CiR-Benct Nr. 219 Eidg. institut fur Reaktorforschung Wirenlingen Schweiz Molten Chlorides Fast Breeder Reactor Problems and Possibilities M. Taube, J. Ligou By Wirenlingen, Juni 1972 We regret that some of the pages in the microfiche copy of this report may not be up to the proper legibility standards, even though the best possible copy was used for preparing the master fiche. EIR-Bericht Nr, 215 Molten Chlorides Fast Breeder Reactor Problems and Peszibilities M. Taube, J. Ligzou *) *) The authors would like to acknowledge particulary the valuable advice and assistence given by G, Markoczy (heat transfer problems) and G. Ullrich (corrosion problems) in the preparation of this report, June 1972 Conter.ts 11. 12, Sumrary RAsuvmé Gengral croblems General description of the reactor Reactor physies Tnerro-hydrealics Quogi- hinetic and pumping problems Fissiun products Repreccessing Aluminiun Tri-chloride as a working agent Corros.on Safety pr-vlems Econonmics Some other selected thermal, and fast, breeder reac.uors Appendix- Physical and chemical properties of the salt components Acknowledgements References page 59 62 70 71 Summary A fast breeder reactior of 2000 MW(t) oulput using molten chlorides as fuel and coolant is discussed. Some of the most significant characteristics are - the fuel contains only Pu013fN301, - the coolant is U013/N301 and also forms the fertile material along with the blanket, again UCl3/N301 - the fuel circulates through the core by forced convection. The thermal stability of the reactor is vary good. Power excursions cf Tuel temperature transients are quickly damped by the phenomensa of fuel expansion pushing part of the fissile material out of the critjcal zone, The balance of fission products including free chlorine seems to be stabiliz=d when some of the semi-noble metals (Ru, Rh, Pd and some Mo) are present in ithe elementary form, Corrcsion effects form the most difficult problem, Thermodynamic atud_e2s8 suggest the use of molybdenum alloys as structural mate- rials, Separation of the fuel éfifl;&@&%& from the fertile compo- nent UCl3 helps to overcome some of the corrosion »roblems, A reprocessing system btased on a salt-metal-transport process seems to be atiractive from the point of view of the economics of thne plant as a whole. The possiblity of using a dissociating gas as a secondary working agent in the turbine for example A12016 and N2q4is discussed. Résumé SURREGENERATE'R RAPIDE A SELS FONDUS (Chlorurss) Problemes et possibilités Un surrégénérateur r-pide de 2000 MW(t) utilisant des chlorures fondus comme combustible et comme réfrigérant est décrit. Quel- ques-unes de ses caractéristiques les plus intéressantes sont les suivantes: - le combustible liquide, contient seulement un mélange PuClB/NaCI - le réfrigérant est un mélange UCl3/N301; i1 comporte donc le matériau fertile tout comme les couvertures de méme composition - le combustible demeure dans le coeur mais doit é&étre entrainé suivant une boucle de convection forcée, La stabilité thermique du réacteur est trés bonne; les excursions de puissance et les transitoires concernant la température du combustible sont fortement limitée par *'expansion du combustible qui se trouve déplacé vers les régions périphériques qui con- tribuent beaucoup moins & la réactivité. L'équilibre des produits de fission incluant des atomes de chlore semble &tre réalisé quand certains métaux semi-noble (Ru, Rh, Pd et Mo) sont présents, Les effets de corrosion constituent le probldme le plus difficile. Des études de thermodynamique suggdrent l'utilisation d'alliages de molybdéne comme matériau de structure. Un systéme de régénération basé sur un procédé de "transport sel- métal” parait intéressant du point de vue économique. Enfin la possibilité d'utiliser les propriétés de dimérisation de certains gaz, par exemple Alzcl6 et N204, dans le cycle thermc- 1, General Problems The biggest attraction of breeder reactors is their ability to utilize directly or indirectly, the non-fissile nuclides U-238 - in fast breeder reactors, and Th-232 - in thermal reactors. The relative advantages of the fast breeder over the thermal bree- der are: higher breeding ratio, higher specific power which re- sults in the shortening of the doubling time as well as in the more intensive use of the reactor volume. In addition it must be emphasized, that each breeder reactor, which is built to last 25-30 years, must be considered not only as a power producing device but also as a source of fissile material. Therefore each breeder reactor should be considered as part of a long term complex ' breeding system ' which includes both the power reactor and the fuel reprocessing plant over this period. From this point of view the reactors with molten fuel are better adapted to the long term ' breeding system ' than are the solid fuel reactors. (simpler reprocessing technology, minimal transpor- tation problems, smaller environmental danger, better economics(?)). (Fig. 1) Such a coupled ’breeder system’ has additional advantages when the reprocessing technology is based on high temperature processes, such as pyrometallurgical or pyrochemical techniques instead of low temperature processes in aqueous solutions. These high tempe- rature reactorkeproceseing systems might be realised in the most favorable manner when the fuel in the breeders is in a molten liquid and not in a solid state. Thus in the future, 'breeding systems’'using molten fuel fast reactors seem to be of interest. The molten fuel fast reactors can lLbe classified in the following manner 1) Molten metallic fuel 2) Molten salt fuel Past reactors with molten metallic plutonium fuels were constructed in the sixties in Los Alamos under the name LAMPRE. The molten alloy Pu-Co-Ce was the most promising fuel proposed. The results were encouraging but further experimehts have not been realised. These reactor types were not breeders. Fast reactors with molten salt fuel exist only as paper studies at the present time. Som: work has been done at QOak Ridge (1956), Warsav (1960-68) Argonne (1965-68) and Harwell (1963-70), In addi- tion some experimental work has also been carried out, When nuclear, physical and chemical considerations are studied it can be shown that the only possible fuel conatituent of these fasi reactors are the molten chlorides. In the case of thermal reactors the most suitable molten salt fuel proves to be a fluoride. (Fluorides are moderators, and therefore could not be used in a fast reactor, because of the dramatic softening of the neutron spectrum). The search for the best components of molten salt fuel for fast reactors must take into account not only thermal and hydraulic properties but also the following nuclear properties: elastic and inelastic cross sections, cross sections for neutron absorbing mechanisms, not only (n,y) but also (r,p), (n,x) and (n,2n). The chlorides of uranium~238 and plutonium-239 diluted by sodium chlo- ride are the selected components of the fused salt fuel. (Taube 1961) Molten fuel reactors differ from the peint of view of thne cooling system. ‘The following are three types of molten fuel reactors: a) Externally cooled, where the molten fuel is pumped out of the core to the external heat exchanger. In this type of reactor, cnly fucl and fertile material are present in the core {no coclant)., The large ruount of molten fuel ouvtside the core does not of course contribute to the critical mass, This type of reactor has been discussed for example by Nelison, (Argonne 1967) and Lane (USA 1970) especially as a high flux material testing fast reactor. In externally cooled fast reactors the loss of a portion of the delayed neutrons could adversly affect reactor control. Also the biological shielding outside the core is very expensive. b) Internally, direct cooled reactors: here the cooling agent is pumped directly into the core where, after mixing the fuel in the lower part of the core is separated and pumped out of the core to the heat exchanger. The direct contact of molten fuel with molten coolant has several particular advantages., Very good heat transfer, no coolant tubes {(or cladding), possi- bility of transporting fission products. The disadvantages are unfortunately, also numerous: problems of mixing and separating the fuel and coolant, corrosion etc. This type of reactor has also been studied eg. - cooled by molten lead (Long, Harwell and Killingback, Winfrith), cooled by boiling mercury (Taube, Warsaw) and cooled by boiling aluminium chloride (Taube, Warsaw). This type of reactor must be considered ag sn 'extreme exotic type'. P g FAST a REACTOR " WITH _ SOLID o FUEL 2 & IRRADIATED FUEL COOLING TRANSPORTATION =!| DECLADDING T | = v w| LIQUIFACTION = 7 v wl SEPARATION 3 L PRocfssss TRANSPO*RTATiON 2| PREPARATION < |OF FUEL MATERIAL ; MANUFACTURE S| OF SOLID FUEL S & CLADDING o 0 FUEL PIN | ASSEMBLY .| n l T Iy TRANSPORTATION FAST REACTOR WITH MOLTEN FUEL CONTINUOUS SEPARATION OF IRRADIATED FUEL POWER STATION WITH FUEL CYCLING Ltk FUEL PROCESSING CYCLE FOR FAST BREEDER REACTORS WITH SOLID AND MOLTEN FUEL ¢) Internallv jndirectly cooled reactor: here the cooling agent flows through tubes in the core. Heat is transferred from fuel to c¢oolant across the tubes. No direct contact between molten fuel and 1iquid or gaseous coolant is permitted. These types have also been studied, in mos. cases using sodium as a coolant, (Nelson, Argonne 1967). In this paper an internally 3‘;{;;%5'@51, cooled molten salt fast reactor is discussed. The unusual difference is in the use of a molten chloride coolant, including uranium chloride in place of sodium or gaseous coolant. The uraniunm chloride component is in fact the fertile constituent which doubles as coolant. Fig. 1 shows the flow diagrams o two types of reactors: with solid fuel and separate reprocessing plant and the molten fuel reactor with integral reprocessing system. 2, General description of the reactor In this paper a molten chlorides fast breeder reactor is discussed. The most important features of this reactor are: (Fig. 2 and Pig. 3) - thermal power 2050 MW(th) - 1936 MW(t) in core + 114 MW(t) in blanket giving - electrical power 1000 MW(e) (in the inost optimistic case) - molten fuel consisting of (in mol%) 15% PuCl; (of which Pu-239 + Pu-241 = 80% and Pu-240 = 20%) 8% NaCl (no 238UCl in fuel) 3 and fission products in the form of chlorides or in elementary state, molten fertile material (in mol %) 65% “>uca, 365% NaCl and newly bred PuCl3 and fission products. coolant flowing in tubes: the same as fertile material (no other coolant in core) blanket material: the same as fertile material the core is internally cooled, there is no circulating fuel outside the core. the fuel and the coolant are flowing in the same direction (see fig. 3) the reprocessing plant is in intimate proximity witn the reactor (under the same roof) for the sake of obtaining relatively high plant thermal effi- ciency a rather exotic working agent is proposed: aluminium trichloride. This agent removes the heat from the primary coolant molen U013/N301) in an external heat exchanger, for full use of the thermal energy a secondary working agent is also proposed: nitrogen dioxide. The theoretical thermal efficiency of such a power station may be significantly higher than those of a 'classical' power stavion. Of course a more conservative steam turbine system need not be excluded, but it is possible 1o argue that one attraction of such system would be to use for 'district' heating, the fuel in the core and the coolant are pumped with the velo- city of 2 and 9 m3™L respectively. FUEL velocity 2m§) FERTILE MATERIAL REACTOR ~2050MW() Pu gain 1000MWie} OUTPU AIR COOLING oriong distance heating system —» 1050MW(y) 02 sink temperature is ~100C) | o, A A1 TEALIL MMt ARIESE BEAASYTAES Y AMT 10 FUEL REPROCESSING Y COOLANT UCt./Nafm PLANT = HEAT EXCHANGER OVERFLOW tor FUEL | R - oy . T TeetwwNNNN | - Y N NN (SN DAY N - /r/ EX PANSION ] BLANKET WITH MOLTEN FERTILE MATERIAL UCl3/Na Cl HEAT EXCHANGER im 6 T& 5 — No,of Tubes 19940 'l_j ) [COOLANT UCls/NaCt HEAT EXCHANGER ;l_‘ FOR BLANKET l R U Fig3 CORE AND BLANKET SCHEMATIC 11 - the possible structural material: molybdenum alloy with small amounts of other metals e.g. Ni, Fe +.uss The advantages of the proposed reactor are the following: - no separate coolant no 'foreign' cooling agent {(e.g. sodium, helium ete.) in the core which results in a more satisfactory sysliem with improved neutron balance. ~ the fuel inventory is very small due to lack of a separate cooling system and because of the small ~ut-of-core inventory on the basis of the direct coupled continous reprocessing plant. - the fuel contains only plutonium and no uranium which simpli- fies the processing technology and removes the danger of uranium trichloride oxidation which also improves the corrosion proper— ties of this medium. -~ the corrosion problems are casier to solve when the aggressive turbine working agent, aluminium chloride is situated outside the core. - the high velocities of both fuel and coolant significantly reduces the temperature gradients at the equilibrium state and reduces the mass transport mechanism, The mass transport mechanism is very sensitive to temperature gradients and plays a large part in corrosion mechanisms, However the disadvantages are numerous: -~ the first and most important disadvantage is of course corro- sion. The molten cnloride medium, especially in neutron and gamma fields, at high temperatures and velocities with chlorine 12 being virtually free in the fission process of plutonium chloride presents a very serious problem which must, and pro- bably could be solved, - the most likely structural material seems to be molybdenum alloy which among other things gives rise tc¢ parasytic absorp- tion of neutrons, - the fuel is circulated by a pump which must be located in or close to the core which increases the corrosion problems. - the high fuel and coolant velocities result in high pumping costs and couwld cause severe erosion. Table I Molten chloride fast breeder reactor 'CHLOROPHIL' Blectrical power (approximate estimate) MW(e) 1000 Thermal power, total MW(t) 2050 - in core MW(t) 1936 - in blanket (approx) MW(t) 114 Core volume m? 7.62 Specific power (core) M~ 255 Core structure see fig. 3 Plant efficiency {estimate) % 49* Liguid fuel properties PuGl3 mol % 15 NaCl mol % 85 Liquidus point % 685 Boiling point (at 1 bar) °¢ 1500 Fuel temperature (mean) ¢ 984 Puel volume in core m° 2,66 15 Density at 984°C kflme Keat capacity (for 984°C) KJ.kg’ldeg' Viscosity (984°C) g .om ts~t Thermal conductivity (for 750°C) H.cm-ldeg- Fuel in core kg Total plutonium in core kg Total plutonium in salt weight % Total plutonium in entire system kg Mean plutonium specific power (core) MW(t).kg * Plutonium specific power (entire system)flfift).kgfl Coolant Prgyerties - . s el o . i Al (U,54)C1g (depleted uranium) mol % weight % NaCl mol % Uyzg in salt : % Liquidus point C Boiling point °c Coolant temp. inlet °¢a Coolant temp outlet ¢ Coolant fertile salt in core m3 density kg.mf3 in core kg Corlent fertile salt total (blanket + core) n° kg Uranium inventory (blanket + core) K& Uranium inventory reactor + kg reprocessing Molybdenum a2lloy (80% Mo) in core kg Core geometry - cylindrical height diemeter m volume m3 2340 0.95% 0.,0217 0.007 6210 2252 36 .4 2500 0.86 0.775 65 92 35 63.5 710 1700 750 795 4.96 4000 20000 46.14 185000 171000 180000 3000 2,00 2.20 7.62 Axial blanket height Radial blanket thicknesses Blanket + in core coolant (tubes) volume Thermo-hydraulics Fuel velocity Fuel pump Fuel Coolant velocity Coolant pump Number of coolant tubes Tubes - inner dia - outer dia - tube material - tube pitch piteh/inner dia ratio Secondary working agent Reprocessing Efficiency for Pu + F.,P., separation (assumed) Fuel stream to plant Mean cycle time for fuel Fertile stream to plant Mean cycle time for fertile medium Breeding ratio Internal (in core only)** Quter (in blanket) total Doubling time * Total station efficiency only roughly estimated cm cm cm years 0.80 1.00 38.25 2.00 in core shell side 9.00 in external heat exchan- ger 19941 1,20 1.26 Mo. alloy 1.58 1.15 Al1C1. gasegus continuous 50 0,003 21 0.216 56 0.709 0,680 1.389 9.2 ** Neutron calculations neglecting parasitic absorbtion by vessel atriintivo 15 Reactor Physics The neutronics calculations have been made in 3 steps: - rough calculation with one group cross secticns taken from sodium cooled oxide fuel fast reactor data. - calculation with 15 group cross sections for given chloride conposition. - recalculation with newly calculated one group cross sections, normalized from spectrum obtained in the previous 15 group calculation, (see fig. 4) The one group cross sections used in these calculations are given in table 2, The neutron balance is given in table 3, There are references in the literature to the adverse effect of neutron absorption by the chlorine isotopes. For fast neutrons the two isotopes of chlorine have the following cross—sections. % isotope fast neutron value Nuclide in natural C1 c¢ross sections (barns) 01"35 75-53 U(n!p) 0.072 o(n,a) C1-37 24,47 o(n,p) oln,a) 0.0015 From the data obtained the adverse influence of chlorine-35 is rather small and an isotopically enriched chlorine is not required. Such a suggestion was made by Weinberg and Wigner (1955) but was hflflfla pile. mmile .fifi“1 ‘A“ il . el e “Aflfl‘ _‘ fl*‘fl flflflflflflflfl *4 e Maalle "fifi - e 16 Table 2 One group cross-sections for the molten chloride fast reactor on the basis of 15 group data of Bondarenko. Nuclide O¢ Gcapture ‘r Pu-239 (and Pu-241) 1,826 0,256 2,962 Pu-240 0,546 0,437 2,660 (estimated) U-238 0,0743 0,3145 2,660 Na - 0,0013%3% - Cl - 0,015 - Mo - 0,0743 -~ Fission products - 0,273 - Table 3 Neutron balance for 1 cm height of cell (;,21 cm32 20 Nuclide x10 "~ Neutrons produced Neutrons sbsorbed Pu~239 11.36 (n,f) 61.4 20,7 (n,Y) - 2.9 Pu-240 2,85 (n,f) 4.14 1.55 (n,v) - 1.24 U-238 72,39 (n,f) 14.3 5,40 (n,y) - 15,53 Na 104,00 - 0.13 Cl 375.00 - 5.62 Mo 56 .00 - 4,2 F.P. 1.13 - 0.3 79.8 57.57 = 12:8. . 3,387 57457 17 To a similar extent the problem of the exact choice of construc- tional mnatexrials for the coolant tubes will also affect the neutronics calculations. At least two criteria must be considered in the selection of tube material: - the absorption cross-section for fast neutrons (agD.l MeV) because of its effect on the breeding ratio. (fig. 5) - chemical stability against the attack of chlorine ions which can be partly represented by the free enthalpy of chloride for- 1000 QK) mation (Gform Fig. 5 shows both properties of some selected materials. From this it appears that one of the most suitable metals is molybdenum. Un- fortunately this answer cannot be regarded as completely satisfac- tory because of the lack of real experirmental data and because of some thermodynamic questions which are discussed in sections 8 and 9. 4. Thermohydraulics The calculations for this type of reactor have been based on the following more or less arbitrary selected parameters: - fuel in shell side, with tube pitch to diameter ratio equal to 1.10 to 1.18, - fuel velocity: 0.5 to 5 m.s'l - core dia: 2 and 2.2 n - core height: 2 m bt ) ) wn i Capture Cross section [barng 0-06+ Q@ r ] Relative Flux Elrbiturr Units] %3/\/ 18 NEUTRON ENERGY, MeV 1] - - L ) o - w 88 888383~ vy ®y oo wl! T 2 2B @ @0 O 00~ ¥ 7 2 . 1 S 0-5- l Spectrum in this Reactor — 024 01 - oosd — ~.Fission Spectrum 002 | 00t 1§ "1 '%3'12 N D'9°'8'7 "6 8§ 4 3 21 [Groups) Figé NEUTRON SPECTRA Preferred Metals Co ® Fe Forbidden vV Ti ® @ Metals Zr ® 1000K ® 20 40 60 80 D0 120 %0 160 180 200 220 AG, Fig5 CHOICE OF STRUCTURAL MATERIALS ] 19 - coolant in tube with tube internal diameter equal to: 1.0 to 1.5 cm -~ velocity of coolant: 1 to 17 me T - coolant inlet temperature ?500 and 800°C The calculation of neutronics and thermo-hydraulics were made for 1 cm of the core height (see fig. 6 - flowsheet of program) The data given in table 1 obtained from these caiculations are for a steady state reactor. The detailed representation for the temperature distribution in a typical power reactor with a core output of 1936 MW(t) are given in fig. 7 (for a position 43 cm above the bottom of the core where the neutron flux is normalised to 1). The bulk temperature of the fuel is here 99800, the temperature of the tube walls 857 - 839°C and the bulk temperature of the coolant 781°C. For the total output of the core 1936 MW(t), the power distri- bution is as shown in fig. 8. Of course a flatter power distribution could be obtained by ad justing tube diameters and pitch across the core. (Note that in this calculation the radial neutron flux distribu- tion has been taken as unperturbed). A very encouraging indication of the good temperature distribution with very small temperature gradients is skown in fig, 9 which indicates the axial bulk temperature distribution in the fuel and in the coolant in the core, X = height of core x no. of passes [cm) D =density of component C =heat capacity V =viscosity K =heat conductivity G =geometry of channels ¢)'neutron flux arbitrary values for x=0 from previous calculat) G, Uy, Ug, Ny, N, T output 20 E=energy produktion H =heat transter Ti=temperature of fuel TC=inlet temperature of coolant U =velocities of fuet and coolant Nf-fuel composition Nc-coolant composition Re-Re)notds number Pr=Prandtl number Nu-Nusselt number v ' — ¢-' (x) ‘ Yo X1 —® E =1(4,0.1c) ) > Re=tx1...) " Pretar.) — Nusf (xT.,.) | Y x«200,40 ILOOcn?,d pasos’ > —r T flx) ————pm —» D=tr,T) — C=tt,0,1) ™ Ve=tx..1n ™ K=f{x.1) ' | X=2400 corresponds to 6passes of fuel trough core 2l TEMPERATURE °C COOP.ANT : 0-3mm 12!mm ‘r £ Fuel \ (S - %) Zms 4 / ; 1000 338C g COO|LANT 7 /I U j 9mls’ 2 / % /] j 900°- /! / % fl 854°C / 4 / 800" % % ; l7e1°C j | o — ; total 24455 | 4 Heat Transfer [wem™a 750 0-0216 /] 0-0937 ] Viscosity 0:494 /// 0957 )g Heat Capacity [J’grfi‘deg' 200 175 22 Mean total value 1936-27 MW(t) 150 125 M —f Q o ) un Core Height [C N o 25 3 4 5 6 7 8 9 10 W 12 Power produced per cm height of cell MW(t) Fig 8 Power distribution in the core 260 D . 175 150 - il N o 100 CORE HEIGHT [CM| ~3 o Il 50 - \ 25 - + \ 23 P-1936MW(t) P:20x10"herids™ A Coolant 9ms™ Fuel 2ms \ 0 2 i ' .\ 3 4 P i A A 2 1 2 | 750 60 70 80 90 800 % 900 910 920 940 950 960 970 980 899G 100 TEMPERAT' RE [ C Fig 9 TEMPERATURE of FUEL and COOLANT 24 The fuel bulk temperature changes form 980°C at the bottom to 96500 at 1/4 core height and is 998°C at 3/4 of core height. The coolant temperature lies betweaen 750°C inlet and 79300 outlet. Both these small temperature gradients in the fuel and in the coo- lant (fertile and blanket material) may prove beneficial in redu~- cing corrosion processes due to the minimizing of mass transport phenomena, The stable behaviour of this type of reactor results from many parameters. Two of them are the velocity of the coolant and its bulk temperature. The mean power output of the core is strongly dependant on the velocities of both fuel and coolant. (fig. 10) For a fuel velocity of 2 m.s-l, when the coolant velocity falls from 12 m.s-l to 1 m.s'l the coolant outlet temperature increases from 784°C to 893°C for constant inlet temperature of 750°C. This change of coolant velocity and its bulk temperature results in the decrease of the mean core output from 2088 MW(t) to 598 MW(t) - that is approcimatly a factor 3! It is clear since the lower coolant velocity results in a higher coolant outlet temperature and lower power outi.. .2 have definite negative temperature coefficient (power output) varying with the given coolant velocity. 1 If the fuel velocity falls from 2 m.s - to 0.8 m.s-l we again get an important decre se of power output (see fig. 11) The decrease in both fuel and coolant velocity results in a sharp decrease of reactor power (see fig. 12). This means that such a reactor can be considered as a surprisingly stable and self regula- ting device, In the case of a sudden fall in coolant and/or fuel velocities the power output decreases to a safe level without inter- vention. 200 - 1938MW(:L PR - ) -~ | TR08°C " i%93°C 150 - CORE - POWER | OUTPUT ~7 il — MW() - ’ i ' A o> 1000 - { 2 3 4 5 6 7 8 9 10 1 12 coolant velocity [ms"] Fig 10 CORE POWER OUTPUT YEBSUS COOLANT =» = R & . i 26 2000 ~ A ’/' ] | \“'\fy/" S K 7 A & 7 y 1500- ‘//’ o< CORE / POWER 4 MW (8 Y 1000 - i L ims” 4 500 - 0 T T T 0.8 | ’ Y ! ] 2-0 FUEL VELOCITY ms™! Fig 11 CORE POWER OUTPUT VERSUS FUEL VELOCITY 27 2000+ i : E Coolant velocity [MS' ———@ Fuel velocity [HS'fl 1500 Core 7 Power output MW(t) \ - A \ 1000- \-\\. \ ™~ \ N M—\ 500- \é) - ] 0 T T T 1 ) ¥ ] T | ¥ | R | T 1 T ¥ 995 1000 1005 1010 Bulk fuel temperature [°C] Fig 12 Power output versus bulk fuel temperature 28 The reference core with 1936 MW(t) coolant velocity 9 m.s-l, fuel velocity 2 m.s'l, tube pitch/dia ratio 1.1%, has a plutonium in- ventory of 2252 kg of Pu (Pu-239 + Pu-240) which gives a specific power of 0,86 MW(t)/kg Pu. The breeding ratio in the core for the reference case is calcula- ted as The blanket breeding ratio for a 1 m blanket is calculated as approx 0.680 which gives a total breeding ratio 1.389. The doubling time for this reactor is given the following defi- nition (at 80 % load factor) Pu inventory in reactor Doubling time = Pu gain 2252 kg Pu 1936 MW(t) x 1.1 kg Pu/1000 Mwg ¥ 300 days/year x (1.389-1) 9,7 years It must be stressed that this doubling time is a linear one. 29 5, Quasi-kinetic and pumping problems The achievment of the required fuel velocity in the core seems to require a forced circulation system since the rough estimate using notural convection gives a heat transfer coefficient which is too low. Such a forced circulation system (core only) can be one of the following types - pump installed directly in core - pump outside the core - an external pump with injector - a gas-1ift pump using inertgas (argon) Intensive consideration of the factors involved, using criteria such as ~ reduction of the out of core inventory, elimination of additional heat exchangers, minimization of the fuel leakage, mini- mization of the auxilliary opwer, optimisation of the fuel flow regulation, points to an in-core pump solution, Of course this gives rise to considerable technical problems (cooling of the rotor, corrosion and erosion, maintenance, neutron activation ete.) ible The postulated fuel velocity makes it gsfiaible to mal.e some calcu- lation on the heat transfer problems and also gives a feel for the kinetics of the reactor under discussion. It must be stressed that these kinetics studies have no strong physical sense and use an ifl%fi%gEZhe approach but it is clear that they give some useful information about the general reactor stabi- lity. (see fig. 13 and fig. 14) There is little information on the density of PuClB/N301 and U013/Na01 in the temperature range of interest. For these caleunla- I — I - e | - ] P I [ [ I N I o =Ly PR . e Y ™y ™ ™ Fuel velocity 2ms™ -1 1200~ Ccolant Ims - ""'\‘ - ’I \\ Neutron flux step change -16!10' i\ .‘ 1100 | {1 | . I ! = |/ O “10004/ . \ \A/\/\ = . \ [ ; | E | " Reference case ® ‘ $=0-8x10"nem s~ _g ! T;uel'ggo.c “900 - ,’ -KCold fuel case 8 00- I $=0-8x10%ncm s 10 =750°C { fuel 750| I I | | I I ] | L ! ' ' ! ' § | ' L ' | J 0 1 2 3 4 5 Botton Top Botton Fuel No. of passes of core Fig 13 Fuel temperature during several passes through core under different inital conditions 31 1200-: "50_f | Fuel velocizy 2ms™ ] Tryey= 990°C . $= 0-8x10*ncms™! Coolant 11004 velocity . ims=-! 105 0- Reference case Coolant velocity 9ms-’ d :f‘% \ 1 Ly N o \l ;’! “; IR ,o‘\ d1 5 5 i 1 N YRR N VA A A A AT ARV W I AN A N A N N AN :!\_g ’--lf.‘,l‘,,"ll!“f‘;‘ VI \,‘.' | | “U! ‘J b ‘\ i ‘l ;' W “ ! | cootant” * . ¥ VoV vy 4 Coolan velocity 12ms™ 950 ' ’ | J i ¥ ' | T J | I T 1 0 1 2 3 4 5 No. of Passes 32 given in Tigure 165. The influence of the temperature coefficient of deusity (Ap/p per deg) is calculated here for 3 cases, 0.5 x 10 =3 1x 10 =2 ane 1.5 x 10 3. Fig. 16 gives the results which show only a small influence on reactor power. 6, Fission Products The fission process of plutonium in a molten chloride medium may be expressed as PuCl:_.’ (n,f) ~— Fiss, products + 3 Cl From the earlier published data of Taube 1965 and Chasanov 1965 and the more recent data of Lang 1970 it appears that the problem of the pessible existence of fission product oxides states was not sufficiently discussed. On t'.e basis of the fission yields of Pu-239 in a fast reactor after 10 days irradiation and without cooling time the yield of the indi- vidual elements are as shown in table 4 The most difficult problem is to predict the probable valency (oxidation state) of some metallic elements and metalloids., In this paper the oxidation states have been determined on the basis of the free enthalpy of formation of the chlorides (according to Veryatin et al. 1965) and are shown below, 33 TOP 200 232 233 234235 ?s 398 339 400401 402 980°C 150 998°C Height of core - [em] Fuel 100 | T=8847C 2ms™ Power Output ! _ 1939 MW(t) $'-" 0*8:(10" m:rfi’s!s"1 50 - 967°C 25 - ' 750 BOTTOM 0 r— I ,m:c T ¥ ] ) ! ! 2:34 235 236 398 399400401 402 DENSITY OF SALY Fig 15 ger) Density of fuel and coolant for _— e Reterence Core + 2000 - i Fuel velocity 2ms™ - Coolant velocity 9ms™ 1500 - Power | MW (t) 1000 - i ‘\+_ Fuel velocity 0-8ms-’ 500- ~® Coolant velocity 1ms"! 0 | ] 1 ) 1 I i ' L 1 L ] ] i 0-5x10 +0x10™> 15x10 Temperature coefficient of density _5'2“9" Fig 16 Impact of the density temperature coefficient of the core power outnut 35 Table 4 Fission products in irradiated chloride fuel Oxidation state Fission products 0 Kr, Xe 0 Se, Br, I 0 Tc, Ru, Rh, Pd 0 Mo (partially 30 %) + 1 Rb, Cs, Ag + 1 In + 2 or, Cd, Ba + 2 Sn, Mo {70 %) + 3 Y, La, Ce, Pr, N4, Pm, Dec + 3 Sm, Eu + 3 Sb + 5 Nb + 3 Zr Remarks noble gases volatile metalloids semi-noble metals the only stable oxida- tion state the most stable (but unimportant) the only stable oxida- tion state the most probable oxi- dation state in this med ium the only stable state the most probable oxi- dation state the lowest oxidation atate the only stable oxida- tion state the most probable state: On the basis of this assumption the balance of chlorine is fully realised: no free chlorine is to be expected in this molten chlo- ride system., The following elements are fully or partially in a nonoxidized state, that is in metallic form: Tc, Ru, Rh, Pd 36 The total amount of these metals equals approximately 50 atoms per 100 fissioned atoms of Pu-239 (see also fig. 17). (Note that all these considerations have been made for standard free enthalpy; but even a change in the thermodynamic activity from Yy =1 to vy = 0,001 which means a change in free enthalpy of 14 kJ-mol-l, thus appears insignificant in these rough considera- tions). Table 5 Fission products in a molten chloride plutonium fuel after 10 days irradiation in a fast reactor !no cooling}= Yields per 100 atoms of Pu-239 fissioned. (Burris, 1957) Element Yield Oxidation state Chloride Se 0,008 0 no Br 0.003 0 no Kr 0,942 0 no Rb 1,050 + 1 RbC1 Sr 5,487 + 2 SrCl2 Y 3,028 + 3 YCl3 Zr 21,520 + 3 ZrCl3 Nb 0,289 + 5 Hb015 Mo 18,160 + 2 MoCl,, Tc 4,014 0 no Ru 31,445 0 no Rh 1,736 0 no Pd 12,657 0 no Ag 1,88 + 1 AgCl Cd 0,66 ¥ 2 CdCl2 37 Element Yield Oxidation state Chloride In 0,00 + 1 InCl Sn 0,324 ¥ 2 SnCl3 Sb 0,674 * 3 Sb013 Te 7,654 + 2 Te(‘l2 I 6,177 0 no Xe 21,234 0 no Cs 15,355 +1 03012 Ba 9,502 + 2 BaCl2 La 5,79 + 3 LaCl3 Ce 13,986 + 3 CeCl, Pr 4,278 + 3 PrCl3 Nd 11,870 + 3 NdCl3 Pm 1,44 + 3 PmCl3 Sm 3,737 + 3 SmCl3 Eu 0,595 + 3 EuCl3 Gd 0,028 :_Z EEEEE Total 200 Mean +15 200 M Cll.5 Fission products in fertile material The most important reactions in the fertile material are fission process UCl3 -» Fiss, products 1;58 g% , oxidation process Um3 + 1/2012--0014;49 = 25 kJemol disproportionation UCl3 + 3 U013-P3U01 st Umet (see alsotHar- der, 1970) Because some of the fission product chlorides have a free enthalpy of formation of the same order of magnitude as the oxidation 1300 - 1250 - 1200 1 1150 - 1100 - Temperature °C o ot [ =] . | 900+ 850 800 38 Remark: Fig. 25 is not in the right place, it should be on the page 53 To 990°C Pox 2000 MW(t) ¢° = 0-8x10% Fuel Temperature Fuel velocity=0-§ ms"' Power S 540MW(t) Power = 1936 MW(t) ower wl, 750 No.of Passes Selfdamping mechanisms of the reactor tallawina cten rediwctinne of canlant and 39 process of U013--U014, a reaction of the following type is possible UC1 + 1/2 MoCl uc1 +1/2 M : 3(s) e Ae) "(m) The corrosion aggressivity of UCJ_4 is of course similar to that of HoClz. 7. Rep ‘cessing Breeder reactors as we already know form part of a 'breeder system' which includes not only the power reactor but also the reprocessing plant. The advantages of molten salt breeder reactors become particularly apparent when the reprocessing plant is under the same roof as the power reactor and when chemical separation processes take place in the high temperature molten salt media in a continuous cycle. The separation of plutonium and/or uranium from the irradiated fuel by means of pyrochemical techniques could be carried out, for example, in the following way Molten salt, primary phase Pu, FP (part of FP remains) Transport of Pu and part of FP, Metallic phase (part of FP remains) Transport of Pu Molten salt, secondary phase containing only FPu. This is the so called 'metal transport’ process (fig. 18) 40 0—-1(6+17)—Xe(212)— Br—Kr(0:84)-Rh(1.73)-Ru3144) Te - — 1€Cls (4-01) o PdCl2(12-65) Cm— o e=—=2 o «====Mq(C|x(18-16) level for 300 atom of Cl NbCL&(0-28) — TeCl,(7:65) e AGCL (1-88) o SHCl3 (0-67) —100 + e CdC2(0- 66) FUEL e SNCL2(0°32) / COMPONENTS 1250 K JA\C eI CL(0-06) EG .mol] —Zr C1,(215) —2004 e 7 Cl2 —PuCls Rt ) ——rurte gfnl%fiig%—-cagmw flLaClsts-m e —tac s (8 . — SmCls e S1C1 (5 °48) o £ 2C12(9+50) Free Energy of formation for fission product chlorides |Yields shown in brackefi] EFia 17 URANIUM - PRIMARY SALT PHASE irradiated Fuel UClg , Pu Cla FPA+4FPS+FPE in Na Cl IMETALLIC PHASE Mg Metal In Moliten Metal IME TALLIC PHASE Mg in molten metal U metal FPS + F met I:’Ernet SECONDARY SALT PHASE Pu Cls +MgCl g in NaCl OXIDATION SALT PHASE UCls in NaCl XTRACTION- FROM REACTOR / PRIMARY SALT PHASE SALT METAL ) /] EXTRACTION- _ irradiated Fuyel _FPA Chloride in Na Ct + Mg Cl» REDUCTION T 800°C _ [METALLIC PHASE g in moltenmetal Unet PYmet +FPS +FPE met met ME TAL SALT ) EXTRACTION- SECONDARY SALT OX IDATION PHASE T == 800°C J\ MgCl, in NaCl FRESH FUEL ) AND FERTILE MATERIAL PREPARATION T 700°C_J 7O REACTOR / MOLTEN FUEL OR FERTILE MATERIAL FPA — Alkali and Alkali earth fission products e-g. Cs Ba, Sr. FPS — Semt and noble metals and metal chlorides . FPE — Noble metals in metallic states and noble gases. 42 From fig, 19 it ~an be seen that all fission products inight be classitiad into 4 clasces, I'A - rission products of alkali, and alkalli earth b .t also rare earty elemenis whicn nave free enthalpy of cnloride formaiion »reater *‘bhan those of PuCl3. FP3 = fission products of ceminoble metals with free enthalpy of formulations smaller than those of PuClj. FPE = fission products existing in elementary form because of small free enthalpy of chloride formation or negative balance of chlorine. The proposed schema of the separa*ion processes utilizing metal transport is given in fig. 20 8. Aluminium trichloride as_ turbine working ageut One of the most important features of the proposed reactor is the relatively high total thermal efficiency of the power staiion, Such a high thermal efficiency is poscible only under two condi- tions: - working agent at a higher temperature - certain required thermodvnamic properties of the working agent must be met. Among several possibilities~ aluminium trichkloride, a rather exotic working agent is proposed here. 300 400 e Pd C'.z fMoCle TeCta -Te Cly ~NbCls -SbCly -CdClg -ZnCle ZrClg : Nd Cls, Lfl.Cls ~CeCly, PrcCly _RbCl [Cs Cl "SrClz BaClg e 10°K A Gfarmation i mini S e PRIMARY SALT PHASE (IRRADIATED FUcL) --—---fl-----a---“fl—-i———“-—- ———-—-————————-———-" L FPS — - : ‘Metallic Phase E iMetallic Phase : met"'"arnet 1 Pu+FPS 4 FPE ~ L------——“-JYEL-"!&*“M Lod L ------- = | ? 3 3 g ™ > 3 [ ] (7] ¢ 2 - v WASTES r- l | g i | | I T U RECOVERY i | | | ' Metallic Phase :U + FPS + FPE met met L-____ A pe-————=====1 t=-==y [Secondary E( hSalt ' lSdtH‘ISC: ' iMgCly PIJCl;pure' Pu RECYLING Fuel Reprocessing & Seperation Fig. 19 44 IRRADIATED FUEL FROM CORE 379 Salt.s"' _ et 21[?9PUS CONTINUQUS 00229 Pu-s \ 0:045¢ FP- s FUEL REPROCESSING (509% EFFICIENCY FOR ALL F.P) Pu INVENTORY ~ 600kg 3.7g Salt+ s~ 2329 Py. s~ RENTENTION TIME 00229 FP- s~ ~3 DAYS 0-0225¢ FP-s~' FRESH FUEL TO diluted In o THE CORE 1-85g Salt E o i FROM REACTOR o c 21-6g Salt: s~ g _ ce=1 - 1359 U-s7" CONTINUOUS - 0-066g Pu-s” & 0-005gF P- s~* FERTILE MATERIAL REPROCESSING 50% CFFICIENCY Pu recovery 0:033g Pu.s™ < —- Udept INVENTORY=*11600%g FERTILE MATERIA RENTENTION TIME o T0 THE Fission Products . REACTOR ~10 DAYS to waste g 21:0g Salt. s~ ' e 00339Pu S U‘deplflted . 0-0025q.5" _ ' 000259 FP- s~ input in 10-8g Salt. s 13-633g U- s~ +10:'8g fresh satt U-depleted ; Pu gain 0:0355 9.8~ 00115 9.5-’ FUEL REPROCESSING MATERIAL BALANCE Eimn 27N 45 The most important property of this substance is its spontaneous dimerisation at lower temperatures ALCL, + AL0L, e X o 3 1000 °K A12016 + AH AH = 125 kJ/mol This reaction results in a two fold decrease in volume but also release of some amount of heat energy. The physico-chemical properties of aluminium trichloride are very well known (see Blander 1957, Krasin 1967). The phase diagram is given in fig. 21. Table 6 Physical and chemical properties of dissociating-gas systems A1,Cl, and N.O 276 2 4 Al2016 N20] Molecular weight (g-mol—l) 266,7 92,02 Normal boiling point (OC) 193 21,5 Critical temperature (°C) 362,7 158,73 Critical pressure ( bar) 26 1033 Melting point (°0) 195 - 11 (at 2,46 bar) Heat of evaporation(kJ-kg-l) 15C 415 Heat of dimerisation(kJ-kg_%) 470 620 Type of reaction Al§016 Ni04 2!01 g NO ’ $ ° 2 NO + 0 . . N s LY LY - g - - I . Corrccian From the point of view of corrnsion the following regions can be distinguished, see Table 7. Components Temperature gradient Neutron dose Gamma dose Velocity of medium 2 m.s State Region F {fuel in core) NaCl PuCl 965-998°C l \N ‘M very high very high -1 liguid Region B iBlanket and coolant) NaCl PuCl_(little) . > Cl3 FP (1little) Sy ek Sy w— 750-793°C =73° high high 9 m,s 1 liquid The selected structural material is molybdenunm, Region A (outside core) —— — — — ——— - 750-400°C = 350° none very small 40 m.s-l gaseous/liquid The maia corrosion processes result from the following mechanisms (m = metallic phase, s = salt phase, Me = metallic component of irradiated fuel or coolant). 2 MD(m) + x For the reaction in region F of fresh fuel PuCl MeCl X (s) likely reaction is (1250 °k) A 0 C'lOC“J K » MoCl 3 S ) + 450 kJ/mol C1 > 2(s) 3 + E Me(m) in NaCl, the most H0012 + 2/3 Pu(m) 47 The equilibriuwum constant of this reaction is so small and equals 10717 that this reaction has no practical meaning. In region B the most dangerous reaction is connected with uranium tetrachloride, the product of the oxidation of uranium trichlorides ): (chlorine from fission of PuCl 3 + 1/2 C1 » UC1 UCl1 3(s) °(s) 4(s) + MoC1 “(s) °(s) 2 UCl4{S) + Mo(m) - 2 UCl The control of the UClj/U014 ratio in the fertile-coolant material might be feasible due to the continuous reprocessing of this material together with the control of zirconium from the fission products oxidation state, In the region A that is in the external heat exchanger the main corrosion process results from the action of gaseous aluminium chloride (the secondary working agent). (see fig. 22) A1C]1 — = MoC1 + = Mo + Al1C1 3(8) X " (m) X (g) This reaction was discussed in previous publications (Blander 1957) but unfortunately not all the thermodynamic data is known. Molybdenum forms four compounds with chlorine: MoClz, H0013, MoCl4, M0015 (fig. 22). The stability of these chlorines is strongly influenced by the concentration of free chlorine and also by the temperature, A more detailed calculation of metallic molybdenum corrosion in the aluminium trichloride is needed, see fig. A7. These calcula- tions are very sensitive to the vapour pressure of chlorides (fig. 23). In connection with corrosion problems mention must also 6360 : : 98wy = ~ 232 43-777-0-00612T+46 7?' acc,to BLANDER 3 « (@, 68000 - W x | n n tot x 0 6800 1 gouLID 680 1 S0 bar S0bar S0bar noregenerating __________ 55,6,_9”_ __80bar 386 C no overheating — 68 - ig.fi.,.w MC[’ B ! L. ALBry @, 34 : : 26 4. E 1 13-6450Up/ LIQwID ¥ GAS |} 100 - 68 Compe. | Turbine inlet 50 4 3-4- — e —————— e ——0 60 psIg 2-261 ; _- 2000 R. | -~ 20 136 |1 ] - 1alm. : 1 - 10-1068 § | : =7 i: i ____..--'""' 0344) ) [ - ll | H .——" :i o i ,..-'i' oulymt &g 0-068{)7 ["—‘ 172° 282° 392° 502° 612° 722° 832° 942° ©°C 800" 1000° 1200° 1400° 1607° 1800° 2000° °Rankine TEMPERATURE FREE ENTHALPY OF FORMATION [KJ'- mol"Cl] Molybdenum --Metal Melting Point-2610°C Boiling Point-5560°C (P=10-22 g.cm™ PRICE 30 $ /kg avc cast ngot L o | N ' ) el I o ° S MoCls flifigifi -8 ' { ' ‘ - 0300 500 | ' ' 1000 ' _,1.500' -P"’ TEMPERATURE °K _--" .-"'"‘- 180 |- - MO~ -200 | = ’.fl" r”f 1000+ 50 UC£3 .PU Cl5 SALT- FUEL COMPONENTS M.q Cla N.aCl 100 200 300 ) Free Enthalpy of Formation AGM [KI.molCEl CHLORIDES — BOILING POINT V 51 be made of the problem of the reaction between metal chlorides and oxygen and water, These reactions (for oxidation state + 2) could be written in simplified form: MeCl 5 + H20 - MeO + 2HC1 H9012 + 02 - MeO + C1 2 The metal oxides are mostly insoluble in molten chlorides, which results in a serious disturbance of the fuel system. From this point of view the metallic elements could be divided into three classes: (see fig. 24) - those which are stable against H20 and 02, that is the chlorides are more stable than the oxides (eg. Na, Cs, Ba) and partially Ca. - those which are not stable against H20 and 02 and the resulting product is a mixture of chloride, oxychloride and oxide (eg. Pu, U but also Zr, Ti, Al, PFe, Cr, Mn, Mg - this is the most numerous group of metals). - those in which chlorides are converted to the most stable oxide in the presence of H,0 or O, (eg. Mo, W), Metals of this class seem to be not so numerous as in the other two classes. This property causes the rapid elimination of traces of water or oxygen in the molten salts of Pu and U chlorides. It is also well known that traces of H20 and 02 have a very big influence on corro- ajion rate, 52 —400 - c ~3004 ol —200- cHLORIDES [KT.m 1000"% form AG —100- ¥ wfk—zao -300 . AG, OXIDES [KJ-Jymei 0] orm Fig 24 CHLORIDES—O0XIDES EQUILIBRUM DIAGRAM at 1000°"% The same property may be advantageous in establishing a thin coa- ting of oxide on molybdenum surfaces., This suggestion must be proved thermodynamically and experimentally. It must be stressed that the problem of removal of oxygen and water and other oxygen containing substances from the salts may be crucial for the corrosion problem as well as for long term fuel stability. 10, Safety Problems The molten chlorides reactor seems to be a relatively safe system because of the following reasons - an extremely high negative temperature coefficient of reactivity, because during a temperature rise part of the liquid fuel is pushed out of the core into a non-critical geometry burfer tank. The dumping of fuel in case of an incident is also possible in an extremely short time. (fig. 25) Pig. 25 is on the page 38, - in a more serious incident when the fuel temperature increases to 1500-1700°C (depending on external pressure) the fuel begins to boil. The vapour bubbles give rise to a new and unique, very high negative 'fuel void effect' - the leak of fuel to the conlant is probably not a serious problem because the coolant is continuously reprocessed. - the leak of coolant to the fuel for the same reason cannot cause large problems (provided the leak remains small). A rather adverse property of such a molten fuel reactor is the necessity of initially heating the solidified fuel in a2 non criti- cal geometry with external power. (eg. from the electrical grid). 54 11, BEconomics It is not possible, when considering the econimics of this type of reactor, at such an early stage to make realistic statements of . costs and predictions of economic performance for a station in full operation. Here we merely indicate the main areas in which this type of reactor can be expected to have an economic advantage including some comments on the possible attractions offered by a molten fuel reactor in utilizing the abundant but low grade sources of uranium which may become available when economic or national requirements dictate the need. The most important advantage ¢f the molten chloride breeder power system (power reactor including reprocessing plant) is of course due the part it could play in reducing the costs of power produc- tion. The possible eccnomic advantages are caused by the following features -~ in relation to the 'classical’ solid fuel fast breeder reactor, the molten chlorides fast breeder reactor system removes the need for the following operations: cooling of the irradiated fuel, transport, decladding, liquefaction, manufacture of solid fuel, cladding, fuel pin assembly, transport ete. - the amoun’t of fuel outside the core is, in the molten chloride reactor only a few percent of the fuel core inventory. In the solid fuel reactor the out ¢7 core amount equals the fuel core inventory. The capital zosts for the out of core fuel are of significance, —~ the doubling time for these reactors _s shorter than those for sodium cooled solid fuel reactors and, being equal to 7 years, gives a good doubling time. - the mean burn up in the molten fuel, continuously reprocessed could be a factor 3-5 lower than that for the solid fuel system. 55 - in a 'closed' system of power station and reprocessing plant the safeguards are much simpler and easier to apply. In addition the molten chlorides reactor has further advantages: - a higher mass specific power (MW(t)/kgPu) than the solid fuel reactor which decreases the fuel inventory capital costs. ~ a high power density (M‘a’(t)/m3 of core) than the solid fuel reactor, which may decrease the capital costs of core, blanket and shielding, perhaps in the future an increasingly important part of power production costs, - more attractive, from the point of view of future conditions when the costs of uranium recovery will probably increase and when independance from a foreign market may become an important factor., This last point is here developed further as of being of particular relevance to the Swiss economy and national interests but which may become more and more relevant to the world uranium market in the future. As with the classical fast breeder reactor, the molten chlorides reactor can be used for 'burning the rocks' (according to A.M, Weinberg), that is for utilization of the dispersed uranium present in granites in amounts of the order of 10 ppm (the mean value of uranium concentration in the entire earths crust is 4 ppm). The continuous reprocessing of irradiated fuel and the relatively simple preparation of fresh fuel and fresh fertile material as suggested for the molten chlorides reactor seems to be ideally suited for 'rocks burning'. 56 Let as make some simple calcuiations: 1 m3 content of 10 ppm we have 25 g of uranium per 1 m of granite equals 2500 kg of minerals, and with a uranium 3 ¢f granite. As is known in the Swiss Alps in Piz-Giuv {Aar-Massif) -ne uranium content in the syenite equals 15 - 30 ppm {Prof. Higi 1971) so in this case Tor 30 ppm. 1 m3 of syenite weighs: 2.5 x 103 kg b 6 2.5 x 10° kg x 30 x 1070 = 75 x 10°° kg = 75 g Uranium. In breeding 75 g U-239 -» 70 g Pu-239 for power production 70 g Pu x 1 MWa x 10° kw x 24 x 3.6 x 10° s = 6 x 107 KJ in other words 5 6 TJ or 6 x 1012 Joules, 1l m” of syenite is equivalent to 6 x 109 KJ It is interesting to compare this energy source with oil: 1 m3 and heat of combustion ~ 12000 kCal/kg gives: of petroleum product, with a specific weight og 900 kg/m3 > 10 000 kg x 12 x 10° x 4.18 = 4,5 x 10° ¥J = 4.5 x 10 Joules that is 6/0.045 = 130 times less than the energy content of 1 m of granite (syenite), 5 57 3 In terms of volumes, 1 m” of petroleum product is equivalent to 7 dm} of granite (syenite) (or a cube 19 cm x 19 xm x 19 c¢m) and in addition 1 m3 of 0il requires approx. 3000 kg of oxygen or 10'000 m3 air for combust*ion, In the future output of the Swiss nuclear power industry will reach 10'000 MW(e). Por this amount of electrical energy approx. 10'000 kg Pu must be burnt annually. In a steady state nuclear povwer industry this amount of plutonium could be obtained from 20'000 kg of natural uranium (allowing for losses etc.). So for the Swiss nuclear indu- stry it would be sufficient to recover the uranium from approxi- mately 1 million cubic meters of granite annually - in other words the volume of a tunnel 30 m x 10 m x 3000 m or an open mine 200 x 200 x 20 m, The cost of this rocks burning may be estimated =s folloxa The present price of natural uranium is 20 $/kg U for the case where vranium is recovered from 0.1 % uranium ores, In the future, when uranium is recovered from granites with only 0,001 % {10 ppm) uranium the price of uranium may increase to 500 g/kg that is 25 times. Let use assume that the price of plutonium (currently and in the future) equals 8000 #/kg. At the present time the proportion of uranivm raw material costs appearing in this plutonium price is only 20x2_ o,5% (2 kgU for 1 kg Pu) 58 In the future for the very expensive uranium from granites this part will equal Mz 12.5 % 800 Or in other words an increase in the ccst of recovery of natural uranium by a factor of 2% will only raise the plutonium price from 80008/kg to 90003/kg. The present cost of electrical energy with an optimistic figure of 4 mills/kWh{e) gives 1,1 mills for Pu per kWh(e) which at the higher price would give erergy (electrical} costs of 4,27 mills/ kWh(e) that is only 6,7 % more expensive than the current price, Thus to summarize: - The molten chlorides reactor is an attractive candidate for utilization of low grade ores such as from granite. - Even with the greatly increased cost of recovery of uranium from granite, tlc influence on electrical energy costs is small. - From a resource point of view, the energy content (per m3) of granite is greatly superior to say fuel oil (130 times). - The use of the abundant supplies of granit. is abviously an attractive possibility from the point of view of the economy and interests of countries such as Switzerland. 59 12, Some other selected breeders,