oxides could be volatile. 6.3 Conclusions For the breeder reactor it is essentiaJ .0 confine ]2gI, 95Kr and plutonium. Necessary confinement factors are 2xl09 for transuranium isotopes, 2x10^ for iodine, and 10 for krypton. The tritium inventory of a fusion reactor is much larger than that of an LMFBR. The report I considers a total inventory, in a not yet optimized design, of 25 kg T = 2.5xlC3 Ci, being a factor 101* larger than the inventory of an LMFBR. The analysis of the tritium control by I shows an incongruity. In our report we showed that leakage of the complete tritium inventory from the fission reactor is permissible. In the conclusion of I it is wrongly argued that, since the tritium inventory of the fusion reactor is 104 times as large as that of the fission reactor, the confinement of tritium in the fusion reactor has to be better in the same proportion. he have shown, however, that the confinement factor has to be only a factor 102 higher than that in a fission reactor. Although not yet achieved during continuous operation in large experiments, the required confinement factors appear to be within reach, and demonstration facilities are under preparation. 7. DANGER OF ACCIDENTS 7.1 The fast breeder reactor One can distinguish three phases in the development of safety studies for fast breeder reactors: I. Between 1949 and 1959 much attention was given to small breeder reactors (< 66 MW(e)) with metal fuel elements. Safety analyses focussed on the criticality of the reactor. II. 1959-1970 was the period of the development of large breeder reac tors (- 1 GW(e)) with oxidic fuel. Safety considerations concerned -42- the Doppler coefficient, the sodium-void coefficient, the sodium super-heat, the fuel/sodium interaction, and fuel failure detec tion. III. The overriding feature of the third phase after 1979 is a wide spread and multilevel approach of testing the elements of the design concept that arose from the second phase. In the safety studies accurate probability calculations were introduced. The mostly favourable results were not sufficiently communicated to the scientific public and the technical community. In I it is argued that the handling and control of fast breeder reac tors differ not essentially from those of thermal reactors. This is based on the similarity between the electro-mechanical designs of the control system. Accident chains are considered on two levels: realistic accidents (probability > 10-5 per year, can be avoided in principle by a proper design of instrument control of shut-down systems) and hypothetical accidents (probability < 10-5 per year, are covered by the strength of the primary cooling system and that of the surrounding containment system). In I a hypothetical chain of accidents is considered, consis ting of ccolant blockage in fuel elements, local boiling of sodium, damage propagation within the considered fuel element, integral gross boiling, melt-down of the fuel and sodium-fuel interaction, and finally pressure build-up and coherent core compaction. There are a sufficient number of indications, both experimentally and theoretically, that the complete chain of accidents is highly improbable. With an analysis of an accident with the SNR-300 as an example, I concludes that the LMFBR has a safety standard like the other reactor- types . There exists an alternative view of reactor safety, namely that of F.R. Farmer. He rigourously stresses the probabilistic approach to safety. He emphasizes the importance of accident prevention by early detection of irregularities through sophisticated instrumentation in stead of protective measures for the case of an improbable accident. An important aspect is the safety of the fuel cycle, comprising fabrication of fuel elements, reprocessing, intermediate waste storage, waste solidification, and the final waste storage. From a theoretical study mentioned in I, using the notion of expectation values and per mitting a radiation dose per individual of 25 rem in a 70 year life time, it is found that waste storage is the most critical step in the fuel cycle. -43- There is a discrepancy between the service capacity of a repro cessing facility (about 45 GW(e)) and present-day optimal power of one reactor (about 1 GW(e) if energy is transported in the form of electric ity) . Besides servicing many reactors that stand a large distance apart, the reprocessing facility could be situated in a reactor park. In the latter case one should switch over to a more effective energy vector than electricity, such as hydrogen. 7.2 The fusion reactor The considered dangers of a tokamak reactor to the population are: - the release of tritium; - the release of radioactive corrosion and sputtering products in the coolant; - the release of radioactive structural material and blanket material; - the release of non-radioactive but toxic materials. Possible accidents are: - A rapid release of the nuclear energy represented by the fuel con tained in the plasma. This is an unlikely event on the basis of pres ent knowledge of plasma behaviour. If somehow the entire quantity of fuel in the plasma did react, the 70 GJ that evolved would raise the blanket temperature by only about 100 °C. - If the kinetic energy of the plasma (< 3 GJ) would be deposited on a small section of the first wall, only a local burn-through could result. - Release of the magnetic energy (about 200 GJ). This could lead to very large forces (on the order of 105 t) on the superconducting magnets, so that structural failure perhaps cannot be entirely ruled out. - Loss of coolant flow. The first wall could reach a temperature of 600°C, followed by a rupture of the wall and quenching of the reaction. Such an accident would be disruptive and expensive, but not catastrophic. - Complete loss of coolant, followed by shut-down of the reaction. The afterheat in the blanket is about 1 W/cm3 and entails a temperature rise of 0.1 C/s, a much smaller value than occurs in a fission reactor. - A lithium fire. This may well represent the "maximum hypothetical ac cident". It could lead to ruptures in the structural material at tem peratures above 2000 K. It could thereby augment the volatile tritium inventory that could escape through breaks by converting activation products and toxic non-radioactive metals to volatile form. Therefore, lithium will be used in a compound which cannot burn. -44- The conceptual designs for fusion reactors incorporate many fea tures to minimize the probability of accidents, and to reduce the con sequences if one occurs. These features include double-walled piping, multiple separate lithium loops, stainless steel liners, multiple, widely separated storage bunkers for tritium, and barriers between liquid lithium and the steam cycle. The crucial question remains: How flexible is the choice of coolant and breeding material, in order to maximize the safety? Relative to radioactive materials, the hazard of toxic, non-radio active materials is 2 to 3 orders of magnitude lower. The most dangerous material is beryllium (diluting volume 0.5 km3 air/kg Be), and the dan ger of Pb and Li is not to be underestimated. Be and Pb do not play a role in all blanket designs, they have been suggested to diminish the tritium inventory. 7.3 Consequences of an improbable hypothetical accident War, sabotage, etc., and the psychological effect of a large number of casualties in a very short time make it necessary to face the task of calculating consequences of highly improbable events. There are four levels of consideration concerning radioactive danger: 1) the total inventory of radioisotopes in the reactor; 2) the maximum biological hazard potential in the reactor; 3) the actualization of the BHP by movement of materials along pathways into and through the environment; 4) injury to human beings. A comparison of fusion and fission should best be made on level 4), its realization, however, is very difficult. The consequences of acci dents with an LWR have been analyzed in the U.S. Nuclear Regulatory Com mission's Reactor Safety Study (RSS), also called the Rasmussen Report. The method followed herein can be used for the comparison of fission and fusion reactors although the specific analysis of the fast breeder reactor is still missing, and in the case of the LWR the most unlikely and worst hypothetical accident scenario has not been considered. The risk analysis must involve four steps: - evaluation of the fraction of inventories released and the form of re leased materials; - analysis of the dispersion of released materials in the environment (dispersion models); - relation of environmental concentrations of radionuclides to absorbed doses of radiation in critical organs; - dose-response relations and population densities exposed by various pathways, to estimate expected casualties. -45- In a simplified consideration critical doses are compared in their effect on the bone marrow. For the case of a fusion reactor, tritium is the most abundant volatile isotope. It is taken in as tri tiated water and is absorbed by the bone marrow. One Curie in tritiated water gives a radiation dose of 77 rem to an average individual. The critical dose is taken to be composed of the sum of the external dose of Y~rays (absent in the case of fusion) of a passing cloud, the exter nal dose of emitters having reached the ground level, the internal dose due to tritium during the first 7 days (inhaled during the passage of the cloud), and one half of the internal dose received through the same cause from the Sth to the 30th day. Calculated with the aid of a formula in RSS, the critical dose becomes 45 rem per Ci absorbed tritiated water. Figure 7.1 gives the critical dose to bone marrow plotted versus the area around the reactor receiving this dose or more. For a serious accident with a fission reactor it is supposed that 40% of the I, Te and Sb present, 20% of the Cs and Rb, 5% of the Ba and Sr, 3% of the Rh, Ru, Co, Mo and Tc, and 0.5% of the actinides escape. For the fusion reactor it is supposed that 40% (or 10° Ci) of the tritium escapes. Some lethal doses are indicated in the figure, where LD . means that x% of the pop- x ulation in the region concerned dies within y days (LD indicates the lethal d<_ e at intensive medical treatment) . The results in the figure show that the fusion reactor is better off by a factor of 40 to 50 if compared to the fission reactor (this factor may be underestimated be cause in the case of fission, effects outside the bone marrow have not been taken into account). For both types of reactor, late effects of re leased activity have not yet been considered. Furthermore, the possible escape of activated structural material of a fusion reactor was left out of the risk analysis. The structural materials are non-volatile, so that their release is almost impossible. However, if unexpectedly 1% of the activated structural material is released,its hazard potential exceeds that of the escaping tritium. 7.4 Summary Core recompaction in the fast breeder reactor is considered im possible. Compared with the LWR, licensing is coupled to equally strin gent conditions, in other words, both fission reactor types are equally safe. Risk analysis vilh respect to contamination of the environment shows that the fus.ion reactor inventory is about a factor 50 smaller than that of the fission reactor. For both reactors the accident analysis of radioactive waste management will still have to be done. -46- M «U E 104* c«I o \^ PWB \fission release RSS VI - 13 I io3- 10 O 510 R 50/60 340 R 50/60 200 R 1/60 iO2 N^usion reactor \ \l08CiHTO \25 R — ICRP emergency dose limit 10» • =with extensive 1 - 1 t i 2 medical 10",-11 1 10* 10 care area receiving given dose or more (km ) Fig. 7.1. Critical dose to bone marrow at a hypothetical accident according to the Rasmussen model. 8. PROTECTION AGAINST ABUSE The problems of interest here concern the non-proliferation trea ties and protection against offenses by private groups. 8.1 Nuclear explosives For the production of a nuclear bomb one needs technical knowl edge and materials. It is assumed that in many industrial countries the knowledge is present or at least within reach. In many (non-nuclear) nations the materials for a fission bomb are supposed to be missing, A supply of fission reactors to these nations is one way to obtain the material, but not the cheapest and (probably) not the simplest one. The non-proliferation treaties have been concluded in order to stabilize the number of nations with nuclear arms and the IAEA (Vienna) has the task to signalize any non-observation of the treaty rules. -47- For a hydrogen bomb one needs a fission bomb as an ignition mechanism, specialized knowledge to handle LiT, LiD and neutron reflec tors, and one has to possess these materials. Thus, in comparison with a fission bomb, the construction of a thermonuclear bomb is much more difficult. Advanced knowledge of and experience with inertial confine ment (e.g. laser ignition) could make ignition by a fission bomb ruper- fluous. However, in view of the large dimensions of the beam system and of the energy supply, this mechanism is unfit for the production of bombs. The proliferation problem of a hybrid fusion-fission reactor with breeding of fissile material in the blanket, is comparable to that of a LMFBR; this aspect is called "neutron diversion". Theft of fissile material from reactors is very difficult. Technically it could be easier at other places in the fuel cycle. Theft of tritium from a fusion cycle is difficult because the fuel cycle is in or close to the reactor. 8.2 Radiological weapons Even without nuclear explosion radiological poisons could be used as a threat. Their dispersion could be accomplished by means of a chemical explosion. We do not consider the question if other, non- nuclear methods of poisoning people would not be easier to perform. From table 8.1 it can be seen that dispersion of plutonium in air is more dangerous than dispersion of tritium. Normalized to quantities of Pu and T per unit of installed power, the radiological hazard of tri tium is two orders of magnitude lower than that of plutonium. According to I, dispersion of Pu or T in water would result in a comparable risk. However, of the two threats of contamination that of air appears more direct and more difficult to counteract. Other radioactive materials such as fission products and acti vated structural materials could also be dispersed, but these would be more difficult to transport and to handle. They could be dispersed in the vicinity of the reactor by sabotage. -48- Table B.1. Radiological Hazards of Plutonium and Tritium (quantities normalized where appropriate to 1 GW(e) of capacity) PlutonJurn Tritium Inventory Outside Blanket kg 900 25 Annual Flow Outside Reactor kg 1600 32(.) MPC . Ci/km3 air Insoluble Pu239, HT or T2gas o.ooi - 40,000 Soluble Pu2 39, HTO vapour 0.00006 200 BHP km3 of air per gram of Pure Pu239. elemental T 63 to 1000 0.25 Reactor Pu , T in HTO 300 to 5000 50 BHP/GW(e)«yr 106 km3 of air per year (d) Best case , , 450 0.008 (e) Worst case 7500 1.6 'f) c -> BHP/GW(e) 106 km3 of air (d) Best case , . 270 0.006 Worst case 4500 1.25 MPC Ci/km3 for wate t r , , , , soluble forms 5000 3,000,000 BHP m3 of water per gram of Pure Pu239 in soluble compound 12,500 Reactor Pu in soluble compound 62,500 Pure T in HTO 3,JOO,000 BHP/GW(e)«yr km3 of water per year 94 110 BHP/GW(e) ' km3 of water 56 IBHP/GW(e) km3'year 2,000,000 15C0 a) At breeding ratio = 1.25; b) Contains Pu23Q, Pu239, Pu240, Pu241, Pu242; c) Based on flow outside reactor; d) Reactor-grade Pu dispersed in insoluble form, tritium dispersed as T gas; e) Reactor-grade Pu dispersed in soluble form, tritium dispersed as HTO vapour; f) Based on inventories outside blankets. -49- 8.3 Safeguards Attempts to establish an international supervision on nuclear energy have been made ever since 1946. In 1969 the Treaty on the Non- Proliferation of Nuclear Weapons became effective with the intention to signalize any non-peaceful application of fissile material. Since that time the concern for safeguards has broadened to more active measures of protection. With respect to both detection and protection, it is mostly the reprocessing facility that deserves attention. There are four classes for need of protection: almost no protection (irradiated material), some protection (uranium with enrichment of less than 5%), significant pro tection (Pu, 233U), and high degree of protection (uranium with enrich ment of more than 20%). This division is based on two factors: self- protection of radioactive material by emitted radiation and the indus trial effort concerning isotope enrichment that is necessary before abuse becomes possible. 9. MATERIALS AND RADIATION DAMAGE 9.l The fast breeder reactor Por the fast breeder reactor the material research is concentrated on fuels and fuel-cladding. Research on the structure of the core, the pressure vessel, c^tc. is less important and is not considered here. iLüël-deve^ggnient From the point of view of available methods of fabrication, eco nomic aspects, and mechanical and thermal behaviour, the mixed oxide fuels are the most attractive. That is why research is concentrated on these fuels. The mixed carbide and nitride fuels have a higher burn-up and power density, but they have not yet been intensively studied. The important present-day research aspects of the mixed oxides are closely related to the high burn-up and the high specific power density. Important effects are the changes in physical, mechanical, and chemical characteristics, a.o.: . influence of radiation-induced sintering and swelling of fuel, caused by the fission products, on changes of the fuel density; . strong mechanical interaction between fuel and fuel-cladding during power changes; . thermal diffusion of plutonium, oxygen, and solid fission products due to high temperature gradients, that leads to a redistribution of -50- plutonium and possibly to corrosion of the inner side of the cladding (mainly by 0 and Cs). Experiments up to > 105 MWday have already been performed. Exper iments in Phenix and BN 350 have shown that at radiation up to 6X101* MWday no failure of fuel pins occurred. From the point of view of safety, an important research aspect is the behaviour of mixed oxides under abnormal conditions, such as coolant losses, power transient, cyclic power changes, and interaction between sodium and molten fuel. The experiments concerning the safety (e.g. in HFR-Petten and Cabri-France) will certainly continue during the next ten years. The fuel cladding can be considered the most strained part of the fast-reactor core. In a temperature range of 350 C to about 700 °C it is subjected to internal mechanical loads by fission-gas pressures and fuel swelling, inside-corrosion attack by fission products, external corrosion due to sodium impurities, and a high radiation load. Suitable materials must fulfil a number of additional conditions, such as a small neutron absorption cross-section, good weldabi"ity, and good creep-rupture properties at high temperatures. The materials with the best combination of properties are the austenitic steel (SS 316, Incoloy), with Ni- and V-alloys as developable alternatives. Irradiation effects that have to be investigated experimentally are . swelling due to void formation by vacancy condensation. This is impor tant at neutron fluences nvt > 1022 n/cm2 and depends upon the irradi ation temperature, the neutron fluence, the material composition and the cold working during fabrication, . atomic displacements (are of importance at T > 600 C), . high-temperature embrittlement caused by the formation of helium bubbles after (n,a) reactions (important if nvt > 1022 n/cm2), . low-temperature embrittlement (already effective .: t a low dose of 1017 n/cm2 and at temperatures < 600 °C), . irradiation-induced creep. This occurs also at temperatures below 600 C and depends on swelling and stress. A schematic view of these phenomena is given in Fig. 9.1. Within the next few years much more data will become available for higher neu tron doses and for alternative structural and cladding materials. -51- h* rgfi. MO FlSVOft «O0*C S TtM»CM~ OuTVOt TUC EMWkTn.CMINT COMOVO* •» S«tU' MOT too-c I ClMMMMr EftAftlMC CKHiTUtMtW?. — IDHTOW MOUCED CtXtP ->r. th-e fist r^.».:'-.'r :"iel ~I 9.2 The fusion reactor It is highly probable that the material requirements for a fusion reactor are much more stringent than for a fission reactor. Beside the gereral structural materials the fusion reactor contains other materials sensitive to radiation, in the temperature range frosc 4 K to 1000 K, such as •- . liners with a low atomic number . electrical insulators, . liquid and/or solid materials for tritium breeding and neutron multi plication, . superconducting magnets, c.q. optical systems in laser applications. A survey of typical operating conditions *nó material candidates is given in table 9.1. The most critical part of the structure is the first wall, which is subjected to a bombardment of neutron*», 'ons and atoms, and to optical radiation from the plasma. The solid lithuT» compounds for tritium breeding (LiO , LiAlO,, etc.) pose about the same thermodynamic problems as the fissile materials in the fast breeder reactor, but besides those they introduce extra problems such as the influence of radiation on the -ri2- Table 9.1 Summary of Materials Requirements in the Nuclear Island Portion of D-T Fusion Reactors Approximate Maximum Typical Neutron Flux Primary Principal Operating (n/cm2*yr per MW/m2) Function Main Requirements Candidates Temp(°C) 14.1 Mev Total Structural High temperature strength Al alloys <300 Acceptable heat-transfer properties SS <650 Low macroscopic absorption cross-section Nickel-based alloys <700 1.4 • 102L 7 • 1021 Low induced radioactivity V alloys <800 Compatible with coolant Nb alloys <1000 Fabricable Mo and Mo <1000 Readily available, low cost alloys Some resistance to radiation damage Non-magnetic (except for laser reactors) Coolants High heat capacity He High thermal conductivity Li Low viscosity Na Depends Low pumping power (espe Li salts on 1.4 • 1021 7 • 1021 cially in magnetic fields) K structure Low macroscopic absorption cross-section Breeding High tritium production Li <1000 cross-section Li7Pb2 <300 Reasonable tritium extraction LiAl <500 1021 5 • 1021 Low T2 inventories LiAl02 <6O0-1500 Li02 <600-1500 Neutron High (n, 2n) cross-section Be <600 Multiplier Low parasitic absorption BeO, Be2C <15Q0 1021 5 - 1021 Pb <30ü Reflector High scattering cross- section C, steel Coolant 1020 5 • 1020 Low parasitic absorption temp. Shield High macroscopic absorp B 18 20 tion cross-section BltC 100-200 5 • 10 5 • 10 High gamma-ray attenuation Pb 2: 21 Electrical High dielectric strength Al203 500-800 1.4 • 10 7 • 10 Insulator High temperature stability (blanket) MgO Low temperature stability (magnets) Phenolics -269 3 • in13 3 - 1015 Resistant to radiation damage c Magnets High He, Tc, Jc (S/C ) NbTi, Nb3Sn, Va3Ga(S/C) Low resistivity (non S/C) Cu, Al -269 3 • 1013 3 • 1015 Strength at low temperature Ductile at low temperature Thermal Low thermal conductivity Mylar -269 to RT 3 • 1013 3 • 101' Insulator Good radiation damage (low temp.) resistance Optical High thermal stability CdSe Materials Good reflectivity (Mirrors) ZrC>2 , TiC>2 100-200 Depends (a n design (lasers) Resistance to radiation damage Al, Au, Cu a) Refers to all materials considered for this function; b) 100% plant factor; c) S/C = superconducting. -53- tritium diffusion. A severe problem which arises when the neutron multi plier Be is used, is its swelling due to the helium-producing (n,ot)- reactior.. The influence of irradiation, the radiation damage and its con sequences are the main subjects of material research, with an emphasis on the behaviour of first wall candidates. The irradiation conditions in a fusion reactor differ largely from those in a fission reactor: a) The neutron spectrum has a very important high-energetic component (> 1 MeV). Owing to tnis, the number of atomic displacements is larger, and other reactions like the (n,2n)~, (n,n'p)-, and (n,n'a)- reactions will play a role. In a fusion reactor spectrum, the cross- section of the (n,a)-reaction for Mo, Nb, V, and Al, averaged over the spectrum, is more than 100 times larger than the same cross- section in a fission reactor spectrum. The result is that the forma tion of helium and thus the swelling in the fusion reactor goes faster, see table 9.2. b) The first wall of the fusion reactor is bombarded by energetic charged particles, mainly D, T, and a. In a typical fusion reactor this flux is of the order of 1012 - 101" cm"2 s-1. c) The intermittant radiative load due to the cycle processes in the fusion reactor may have an impact on the resulting damage. Hardly anything is known about possible recovery mechanisms during the down time between two pulses. Heat cycles may also lead to fatigue. Table 9.2. Helium production in various materials (in appm/yr). fissio-. n a) fusioe • n b) Mo 2 47 Nb 1 24 V 0.3 57 Al 8 330 C 34 3000 SS-316 5 210 Be 3050 B4C 3600 LiA102 15500 a) EBR-II spectrum. b) UWMAK-I spectrum, wall lo.d 1 MW/m?. -54- The primary radiative effects in materials are atomic displace ments and nuclear transmutations. A survey of the resulting material damage, which is of dimensional, mechanical, and physical nature is qiven in table 9.3. It should be noted that important surface effects like sputtering, blistering, and evaporation have not been considered in report I. These effects are expected not to be alarming, provided the erosion rate can be limited to the order of um per year. Because of plasma contamination, however, the plasmaphysical conse quences may be important. Table 9.3 Radiation-induced effects. primary effect resulting damage to material swelling due to void formation blistering ^ sputtering surface erosion atomic I = evaporation^ displacement loss of ductility, embrittlement creep and fatigue change of physical properties change of physical properties nuclear swelling due to gas bubbles (He) transmutations activation afterheat Swelling . Helium and hydrogen, formed bv (n,a)~ and (n,p)-reactions, possibly in combination with vacancies caused by atomic displacements, give rise to void formation. The result is a swelling and embr.ttt lemen t of the material. Maximum swellinq occurs between 0.3 T and 0.5 T , wher^ ^ mm T is the melting point of the material in °K. Important reactor im plications are the limitation of the maximum permissible temperature of the material, the loss of vacuum integrity, a reduction of coolant flow, and the loss of mechanical strength. . Swelling of non-structural materials (e.g. Be, Li-compounds) could be an economic problem in that the reactor would have to be shut down periodically to replace swollen components. • Still insufficient knowledge is available concerning the permissible uniform and non-uniform swelling, as well as the influence of e.g. periodic tempering, transmutation products and cyclic loads on the void formation. Much more research in these broad fields is needed. -55- Ductility . Due tc the high radiation field, the high temperature, the presence of a magnetic field, the cyclic load, and the requirement of vacuum integrity, the conditions to be fulfilled by the structural materials of a fusion reactor are more stringent than those of a fission reactor. . The high He-production has a large influence on the deformability of the material. A requirement of e.g. 1% minimum uniform elongation limit will put a limit to the time of exposure of the material (for SS-316 this time is 2 years at a wall load of 2 MW/m2). . The plastic deformation by neutron irradiation is accelerated at high temperatures and stresses. This means a further limitation of the wall life and/or the permissible wall load. . Cyclic load can cause fatigue of the material, but one knows even less about its basic mechanism and the effect of irradiation on it than about creep. . As for swelling, much information is still lacking, and many expensive, long-lasting experiments will be necessary. For insulating materials there is a lack of data on changes in the electrical conductivity caused by high-energetic (up to 14 MeV) neutrons. This is also true for metals at high temperature. Many insulating mate rials contain oxygen, that has a large cross-section for the (n,a)- reaction. Hardly anything is known of the influence of a high He-concen tration and transmutation effects on dielectric properties. In superconducting magnets, the most radiation-sensitive materials are the organic super-insulators (such as mylar). Less sensitive are the stabilizers like Cu, and least sensitive is the superconduction material. A good design of the blanket and shield should provide for a low radiation damage. . From the point of view of the heat leakage, the blanket and the shield are already sufficiently thick, and the threshold for observ able effects in mylar is so high that problems may arise only locally (e.g. near gaps in the shield). The eCfects in mylar are irreversible, so that tempering is impossible. . The stabilization will not pose serious problems when copper is regu larly tempered at a high temperature that depends on the design and the permissible resistivity change. It is a cumbersome procedure since the heat capacity of the material is very large. . The effect of irradiation on superconducting properties (mainly -56- critical temperature and current) greatly depends on the irradiation temperature. However, data are lacking of experiments on irradiation and subsequent testing at T = 4 K. Consequences_of_radiation_dama2e The direct consequences of radiation-induced changes in the ma terial are mainly economical: . Limitation of the maximum permissible temperature of the blanket in order to suppress swelling and embrittlement by He-formation reduces the thermal efficiency. . Periodic replacement of irradiated parts leads to shut-down losses. . Additional facilities for remote-handling, storage, etc. increase capital and maintenance costs. Other non-negligible consequences are: . increases in the weight of radioactive waste which must be processed and stored (for the reference fusion reactor this is of the order of 400 tons per GW(e)yr). . Growing demand on scarce elements. 9.3 Conclusions The ultimate consequences of the limited lifetime of the reactor materials are an increase of the energy price and of the demand on the world reserves of scarce elements. The problems with materials in a fission breeder reactor have already been the subject of a long study. Long-term application of UO /PuO_ as the fuel and SS-316 as the cladding may not be adequate, so that alternative (carbidic) fuel compounds and nickel alloys will have to be investigated. Due to different irradiation conditions, data of irradiation ex periments for fast breeder reactors are not simply applicable to fusion reactors. The material problems for the fusion reactor are much more complicated than those for the fission reactor. There is a lack of experimental data, above all experimental facilities for irradiation research with intensive sources of high-energetic neutrons are not yet available. It must be concluded that, next to the plasma physics prob lems, radiation damage is the second most serious obstacle to the com mercialization of fusion power. The transition from technological to commercial feasibility will need an intensive long-term development. A large, fusion-directed, materials test reactor is indispensable. -57- 9.4 Consumption of materials While the choice of materials to be used in LMFBR's is rather definite now, it is too early to make a definite choice on all the ma terials in the field of fusion research. Research on consumption of ma terials is carried out mainly in view of long-range effects and the de mand on raw materials. £3 51 Jareeder _reactors Based on known designs, the consumption of materials for the con struction and the operation of an LMFBR is given in table 9.4. The numbers concern the nuclear part of the plant only (reactor core, vessel, primary coolant system) , the conventional part (turbines, generators, buildings, etc.) are excluded. The nuclear part contains about 9 t/MW(e) of essential materials, 4.5 t/MW(e) of which is used for construction, 4 t/MW(e) for coolant, and 0.1 t/MW(e) for fuel. The annual consumption is 0.03 t/MW(e) for fuel, 0.25 t/MW(e) for structural material and about 0.17 t/MW(e) for coolant. The inventory of the conventional part of the reactor is estimated to be 60 t/MW(e) (exclusive of the amount of more than 100 m3/MW(e) required for concrete structures). If the in ventory and the annual consumption are averaged over a lifetime of the plant of 30 years, the material requirements are 0.4 t/MW(e)yr. When we take into account the world reserves of raw materials, minable at present-day prices, each material has an upper limit of the total amount Table 9.4. Categories of consumption of materials. Inventory Consumption rate Classification Reactor materials " 2 cladding material 0.2 69 3 primary-circuit structural 2.8 119 steel (cannot be recycled until after at least 100 years storage) 4 secondary-circuit structural 1.6 66 steel (can be recycled after use) 5 sodium coolant 4.0 168 Total 8.7 454 -58- of energy that can be generated, expressed in MWyr. This is shown in table 9.5 for the most important materials (elements) in the breeder reactor. It is clearly seen from this table that instead of the fuel with its energy content of ab-ut 2*109 MW(e)yr, the amounts of Ni, Cr, and Mo form the limiting factor for energy production. It was assumed that apart from the fuel no reprocessing takes place. Only 10% of the structural materials belong to the category of materials that become radioactive in the core. The remaining part can be reprocessed and is then no longer a limiting factor.' Table 9.5 Summary of maximum possible Energy Generation and Materials Requirements for Fast Breeders under Resource Limitations. Maximum Energy qener- Maximum Materials World able from Resource Element Requirement Commitment Reserves Limitations 6 (t/MW(e)) (t/MW(e)*yr) (10 t) (MW(e)«yr) B 0.001 4 • 10_i' 20 4 -1Ü10 9 Cr 0.82 0.048 370 4 -10 ! ? Fe 2.9 0. 17 276,000 10 o « t n i ?. Mn 0.09 'J. 005 72,000 9 Mo 0.09 0.005 13 1.5-10 Ni 0.54 0.028 24 5 -10" Na 4.1 0.18 large large 13 Ti 0.0001 - 181 10 9 U 0.09 1.5 • 10~3 4 2 -10 EyËi9Q_Enactors The low power density in a fusion reactor (1-5 MW/m'), compared to 100 MW/m3 in the fast breeder reactor will have a direct bearing on the consumption of materials. Elaborate reactor studies for a tokamak yield an inventory of materials of 20-40 t/MW(e) for the nuclear part and 80-130 t/MW(e) for the total plant inclusive of buildings and the re processing of electrical energy, but not the concrete structures. The inventory and the annual consumption is greatly affected by the choice of blanket materials, in particular by the decision whether or not solid lithium compounds are applied. Due to parasitic absorption, these com pounds ask for a higher enrichment factor of the 6Li-component and the application of additional neutron multipliers, e.g. Be. Furthermore, -59- in that case liquid lithium could not be chosen as a coolant. The annual consumption of materials is determined by the integral first-wall load that is permissible and by the transmutation-induced burn-up. For vari ous reactor designs the estimated integrated wall loads vary from 2 to 10 MW yr/ra:, leading to a replacement of materials of 0.5 t/MW(e) yr. Table 9.6 gives a survey per element of the maximum inventory, the yearly consumption, the world reserves, and the producible amount of energy, based on a lifetime of 30 y<^ars. It has been assumed that all components that are replaced due to radiation damage, cannot be Table 9.6. World reserves and maximum onercy production of materials in a fusion reactor. r • ~ • ' ^—'— ~ 1 Maximum Energy gen- i Maximum Annual Element Consumption World erable from Resource Requirement Reserves Limitations (t/MW(e)) (t/MW(e)) (106 t) (MW(e)yr) Al 2.42 0.157 2000 8 * 11) ' Be 0.43 0.011 0.04 1.4 < JO6 B 3.61 0 20 1.3 * 10S '2 8.95 0.24 V "S. >> Cr 3.72 0.038 370 1.5 < 109 Nb 1.93 0. 193 cu 6.11 0.27 370 7 * UP r, He 0. 30 0.004 4.44 1 •: UP Fe 12.79 0.304 27b000 3 * 10i: Pb 13.90 - 210 4 x i03 Li (liquid) 1. 15 2*10~3 "1.3 1.4 "• 10' (solid) 0.35 0.147 I •<• If;" Mn 0.41 0.01 72 000 3 < 10I: Hg 0.02 - 0.2 5 3 >: 1(P Mo 2.82 0.060 13 7 < iO7 Ni 2.86 0.060 24 1. ) * 10i! _ Na 11. 99 y • Ti 1.24 - 181 4 •- U)'3 Sn 0.07 - i.l i.i y- to'1 Zr 0.O7 " 717 2.4 * 10' ' —.._ .. — — • .1 ,. „__, -60- recycled economically, with the exception of Be. The only consumption of Be is the burn-up of the atoms themselves and 2% loss during fabrication. The most critical material is Be if applied in the blanket, followed by Nb and Mo if applied as structural materials. Fusion reactors with Be as a neutron multiplier would allow an energy output of 1.5 x 106 MWyronly. Therefore one should look for another blanket composition, at least on the long run*. Obviously, fusion reactors may require four times more material per MW(e) in the nuclear island than fission breeder reactors. Integrated over a lifetime of 30 years the consumption of materials in the fusion reactor is again four times that in a fission breeder. The materials in ventory of the non-nuclear part has about two times the nuclear-island value, if the conventional parts of both types of reactor are considered identical, the total fusion power plant has a consumption of materials of 1.5 times that of the LMFBR. The finding of new resources and a possible tendency to extract ing materials from ores which art at present uneconomical to mine, have been left out of the considerations. The real resources known at present are much larger than the resources that can be mined at to-day's prices. On the other hand, it is impossible to use all reserves exclusively in reactors. The fraction of 3ach material, available for this purpose, is difficult to estimate. Since minerals are not scattered uniformly around the world, local shortages could cause significant problems on an international or regional level. ?2D£lusions The amount of raw materials consumed by the nuclear part of a fusion power plant during its lifetime is four times larger than that consumed by the nuclear part of an LMFBR. For the total plant this factor is 1.5. The only non-fuel elements that appear to be critical in the LMFBR deployment are Cr, Ni, and Mo; their maximum energy output is 109-1010 MWyr (the fuel reserves represent 2 * lO9 MWyr). The fuel re serves for the fusion reactor are essentially unlimited (see chapter 2). If Be would be used in the blanket, the energy output would be limited to 106 - 107 MWyr (this is about 2 to 20 times the annual electricity According to refe:ences cited in I the Be-reserves are 600,000 tons. The authors of I, however, use world reserves of 54,000 tons, on which the number 1.5*10r' MWyr is based. -61- consumption of the world at the present-day level). However, there is no doubt that the use of Be car. be avoided. Reliance on Kb or He limits the energy output to I> * 107 MW yr (100 times the present world consump tion). Li limits the output to 10" -10? MW yr. It will be clear from the foregoing that fusion reactor designs will fave to be optimized to min imum consumption of materials. 10. SUMMARY Fast breeder reactors and fusion reactors are suitable candidates for centralized, lonci-term energy production, their fuel reserves being practically unlimited. In the breeder reactor nuclei are split by neu trons, the fuel can have a moderate temperature, and the reaction system is relatively simple. Large-scale experiments have been carried out and demonstration reactors are operative. The complicated fuel cycle has been developed on a small scale, and a scaling-up is underway. A repro cessing facility could serve many fast breeder reactors. The fusion reactor opens a new unlimited fuel reserve. The fuel cycle is simple and it takes place near the reactor. The reaction prod uct is the inert helium. But the fusion reactor requires fusion of atomic nuclei and the Coulomb repulsion of these positively charged par ticles leads to an extremely high fuel temperature, a complicated reac tion system, and a relatively low power density. From experiments it is expected that the energy output will soon be larger than the energy needed for heating and confinement of the plasma (scientific feasibility). The technology of a durable and economical fusion reactor is still to be developed. Such a development parallel with the fast breeder is valuable by reasons of safety, proliferation, new fuel reserves, and by the very broad potential of the development of the fusion reactor. In order to facilitate a discussion of these aspects, the fusion reactor and the fist breeder reactor were compared in the IIASA-report. The UWMAK-I tokamak reactor design, taken in the comparison, underwent a series of improvements. Anions other things, the size and the needs for material have been decreased. A special choice of the structural mate rials could reduce activation even further. There remains a broad devel oping potential. The radioactivities of both reactor systems differ. In the fast breeder reactor the distribution of the activity over the fuel, the reaction products, and the structural material is determined by the fuel, and manifests itself in the fission products and the actinides. The radioactivity remains concentrated in the fuel elements, the struc tural material is much less activated. The fuel tritium in the fusion -62- reactor is, by its soft radiation and short biological and physical decay time, relatively harmless. In a fully operative fusion cycle, the amount of tritium circulating in the fuel and extraction systems is in the order of 1 kg. Besides, an amount of tritium is stored outside the reactor. The hard neutron spectrum and the large contribution of the neutrons to the primary energy lead to an activation of the materials in the reactor, which depends on the choice of materials and on the details of the construction. With conventional materials the biological hazard potential in air of the inventory of a fusion reactor becomes a factor 30 to 300 smaller than that of a fission breeder inventory. It should again be noted that the tokamak design, used in this comparison, is ca pable of improvements. The authorities that provide the licenses have to guarantee the safety of both systems. It seems that to this end the fusion reactor system will need less investments and personnel. The costs of plasma containment and heating, and the relatively low energy density could be compensated by lower fuel costs. In that case the energy price of fusion did not need to be much higher than that of fast breeders. We have made an attempt to put comparable aspects of both reactor systems side by side in the summarizing table. The numbers indicate how serious the problems with each of the aspects are. The meaning of these numbers is: 1 = favourable, 2 = less favourable, 3 = unfavourable. -63- — . ... r ASPECT FAST BREEDER REACTOR FUSION REACTOR } fuel practically unlimited 1 practically unlimited 1 scientific demonstrated 1 expected in 1985 1 feasibility technological practically demonstrated 1 uncertain 2 commercial expected in 1985-1995 1-2 uncertain 2-3 inventory 1.4*lOln Ci/GW(e) 2-13 <109 Ci/GW(e) ,. „. .„ BHPvair) 3xl0lc km3/GW(e) 108-10" km3/GW(e) radioactivity „ ,„, , 3 2 J BHP(water) 5.5*105 km3/GW(e) 0.1-6*10r' km3/GW(e) IBHP(water) SxiO13 km3s/GW(e) 9-2S0*1010 knrs/GW(e) after-heat 5 s, 2 1 reprocessing/waste reprocessing 24 t/yr 3 ~~ 60 m /yr (iow-level) 2 waste (high-level) 10 m3/yr 2 safety requirements high containment factor Pu 2 high containment factor T 2 (external) normal operation biological shield 1 biological shield 1 (internal) accidental hazards probability of serious 1-2 probability of lithium 1-2 accident fire with escape of tritium hazard potential 3 hazard potential 2 proliferation dependent of a.o. 2 comparable to fission 1 reprocessing procedure reactor only in case of hybrid subnational terrorism radiotoxic threat 2 unimportant threat 1 material properties knowledge and experience 1 research still required 2 available consumption of materials 0.4 t/MW(e)yr 1-2 1.5 t/MW(e)yr 2 reserves of structural practically unlimited 1 more stringent limita 2 materials tion than fuel costs high: 2000 $/kW(e) 2 very high: 3000-4000 3 reprocessing facility $AW(e) not included estimated energy price 0.09 $/kwh 2 0.09-0.12 $/kWh 2 -64- Fusion offers the possibility to select out of a broad spectrum of reactor concepts one that is scientifically, technologically, and commercially feasible. Fusion is then an option for the energy supply in the coming centuries. Fusion technology will partly benefit from the progress in fast breeder technology. With a well-developing fusion technology the stimulating influence can point both ways. REFERENCES 1. W. Hafele, J.P. Holdren, G. Kessler, G.L. Kulcinski, "Fusion and fast breeder reactors", RR-77-8 (1977), IIASA, Laxenburg, Austria. 2. SNR-300 documentation. 3. B. Badger et al., "UWMAK-I, A Wisconsin toroidal fusion reactor design", UWFDM-68 (1974) , University of Wisconsin. 4. R. Wilson, "Physics of LMFB reactor safety". Rev, Mod. Phys. 4_9 (1977) 893-924. 5. Nucl. News 2± (1978) 46-47. 6. Nucl. Eng. Int. 23 (1978) 43-45. 7. B. Badger et al., "UWMAK-III, A non-linear tokamak )ower reactor design", UWFDM-1S0 (1976), University of Wisconsin. 8. "Fusion reactor design concepts", Proc. of a Technical Committee Meeting and Workshop, Madison, 10-21 Oct. 1977. 9. E.T. Cheng et al., "Nucleonic design for a compact tokamak fusion reactor blanket and shield", UWFDM-256 (1978), University of Wisconsin. 10. J.W. Davis, G.L. Kulcinski, "Assessment of titanium for use in the first wall blanket structure of fusion power reactors", EPRI-ER-386 (1977). 11. G.L. Kulcinski et al., "Considerations of the recycle tine for radioactive titanium alloy structures". Trans. ANS 2J_ (1977) 79-80. This work was performed as part of the research programme cf the association agreement of Euratom and the "Stichting voor Fundamenteel Onderzoek der Materie" (FOM) with financial support from the "Neder landse Organisatie voor Zuiver-Wetenschappelijk Onderzoek" (ZWO) and Euratom. -65-