EIR-Bericht Nr. 276 EIR-Bericht Nr. 276 Eidg. Institut fur Reaktorforschung Wiirenlingen Schweiz Breeding in Molten Salt Reactors Lectures at the University of Liege/Belgium 15th May, 1975 M. Taube | I ,_l Wirenlingen, April 1975 EIR-Bericht Nr., 276 BREEDING IN MOLTEN SALT REACTORS Lectures at the University of Liége / Belgium 15th May, 1975 M. Taube April 1975 Remerciements Je remercie M., le Prof., G. DUYCKAERTS de 1l'Université de Liége, de m'avoir permis de faire cette conférence dans le cadre de la licence speciale en Science Nucléaire. M. Taube WHAT IS BREEDING? WHY IS BREEDING POSSIBLE? WHAT IS BREEDING GOOD FOR? Breeding is a process in which two mechanisms are occurring simultane- ously. 1) Fissile nuclides are 'burning' and producing energy and neutrons. 2) Some of the neutrons are transforming the so called fertile nuclides into fissile nuclides at a rate greater than the rate at which the fissile nuclides are being consumed. Breeding is possible because of three factors: 1. Some fissile nuc- lides have a rather large nett production of neutrons so that more than half of them can be used for breeding. 2. The fertile nuclides are converted to fissile nuclides only by the simple, and energetically not expensive act of neutron capture followed by two spontaneous processes which occur at a much faster rate than the rate of neutron capture. 3. Fertile materials are present in the earths crust at relatively high concentrations. (This also means that these fertile nuclides must be rather stable, beta stable of course, like all heavy nuclides they are alpha unstable. The a-decay rate must be matched to the age of the solar system). Breeding makes it possible to use not only the uranium-235 (the only ‘naturally occurring fissile nuclide) but also the uranium-238 which occurs in amounts 140 times greater than U-235. These factors permit the use of ores with low uranium and thorium concentrations even down to granites (>50 ppM U+Th). Practically only some reactor types are suitable for effective bree- ding: - all reactors with a fast neutron spectrum (e.g. sodium cooled fast reactor, gas cooled fast reactor, molten salt fast reactor) - one reactor type with a thermal flux spectrum (molten fluoride thermal reactor). FISSILE AND FERTILE NUCLIDES Definitions Fissionable Nuclides Fissile nuclides Fertile nuclides Characteristics Neutrons of any energy Only neutrons of high energy can produce fission (>1 MeV) can induce fission; Neutrons with lower energy are captured and after gamma emission (and in some cases then by beta-decay) a trans- formation into a fissile nuclide occurs. Examples U-233% U-235 Th-23%2 U-238 Pu-239 Pu-201 U-2310 Pu-240 Binding eénergy of neutrons and barrier to fis- Binding energy of the captured neutron is greater than the fis- Binding energy of the cap- tured neutron is lower than the fission barrier sion sion barrier (more %@ < Q exactly: fission n a activation energy Qg) only fast neutrons with kinetic energy E, can cause Qn > Qg the fission Qn + Ec > Qa Fertile Nuclide [Fissile Nuclide °q re s ) TP § \/ t 0.8 MeV MeV k 74 29 ‘ bar =4 Upatilug Transformation \ after neutron capture U-236 N Qn = Neutron Binding Energy > Qac Qbar Fission Energy Release Activation Energy for Fission = Energy barrier 198 o and Fission activation energy Q n - — Neutron binding energy Q 54 200 N/\r* Deformation 233 234 235 236 237 238 239 240 241 242 atomic mass, A THERE IS ONLY ONE FISSILE NUCLIDE EXISTING IN THE EARTH'S CRUST Atomic Mass, A 5 Fissionability Parameter % = 36 2704 @ 2604 Fm@ © Es non-fissionable c (] o [ 250 _ o ® qg Bk @ A - Aing o @ Pu-241 240 @FPu-239 230 - : Rn fissionable 220 At 104 51—209%’ le-208 0=-209 200 v T 4 T T T T T T | - | VLT v T v 1 g2 84 86 88 90 92 24 96 98 100 Atomic number, Z Half-1life years Es Md T R N Feo B M e (RN P S S WA S SRR RS (LS 82 34 86 88 90 92 94 96 98 100 Atomic number, Z for the following reasons 1. The fissile nuclides must have 2 2 % Z I >%6 (= I fissionability para- meter) 2. The fissile nuclides must be beta stable, because the half 1life of beta-decay in this region of A 1s smaller th%n 10 years. 3. Because for Z >36 gives in the realistic case Zz 92; all fissile nuclides are alpha un- stable. i, To exist in the earth's crust the half 1life of alpha decay of the nuclide must be of the same order of magnitude as the age of the solar system 2109 years. 5. Only one nuclide fulfills all these criteria: - U=-235 Relative 10 abundance t 1/2 = 10x10" "y: Th-232 1 S A t 1/2 = 0.7x109y: U-235 Supernova t 1/2 = 8.3x107y Pu-244 10 -2k | 4 3 2 1 Present time, Giga years J = 1 l ONLY TWO NUCLIDES EXISTING IN THE EARTH'S CRUST CAN PLAY THE ROLE OF FERTILE NUCLIDES The fertile nuclides must fulfill the following obvious criteria 1. They must be abundant on the earth. At least a mean value equi- valent to 1 ppM throughout the earth's crust (or more exactly in the outer layer of the earth's crust. 2. Thus they must be beta-stable 3, Thus the Z value must lie between 90 and 94 and be even 4, These are the isotopes of thorium and uranium with t 1/2 >lO9 years The only two 'natural' isotopes are Th-232 and U-238 These nuclides can be transformed into fissile nuclides by simple neutron capture and spontaneous beta-decay. Thorium Uranium Cycle /\ Stable in Nature (D veta stable (Q bpeta unstable —» beta decay ’f fission *neutron capture e X od = 2xi;;jL/ o = 10710 e ? thermal . . g is relativ big _ (n,y) o0 = l1o 10 Ox =5x10" %5~ Oz .9x107 7571 ® 2334 22 min 2.7 days ‘ i o = 2x107° 232 ] A Th Pa U 1 1 ) 90 91 92 A 2‘%1T 240 - 239 7 238 7 Uranium-Plutonium Cycle o = cross section for neutron reaction (cm2) @ = neutron flux (neutron cm—2s_1) A = decay constant (s-l) all data for ¢ E 1015n cm-zs_l L 7x10 fast O /c:m: 10710 e { ofast fls small o =410 TO PERMIT BREEDING THE NUMBER OF NEUTRONS EMITTED PER FISSION SHOULD BE LARGE The scheme of neutron induced fission of Pu-239 240 fi i 1 1 giPu+n > 9uPu 1581ons}22184 + 2?Nb59 t v ong (prompt neutron) delayed ’/// - V8~ (beta-minus decay) neutron lOOM emission 42 O58 Y stable 136I 53783 lB%XB- (beta-minus decay) 54482 stable The value of prompt neutrons fission = v . The value of v for uranium isotopes 2.5 The value of v for plutonium isotopes 2.9 . The value of v increases when the energy of the captured neutron is ~1 MeV W Table All data approcimate v Value of Neutrons per Fissioned Nucleus: v Nuclide | Thermal o .o .oactor reactor Y n n U-235 2.431 2,09 2.57 2.50 U-233% 2.48| 2.25| 2.51 2.51 Pu-239 2.87| 2.08| 2.88 2.40 U-238 - - 2.66 very small Th=23%2 - - 2.36 very small Pu-240 2.87 Pu-241 2.97 Pu-242 2.18 3.00 8 L1801 3.1 2.06 Y 1.26 habig oo Pu-240 [lu~238 T L] 2 L 1 10 1 10 10 10 10 Neutron Energy, eV ONLY SOME OF THE CAPTURED NEUTRONS CAUSE FISSION THE OTHERS BRING ABOUT A GAMMA EMISSION. THIS REDUCES THE BREEDING POTENTIAL fission fragment 83 fissions—=fission fragment ~\\*239 neutrons (83%x2.88=239) 100 8% atoms 100 atoms prompt neutrons (from neutrons —= of 17 atoms Pu-239=2,88 per fissioned fast Pu-239 of Pu-240 atom) gamma The ratio of non-fissioned atoms to fissioned atoms 1s called «o a = %%%f%% (in this example o = %% = 0.,20) We define: OYabsorption = o(n, ) + o(n,f) The real number of emitted neutrons per neutron absorbed equals: in this example - n = 2.88 ('§217) - 2.40 . om,f) .. L B O(H:f) + O(H,Y) l+a - 1 - L? = 2.88 T3 0.2 ° 2.u?d The value of n is strongly energy dependent The value of n differs for different fissile nuclides Value of n for fissile nuclides : y (simplified) Cross section Cross Section of Pu-239 BADHE (10—24cm2 (simplified) n 1000 1 ~ 100 o e ___.___..__-———-_,/—' ——/——n minimum o~ '-__________,—’ 10 ocapture‘\ o(n,y) Thermal Fast Reactors Reactors T T ] 1 0.01 T T T T ! 2 2 4 6 8 1072 1 102 10" 108 108 10 1 10 10 10 10 Neutron energy, eV Neutron energy, eV THE NEUTRON BALANCE FOR FISSION AND BREEDING IS AS FOLLOWS: Molten Salt (thermal) fluoride breeder for ~100 neutrons Molten salt (fast) chloride breeder for ~100 neutrons Nuclide Absorption Fission 232y, by, 7 0.03 233p, 0.02 233y 41,4 92.3 23L‘U Y 0,01 235y 3.4 6.8 236y 0.3 237Np 0.3 6L1 0.1 14 0.7 IBe 0.3 0.9 g 0.8 Graphite 2.3 Fission 0.7 products : Leakage 1.0 (nE 2.2317) Breeding ratio (total) 1.0708 Nuclide Absorption Fission 2350 (n,v) 22.51 (n,f) 2.99 8:.2% 239py (n,y) 5.58 (n,f) 28.98 85.55 2LmPu (n,vy) 2.214 v (n,f) 1.54 4.72 | § Na 0.26 @ Cl nat 3,16 Fe 1.3%0 Mo 2.04 Fission products A 23%(n,y) 23.15 N (n,f) UeB5 1.50 O Na 0.08 = Cl nat 2.22 3 Leakage 2.90 £ Breeding ratio core 0.716 blanket 0.670 total 1.386 Balance for Thermal Fission U-233 100 neutrons for fission 100 neutrons for fission Fission of 91 atoms of U-233 225 neutrons 6 atoms of U=~233 breeding gain Y £ Balance for Fast Fission of Pu-239 (Fission Products not shown) 100 neutrons u O Fission of 83 atoms v = 2,98 239 neutrons 100 neutrons for fission Pu-240 7 atoms DEFINITION OF BREEDING RATIO AND BREEDING GAIN BR = average rate of production of fissile nuclides average rate of loss of fissile nuclides Rmaximum = Breeding potential = n-1 G = Breeding gain = BR-1 = n-2; for breeder G>O 1) for typical fast reactors; value of breeding gain, G oxide fuel 0.25 (at present for 300 MWe LMFBR; G = 0,12 £ 0.03) nitride fuel 0.3%0 carbide fuel 0.4 + 0.47 (gas cooled FBR) molten chloride U955 2) for thermal breeder molten fluoride 0.06 all other fuels solid and ligquid <0.00 Typical value BR = [Mechanism for Pu fuelled fast reactor (+v |No. of neu- trons per (+2.96 fission -1 one neutron for fission -1 claim £ -0 loss due to . absorbtion 104 in Pu-239 0.24 Metallic fuel -A |losses in 131 Rt Structural - Molten chlorid material, etc. (0.1:0.25) e -L |leakage of 13] Fruoe neutrons -(0.04+0.06) ride -T losses due to absorbtion in |-(0.01+0.015) P +F ., |rate of fission of U-238 +(0.19+0.22). .(v=1)neutrons from U-238 fission |.(2.70-1)) ( 1 rate of ab- ( 1 ) 0.8 l+a/ |sorbtion in Pu-239 U-233 U-235 Pu-239 in (1 + 0.24) (n,y) reaction BRfast= 1.25 + 1.40 10 FOR THE CHEMIST, THE POSSIBILITY OF INTERNAL OR EXTERNAL BREEDING IS OF IMPORTANCE Primary Step .. Secondary Step Cycle Fertile Intermediate | Fissile Intermediate Fissile Thorium Th~232 Pa-23%3% U=-233 U=-234 U=-235 90 91 92 92 92 Uranium U-238 "Np=-238 Pu-239 Pu-240 Pu-241 g2 93 94 QY olu Fissile and Fertile Materials ‘Micro-mixing (U,Pu)O2 Macro-mixing UO2 particles + PuO2 particles In fused salt: PuCl3 in fuel, UCl3 in coolant/fertile material External Small critical mass, Using the fertile medium spectrum hard, breeding as coolant in a 'Chlorophil' ratio high cooling prob- type of reactor. Fissile TFertile lems, low Doppler effect, loss of reactivity due to burn-up <100 MWD/kg possible higher enrichment Internal No blanket spectrum softer, 1 BR, high cross se 299,99, tion of structural materials, good \’”"” Doppler effect, \\’” longer burn-up to \.A’,‘ 100 MWD/kg lower enrichment Note: for ~3000 MW(th) radius of core is ~100 cm thickness of blanket is ~100 cm Liquid fertile material as coolant 11 DEFINITION OF DOUBLING TIME Doubling time To is the period of time (years) in which a breeder produces enough fresh fissile material to fuel a new breeder react with the same power level (Teff includes the inventory of fissile materials out of core e.g. cooling, being transported and reprocessed). The compound doubling time Tigip'takes into account that in a breeder system, the new breeder can be fuelled with fissile material coming from the total system and not only from one reactor. _ Specific inventory (gram fissile/MWth) To - (L + o) *° G (days) G = BR—l; breeding gain r = fuel inventory out of core L = load factor (hours per 8760 hours in year) I = fraction of fission in fertile nuclide Teff - To (1+F) comp L(l-r) 1n? Doubling Time T (years) - 12 Doubling-Time and T 40w Breeding Ratio 30 Sodium cooled, burnup 10 MWD/kg fuel 1 worid Energy Production Oxide fuel, specific power 0.7 MW(th)/kg 20 / Nuclear Pdwer 15 = L 10 G e ] s e e e ] e | I I I ) I ’2 years _J | J 1 20 yea \\\:::;: ; 8 ‘yea 4 BR 0 T i il 7 \ T i 1 1 1 L] 1 Ll 1 1.1 1.2 1.3 1.4 1.5 1970 1980 1990 2000 2010 2020 2030 12 DOUBLING TIME IS COUPLED WITH FUTURE ENERGY DEVELOPMENT The rate of doubling (doubling time) of the total energy consump- tion is ~18 years and may increase in the future, Doubling time of electrical energy consumption is ~9 years. Doubling time of nuclear power capacity is less than 9 years. Doubling time of breeders should be of the same magnitude. In the future, civilization should reach a steady state when the doubling time for the breeder will satisfy the demand with Ts> ~30 years. Combined Breeder/Reprocessing Plant Uranium+Plutonium Uranium=-238 Fuel .3 kg/da Preparation ? e Y 1 kg/day Fresh Fuel 2.3 kg/day (= \ 0.33 GW(el) = Breeder Reactor Reprocessing 1 GW(th) Plant Energy 0.33 GW(el) Pu for sale —_— 0.3 kg/day 0.67 GW(th) E==——"" P - e et Irradiated Fuel L 2.3 kg/day ission Products kg/day BREEDERS ARE NECCESSARY BECAUSE THE WORLD RESOURCES OF 'GOOD' URANIUM ORES ARE RATHER LIMITED if one con- but the very A tremendous increase in available uranium ores occurs siders not only the classical ores (>1000 ppM uranium) abundant granites with 80 ppM uranium and thorium. Even if the price of uranium should increase a hundred fold the price of the raw fuel per kWhr will be no higher than it is now for light water reactors (since the uranium contains only ~0.4% of burnable U-235). 1 kWhr (e) £ 3 cents US, of this fuel % 0.8, from this plutonium 0.5 ecents. 1l g Pu 2 1 MWDth = 24000 kWhr(th) = 8000 kWhr(e) % 40$/g Pu for 1 g Pu £ 1.5 g natU e (114 present price 1 kg U = 30 $; 1 g 3¢ assume future extreme price 1 kg U = 2000 $ lg 2 3 1l g Pu will require 3 $ worth of Unat Each gram of Pu will be not 40 $ but 42 $ per g. The price of electrical energy increases to 3.25 ¢/kWh(e) World Resources of Nuclear Fuel UppM 2000 | 5 iD 16 ppM in Ocean natBLi ~30 ppM 16 in crust 1000 10 2.2xlO2 kg i 6 - ut Li is only 6x10 MJ or 2.4 ppM / 6x1016 kg 6Li 00 4 i > / Granitp 20 nat g U 10.0x102% Mg e 4 ppM in & crust 2 16 5% $/kg 8x10™" kg : 5% 7x102u MJ ’ug% —_— Us’g 200 o | 50 Consumption 23Th 12 ppM in crust s NSZi_gan. 100 = o — o — Granitp 100 A (y 8x10”° popu-U + Th 16 = lation at 21 x 10 kg 15 kW per capita 24 50 o & Cumulative { L 500 17 x 107" MJ =Y World needs to 2000 | | without breeder | I | | Complex Ores 20 o | | - 500 l : l D, Li = 2.75 MeV/atom = 2.75 x 10%" J/kg ‘y & High Grade Ores 1000 Th, U = 0.83 MeV/atom = 8.65 x 1013 J/kg ‘16 Earth's crust down to ~20 km % 2,4 x 19%° kg 10 T = = T- 100 109 12080 1011 1ol2 1913 1ot 1015 1o g Ocean = 0.14x 10 kg 14 FUTURE ENERGY NEEDS COULD BE VERY GREAT BUT WOULD STILL BE MET FROM NUCLEAR SOURCES IN THE FIRST INSTANCE TODAY: 3.8x109 people x 2 kW/cap. = 7.6 TW = 2.4x10°°J/year FUTURE: 8x109 people x 15 kW/cap. = 120 TW_ = 38x1020J/year 1 gram (U,Th) = 1 MW/day = 8.64x1010J 38x1020J/year 4,4x1010g/year = 44000 tons Uranium and/or Thorium per year But granite contains 80 ppM U+Th so the annual need for granites as '"fissile' ores will be 550 million tons. At the present time coal production alone is 4 times greater! Granite: Energy cost for 15 ppM U and 60 ppM Th. Granite 1s the main constituent of the earth's crust (up to 20 km deep) lO6 g 68.6 g/mol Free egthalpy of formation of granite = 210 kcal/mol ' 8.8x10°J/mol ?eghfiglogical free energy (electrolysis?) = 1.8 MJelec/mol + ’ therm/mol 10 Technological free energy for 1 ton granite = 5.2x107J. Amount of wuranium and thorium = 75 ppM = 75 g = T5MWD total = A 6.48x1012% Amount of electricity at 40% efficiency = 2.6x1014g = 1.46 x 104 mol 1000 kg granite = World Population X 109 1205000 1000 %/oo Solar Energy 9 N on Earth's Surface 8 s f$> 120 4 1 %/00 {promille) prodfction present 160 y 200 y 14 T Power per capita kW USA 50 > si°& > 4 10 J I T I 1] fi present 100 y 200 years present 90 100 208 years 15 ONE OF THE BIGGEST CONSTRAINTS TO FISSION ENERGY IS THE PROBLEM OF FISSION PRODUCT MANAGEMENT Each fission releases ~200 MeV; 1 Joule = B.lxlolo fissions. 1 watt = B.lxlO:LO fissions per second equivalent to 6.2X1010 fission product atoms per second. In steady state very roughly 1 Watt of power ~1 Curie of fission pro- ducts (1 Curie = 3.7x1010 disintegrations per sec). All fission products are beta unstable (neutron rich nucleil) Some fission products are long lived, comparable with a human life span or even the life span of an element of social organisation. @ beta stable Watt Obeta unstable T J \beta decay -~ 83 - Sm-l.Sl | \‘ Cs=-135 -8 1 daly lqu 109'd kyez'n' 10 yr 100 y : 1000 y 57 10" 0% 10° 107 10% 10° 1020 1ot a time, sec 16 THE PROBLEM OF FISSION PRODUCT MANAGEMENT TS VERY DIFFICULT AND REQUIRES A SOPHISTICATED SOLUTION The possibilities of radioactive waste management are: 1. Without the use of nuclear transmutation; that is the fission product nuclides are not changed, but merely removed and isola- ted; as for example a rocket to the sun, in discused salt mines or in the polar ice-cap. 2. By using nuclear transmutation in which the nuclear properties are so fundamentally changed that the transmuted products are short living nuclides which decay after a short retention time to a stable nuclide. In this second class of waste management techniques a number of exotic methods have been discussed - gamma laser excitation (does not exist) - underground neutron irradiation due to fission or thermonuclear explosion (e.g. more than 3000 explosions of 100 kT per year in USA 2000 only) - bombarding by protons (e.g. a 10 GeV accelerator with 1 Amp) - neutron irradiation in a thermo-nuclear fission reactor (does not exist) - neutron irradiation in a fission reactor (only pessimistic opinions). 1% Radioactive Waste /’ \ high active low active waste ‘waste any management management dissipation (forbidden!! no transmutation transmutation direct Antrolled without with with controlled storage changing changing changing storage on Z and A A Z the surface (b7<2) coulomb gamma (n,y) (n,2n) (p,y) (p,spallation) exitation laser Earth cosmic space (rockets) ) litosphere hydrosphere solar deep continously periodical (underground) impact cosmic ///////\\ escape salt rocks water ice neutrons neutrons fission thermonuclear mines (deep sheet (primary) (secondary) underground underground see (polar) explosion explosion floor stable geologic fusion fission accelerator conditions nonstable reactor reactor conditions (CTR) fast thermal thermal fast neutron neutron with without thermal thermal column column THERE IS A METHOD FOR PRODUCTS 18 'INCINERATION' OF THE MOST DANGEROUS FISSION Shown here are the most dangerous nuclides: per 100 Caesium Strontium Iodine Technetium Krypton (stable)* (2x106 years) (30 years) (stable)* (28 years) (stable) (1.7x107 y) (2.1x10° y) (stable) (stable) (10.7 years) (stable) 133 135 137 88 90 127 129 99 83 84 85 86 6.91 7.54 6.69 1.44 2.18 0.38 Lsly 5.81 0.35 0.56 0.672 0.882 21.14 302 1.55 | R S S 2. 47 fissioned atoms. \ 33,18 *problem of iso- topic separation The incineration is possible due to irradiation in a very high neutron flux, which gives, for example Cs=-137 53 - (:)Sr-9l gcapfure very| small (n,y) ®Sr-90 9.7 hr 52 = o 50 = () beta unstable ® the most dangerous stable nuclide (n,y) Cs=-13 'Decay' and Transmutation of Strontium=~90 64 hr . Zr-90 } (n,y) reaction: 4 B-decay 8 e 9in Ba-138 (stab "Decay' and Transmutat (n;p) le) ion of Caesium=-137 (:) Cs-138 acapt 32 min very| gmall 82 ‘ Cs-137 Ba~138 A 0y (n,2n) Cn ) \J 81 4 (:) Cs-136 Ba-137 A N 2.9 d 80 - . @s-us Ba-136 Qbeta unstable , 106 thhe most dangerous N y @stable nuclide 75 Q -5 T T T 54 55 56 Z Xe Cs Ba 19 THE NEUTRON TRANSMUTATION OF F.P., NEEDS A VERY HIGH THERMAL NEUTRON The efficiency of the transmutation is limited by the criteria that the probability of the bombardement of an atom by a neutron is ~10 times greater than the probability of spontaneous beta decay. A(neutron capture) 2 10-A (beta decay) We know that for a neutron flux @ (n cm 2 ) and the capture cross section o, (cm?/particle) the probability of reaction equals. A reaction = @+0; therefore @#:-0 >Adecay For caesium-137 we have Adecay = T.7 X 1 P i ccapture (thermal) = 0.06 x 10_2Ll cm2 occapture (fast) * 0.01 x 10°2% cn? For the postulated neutron reaction probability Rt 0 - =15 ?> l%%%él > @fast; @fast L. 7x10 — = 8X1016; Pthermal Y. A50 i P 0.01x10 10.6x10 = 1,2x1016 AL Possibilities of transmutation Thermal Particles Fast Source of Tower _@Q Particles Protons neutrons rating (W/cm ) 200 Accelerator 30 MeV - Fissile _2h_ ~30 MeV protons concentration (N) 10% for 30x10 (secondary) (10—24 700 Accelerator 1 GeV High flux 9fission =2 - >1 GeV protons for of neutrons cm ) 2 Flux 0 _ §;2§l915 (p,spall) _ nem2g + , B = 2.6x10 primary Nuclear - very high explosion flux in very short time C.T.R. = High flux, (controlled continously Fluence, 0.t thermonuclear t = (days) 700 reactor -5 £ 50 Fission = Very high .t = (ncm ) é-ZEipgz reactor flux 2. 2%l continuously 20 IT MAY BE POSSIBLE TO COUPLE A BREEDING SYSTEM WITH A HIGH FLUX TRANSMUTATION FISSION REACTOR We postulate a system with breeding potential and with transmutation potential. The compound doubling time Ts must be at least 30 years, . _ 1000 . M. (1+F) i B} sincey; Ts(year) = Z65 (BR-1) (1+a) L 1n2 = 30 years M, (1+F) (30.(1+a) - L from this;minBR =1+ 2.75 « 1n2 for typical values; minBR = 1.077 we know that: maxBR - vol-a- (A+L+T) +F(v=-1): when T 1s very small (1+a) (T=transmutation rate) For the desired BR . = 1.077 we obtain: min T = wv-1-a-BR(l+a)=-A-L+F(v-1) For typical values: T = 0.364 That is, for a breeding system with a breeding potential of BR = 1.4 and for a compound doubling time of 30 years, approx. 0.36 of FP nuclides could be transformed in the fast system. Organisation of transmutation and breeder system Breeding and Transmutation (history of only F.P. and neutrons) Our system is: - breeding system with compound doubling time 30 years. for/fission ~ breeding potential BRpot Losses ete. = transformation rate 0.36 ~ 1 burner reactor for transmutation - breeder power reactors = 1.4 100 Fissions 00 , We obtain: Producis 140 ' X ° BRmaX = (X+l ) BRmin T . but we know BRmaX = BRmin + TEIET B o from this, for typical values _ l+a _ reedng with X - BRmin T - 3 . 7 dRubliyg time framsmutation M 30 years Result: for about 3.7 power units of breeder reactors and 1 power unit of the breeder transmutation reactor 0.36 of FP nuclides of all reactors can be 'incinerated'. mutation of ‘selected Stal Long 21 WHAT IS THE CONNECTION BETWEEN BREEDING AND THE FUEL CHARACTERISTICS (PHYSICAL STATE: SOLID/LIQUID, CHEMICAL STATE: METALLIC, CERAMIC, SALT)? Breeder reactors are feasable using both thermal and fast neutrons. But for thermal neutrons, breeding is possible only in a very limited region the thorium-232/uranium=-233 cycle in a liquid fuel reactor. Breeder Reactors Neutrons Thermal Fast Fertile/fissile Th-233 U238/Pu239 Th232/U239 U238/Pu239 Cycle Potential Bree- ding gain ~0.1 <0 ~0.3 ~0.45 G=BR~-1-losses Impact of inter- Pa-233 breeding Pa-233 Np-239 mediate mole- th_ not possib- _fast_ fast e — cab—large 1e Oab =small Oab negligi ble be- cause tl/2 small Consequence Continuous _ no limitations, periodic Pa extraction rechanging possible Results liquid fuel no limitations, solid fuel (aqueous or molten) possible Maximum losses in core struc- tural material and coolant ad i P ~0.2 ~0.3 Consequences Moderator D.O _ Metals possible, graphite or graphite forbidden Consequences No metallic metallic tubes in the core tubes in the core possible Impact of coo- ling system out of core no limitations, in core coo- ling possible Fuel system liquid fuel solid and liquid fuel, cooling in/or out of core. Solid fuel: ceramis, metals; liquid fuel: chlorides only. Example only for liquid fuel reactors only, cooled out of core molten |aqueous alt solu- LiF tion BeF2 suspen- iy 560 ho Ty 5! PR UFH- in D2O graphite molten salt molten salt NaCl NaCl 232Th01u 238U013 233UCl3 259Pu013 22 THE LIQUID FUEL REACTOR ALSO HAS OTHER ADVANTAGES IN RELATION TO THE CLASSICAL SOLID FUEL REACTOR Solid Fuel Liquid Fuel Pressure pressure of volatile fission no FP pressure: Very low products. pressure of fuel components Cooling cooling only in the core cooling in the core or Problems out of core Burn=-up periodic recharging continuously recharging e.g. typical case PWR: specific MSBR: 22 KW/litre power 40 KW/kg fuel, burn-up 30000 KWd/kg fuel, dwell time 750 days ~dwell time 10 days Reprocessing | periodic continous Heat Transport in the Core solid fuel liquid fuel "’,——~—“-‘-\\\\~ [’/’,,——-~———-§\\\\ fsggeraturef‘ A ¥ g 30004 20001 é Coplant lanty Temp. 1000 J \ H 4 ' fd 41 4 23 Fuel thermal fast ambient pressure overpressure liquid solid metal now not possible i \Ha) projected LMFBR 0] 0 e MSBR _ = I= salt (fluoride) 3 % MSFBR & |- Chlorophyl _ % - (chloride) . = - LWR E water HWR P not not ~ g possible possible T o] - English § co French . ) o | 2 GCR 2 o0 No No [63] g - HTGR He HHT - GCFBR GCR = Gas Cooled (thermal) Reactor GCFBR = Gas Cooled Fast Breeder Reactor HHT = High Temp. Helium (thermal) Reactor HWR = Heavy Water (thermal) Reactor LMFBR = Liquid Metal Fast Breeder Reactor LWR z Light water (thermal) Reactor HTGR = High Temp. Steam Turbine Reactor MSBFBR = Molten Salt Fast Breeder Reactor M3BR = Molten Salt (thermal) Breeder Reactor 24 MOLTEN SALT FUEL APPEARS TO BE THE BEST LIQUID FUEL Only limited possibilities exist for choosing a fuel to operate in the liquid state in a reactor. The limiting criterium seems to be the radiolysis in the extremely high field of neutrons, fission fragments and gamma rays. In addition the chemical stability of the dissolved species of thorium, uranium and/or plutonium in the presence of the products of radiolysis is a severe problem. The ionic liquids, that 1s the molten appropriate media for fission fuel ionic salts seem to be the most (see table of electronegativity overleaf). Liquid state Range of 1li- Stability Reactor type quid state (°C) Thermal Fast Aqueous solution| 01.2 Chemistry of molten salt Fluorides only. positive experience exists Chlorides, or fluorides (?) no experience with chlorides Structural materials Only graphite-direct contact, no metals (no parasitic capture) metals in core permitted. Cooling Out of core cooling Either in core or out of core cooling Present day develop- ment USA - Oak Ridge; France Fontenay aux Roses; India - Trombay? UK., -Harwell, Winfrith (2); Switzerland: Wirenlingen EIR Moderation of neutrons Fluoride in large amounts acts as a weak moderator The moderation factor for chlorine is 3 times lower and the concentration is also smaller Electronegativity Super Actinide Components of 0l -of molten salt reactors [] thermal Zl fast Energy Decrement 6 8 10 12 1l 16 18 20 22 o4 26 28 A 26 MOLTEN FLUORIDE VERSUS MOLTEN CHLORIDE FUEL - I Criteria for molten salt as fuel: Melting point: lower than 500-600°C Boiling point higher than 1500°¢ (partial pressure) Density: as high as possible, p= 3 gem - | . thermal . _fast Capture cross section low Jm—— <0.,16; 9 s mpts small Cross section for nuclear reactions low; e.g. for 3501 (n,p)BSS Elastic scattering: high for thermal, low for fast (see previous page) Radiolytic stability high and fast recombination (previous page) Chemically stable (large free enthalpy of formation), no precipitates Simple technology, low corrosivity Low viscosity (low pumping requirements) ~1 centipoise Low price, good avallability, non poisonous 1500 o 3 X only as a small component of a solution ThF . . ; .ll (111°%) limit for single substance PuFu 1000 = Melting Temperature (°c) limit for complex mixture 500 - - - UCl6 non stable 5.35° density of UClx (gcmflj) PuF6 UFg T I i | 1 I5E v v VI Oxidation State MOLTEN FLUORIDE VERSUS MOLTEN CHLORIDE FUEL - IT 27 - MELTING POINTS Fluoride Chloride Reactor Type Thermal (fast possible) only fast because Gcapt Cl is large Components - fertile ThFu only UC‘l3 or UClu(unstable!) | fissile UFM or UF3 PuCl3 only dilutent| LiF Possible but also LiCl forbidden since Bel, plays the role of BeCl, they act as 2 2 moderator moderators NaF forbidden since . T is too high NaCl possible melt Moderator Li.F, BeFo not suffi- none cient. Graphite 1s nec- cessary: does not react with molten fluoride Metallic tubes in the core not permitted due to neutron balance permitted, e.g. a molybdenum/iron duplex Melting Point of - Fluoride 1100 + 1000 4 700 T - Selected Salts 11004 700 500 7 Chloride 0 0:.25 0.5 mol fraction 0.75 1.0 mole fraction 28 MOLTEN FLUORIDE VERSUS MOLTEN CHLORIDES FUEL - IIT CHEMICAL STABILITY The free energy of formation for fluorides 1s greater than that for chlorides. The formation of oxides of uranium is retarded in fluorides. The ratio AG (HF/HZO) 1s not the same as AG (HCl/HzO). Carbon tetra-fluoride is a product of graphite and fluorides. Free Enthalpy of Formation Free Enthalpy of Formation 1007 1oofl G -2007 -200 ‘ ;. .——‘:::::— (KJ/mol EL_ (KJ/mol C — -300 300 NaCl -400 " -1400 4 — /o/ /u/o -500 47, <5004 4o, " ——’b. -600 T T T T T T — -600 6‘0 50 850 JO 1550 1100 500 600 700 800 900 1000 1100 1200 S ° ! ’ Temperature (K) Temperature, K 1 1200 29 MOLTEN FLUORIDE VERSUS MOLTEN CHLORIDES FUEL - IV STABILITY AGAINST FISSION AND CORROSION PRODUCTS For molten fluoride the crucial mechanism UFu(d) + 1/2 Cr(Ss) UFB(d): K = 1.38 x 10 i P 1000K . . . Numerical Values given for AG in 2L1F—BeF2 solution Fuel Structural Fission Fuel Structural R Components Materials Products Components Materials 100 +100 ~ o ™ Moc1 4-10C13 2 CFU - -100 -100 — -FeCl, ' CrCl. .4 AgL000K TeFg AGLO00K 2 RuF o¥ i Rqu _UClu -200 oF , MoFy L pop -200 - 3 Th014-~UCl v (KJ /molC}) o 3 (KJ/molf") L i PuCl3 —PaCl — Cerq-Fer -300 | FAmClB BeF2 CrF_ NaCl — 3 rUFll -~ UF 3 -400 PuFB- ThFu Zer N -400 Aij_ CsF - PaFBA Rbi? L Sml, [~ PrF3 -500 + LiF ga{ - CeI*.j =500 = Skl - -600 -600 Fission Products - SeCl = ZrCl2 SrCle—_ RbC1 BaClz“_ CsCl 50 MOLTEN SALT FLUORIDE BREEDER (THERMAL) POWER REACTOR 1 GW(el), MSBR (Oak Ridge concept) 7 Fuel: 'LiF (71.7 mol%), BeF, (16%), ThF, (12%), *33yr, (0.3%) 2 Tfuel = 839-978 X Tmelt = 172 & Breeding ratio = 1.06 Doubling time = 22 years 1.-1 Density (~908 K) = 3330 kg/mB. Heat Capacity = 1.36 J g K Viscosity (~908 XK) = 0.01 N s m ° Structural material: graphite (in core) Hastelloy N Problem of delayed neutron precursors. 4 uynits 1.8 MWel Primary Salt Pump 1m3 /s NaBFy-NaF Secondary Salt Pump 727 K Purified Salt “Graphite Moderator Reactor 2250 MWth Heat ' e 5 menanser] [ TLiF-BeF_~ThF, -UF 2 4 4 Steam Generator Fuel Salt e f—— Chemica Processing Plant Turbo- Generator Steam 31 A FAST BREEDER REACTOR WITH MOLTEN CHLORIDES COULD HAVE IN CORE COOLING Total power: 2050 MW(th) Core volume 7.75 m3 (2m x 2.36 m) Fuel: PuCl3 (15 mol%) NaCl (85 mol%). Tmelt = 685005 0 2,34 gem Coolant, fertile: ZB8UCl3 (65 mol%) NaCl (35 mol%) Tm = 71000; 3 p= 4.00 gém- 5 elt Specific power: 220 W/cm3 core; 705 W/g Pu Velocity: Fuel 2 m s—l; coolant 9 m s+ Number of tubes in the core: 20,000 (1.26/1.20 cm dia.) Reprocessing: continous, 3 gs-l; dwell time: 10 days Breeding ratio: internal 0.716; outer 0.67 total 1.386 (corrected) Mean fast flux across core: 7 X 1015 n cm-2s—l fast flux in the centre: 1.2 X 1016n cm_2>s-l Temperature reactivity coefficent (8k(%/6 T(OC)) fuel = 3.8 x 10°°; coolant +1.29 x 10 ° Structural material: Fe (1.45 kg/l): Mo(0.465 kg/l) ;753 750° 790°¢ k:i) - _::::F:ji=10m % 1L o a4 f?fi: / /;?f;/ / CEEEE:::: 93;g°c fuel ?EEEEEEE 1y m ) e I // ¥ // ” %‘ ////// 32 THE VERY COMPLEX PROBLEM OF CONTINOUS REPROCESSING OF THE FUEL IN A MOLTEN FLUORIDE REACTOR HAS BEEN SOLVED The operation of a molten salt thermal reactor as a high performance breeder is made possible by the continuous reprocessing of the fuel salt in a plant located at the reactor site. The most important opera- tion consists in removing fission products (rare-earths) and isolating Pa-233 from the high flux region during its decay to U-233 in order to reduce neutron absorption. It 1is also neccessary for: - the removal of excess U-233 - addition of fresh amounts of Th-23%2 - the proper redox potential to be maintained - the oxides to be removed - corrosion products to be maintained at tolerable levels In order to achieve the breeding ratio of 1.06 - 1.07 the following fission products must be removed as shown. Dwelling Time of F.P. in Molten Fuel Rave Earths v Noble Metals _. Se, Nb, Mo, Te, Ru, Rh, Ag, Sb, Te, Pd Semi-Noble Metals Zr, Cd, In, Sn Noble_Gases Kr, Xe Volatile Fluorides Br, I Y Non-Volatile Fluorides v Rb, Cs, Sr, Ba Protactinium v Actinides Wp, Pu 1 day 10d 100 d 1 year 10 y T T T T l! T ll = T - L 1 1 102 100 10" 10 10° 107 108 109 time, sec 33 Reprocessing in MSBR Power Reactor ~2250 MW(t UF ) 6 Collection SR— 705%¢ Processing A time 10 days D.02 g/sec A B | | [rradiated Huel Efficiency of RBemoval ~95% \‘-Fluorination U-233 or Bi in fuel <2 ppM Extr ion < 1 (3-6 stagks j Salt + R.E ) i - A ¥ Extpadtion # | 2 L/ Reductive '—‘ * Extractién A k ExtrT T=640°( 5 Molybdqnum Separa -|as g ti¢n ot | Structyral e, R.E and . Pa Mater1q1 Extrl uy Bi/Li (0.5 m) | EL! THE EXISTING REPROCESSING SCHEME IS TOO CONSERVATIVE. IT DOES NOT TAKE INTO ACCOUNT THE PROBLEMS OF RADIOACTIVE WASTE MANAGEMENT PRESENT STATE POSSIBLE FUTURE FORM Reprocessing periodilc quasi=-continuous Removal of F.P. in a mixture as individium Transmutation of hagzar- no possible in a special dous F.P. central station Actinides (Np, Am, Cm..) removal to radio- recycled in a special active waste power reactor? Reprocessing today Reprocessing in the Future Fresh Fuel Fuel Pr:epara- ‘uel 2% Pu + ~100% Np, Am, Cm tion Preparatio 98% Pu + U | Power Sink A Irra- Reactor diated Reprocessiyig Fuel Electrical |_Reactor 2 = . Energy 9 Electrical SpIatessIng / Energy U losses ; RE_D QSSegS ,; Irradiated Fue 2% Pu lossef / Sink Np, Am, Cm / Stable or Np, Am, Cm lossed 100 % ¢ short lives “ 0.001 % ¢ V - Storage Transmutatian ; F ‘:)_ Reactor ; / “ L L long lived F.P. v Stable Nuclide 55 THE LONG TERM HAZARDOUS TRANSPLUTONIUM ELEMENTS CAN BE FULLY BURNED-UP IN A FISSION REACTOR Parallel to the burning of U and Pu also synthesis of Np, Am, Cm, Bk and Cf is going. At present time the actinides, other than U and Pu, are removed to the waste. The problem of the hazard from these actinides in long term must be solved. The recycling of actinides bach to the power reactors gives the so- lution of problem. The methods of the recycling must be developed. Relative Toxicity . Hazard from Nuclides actinides 242 all other F.P. 240 239 4 Cs + Sr 238 36 CHEMISTRY OF FLUORIDE FUEL Fluorides are the only salts with acceptable Oobs? requisite stability (AG) and melting point. Oxidation state UF3 due to stability and Iow Tmelt PuF3 the most stable but in both cases ThFu the only one Tmelt 1ls very high ; . Components 1) 'LiF and BeF2 for lowering Tmelt 2) for moderating 3) 7Li because: gLi(n,a) g Oip = 953 barn ) LiF/BeF2 for low viscosity Tritium production: in spite of 7Li the products are (for ~1 GW(el)) 253 7 U(n,f) ternary:30 Ci/day Li(n,on) 3‘I‘ : 1200 * ®Li(n,a) °T: 1200 " Graphite - as moderator and structural material (p = 1.9 gcm-s) b UFM + C CFH + MUFB: e.g. pressure CF) = 10—8 bar. The problem of xenon-135: This F.P. is adsorbed in the goen pore volume in the graphite; accesible void volume ~10% graphit replacement in the core ~each 2 years Disproportionation K9OOK YUF,(d) 3UFy(d) + U: - 6.3 x 10 ¢ Instability in the presence of oxygen-containing materials: UFM + 02 UO2 T oeeen solubility ofUO2 in molten salt ~40 ppM also precipitation of Pa20 PuO 52 2 27 FLUORIDES F.P. Balance: UFu > 2 atom F.P. + 4UF - i, > MelF + MeIIFZ + Me™"TFy + Me® +~ + F, The Breeding Ratio is very sensitive to the removal rate of RE and Pa-233, Structural material for reprocessing: TZM: molybdenum alloy for fuel system Hastelloy N (modified) (wt%) Ni 75 8 0.1 C 0.06 Mo 12 W 0.1 P 0.015 Cr 7 Al 0.1 S 0.015 Fe 4 Ti 1.0 B 0.001 (10 ppM) Mn 0.2 Co 0.2 Hf 1.0 Ta 2.0 Cost of salt 57$/pound, in this U=-233 $13/g; total cost 23x10 $ Breeding Ratio v Reprocessing days Pa-cycle e [J=2 33 1.05 0.95 4 0.90 « T T T T 100 200 . 300 400 500 Recycle Time days 38 ANALYTICAL PROBLEMS FOR MOLTEN FLUORIDE FUELS ARE OF THE HIGHEST IMPORTANCE In order to fully exploit the unique features of the molten salt reactor concept it will be neccessary to carry out all analyses automatically on line with transducers located directly in the salt streams in the reactor and reprocessing plant. The most significant items to be measured are - Redox conditions of the fuel: the U3+/ULl+ ratio which influences the rate of corrosion and the distribution of certain fission products and tritium in the reactor. This is due to the disproportionation WUF5 + 3UF, + g® (only in molten state) (by spectrophotometric methods also voltammetry) 2 ..+3 - corrosion product concentration (Cr+ Ti -, voltammetry) - oxide levels (by hydrofluorination: MeO + HF » H 0 +‘MeFX) 50 ppM precision | 2 - bismuth (polarographic method 50 ppG) - hydrogen and tritium (diffusion to Pd-electrode some ppM) - protactinium - certain fission products (remote gamma spectrometry and for noble metals mass spectrometry) Note: Most of these techniques are carried out in hot laboratories. The electrochemical technique appears to be the prime candidate for practical on-line fuel analysis because of the simplicity of its trans- ducers. The analytical and general chemical problems in the molten chlorides are also of interest because an important step in the continuous reprocessing technique is the liquid bismuth/molten chloride reduction extraction metal transfer process. 59 Chlorine 35 reacts with neutrons: 35Cl(n,p); 35Cl(n,a)32P Separation of 3501/3701 is relatively easy because 1 3Te1 in the naturally occurring Cl is 24.5% 2 the ratio 2%%25 = 0,057 3 Chlorine is volatile WITH MOLTEN CHLORIDE FUEL IT SEEMS THAT A CONTINUOUS GAS EXTRACTION SYSTEM IS POSSIBLE Iodine isotopes as precursors of delayed neutrons Isotopic separation e.g. Caesium Chlorine in Neutron Flux Z 51 52 53 54 55 5p L L 38 —+Sb = Te—t I—Xe —»Cs —=Ba A=137 from — 4 fission Aj j I 21 T [:] stable 100 0captur 3¢ min Independant (:) beta-unstable very sfall Yield 10 1 0] jom [ O g 5 Cs-137 Ba-137 Half L a te 103 30 year stable Seconds 2 10 10 19 = 1 Y T T T T = Extraction Rate 18 A s-l 14 17 - 4o THE BALANCE OF CHLORINE DURING THE FISSION PROCESS IS "NEGATIVE" PuCl, (n,f) 100 PuCly (n,f) o+ 4+ + =+ 100 PuCl, (n,f) LITERATURE E! + E" + 3C1 0.008 Se + 0.003 Br + 0.942 Kr + 1.05 RbCl + 5.49 Sr‘Cl2 3.03 YCl3 + 215 Zr’Cl3 + 0.29 NbClS(?) + 18.16 MoCl2 + .01 Tc + 31.45 Ru + 1.73 Rh + 12.66 Pd + 1.88 AgCl + 0.66 0.66 CdCl, + 0,06 InCl + 0.325 SnCl2 + 0,67 SbCl, + 2 3 7.65 TeCl2 + 6,18 I + 21.23%3 Xe + 13%3.35 CsC1l + 9.50 Ba012 5.78 LaCl3 + 13%.98 CeCl3 + 4,28 PrCl3 + 11.87 NdCl3 + 4y PmCl3 + 3,74 SmCi_, + 0.60 EvCl. + 0.03 CdCl 3 2 3 200 ECll,.5 For molten salt reactors only, estimated to the recent publication Harder B.R., Long G., Stanaway W.P. McNeese L.E. Rosenthal M. 3 W. .Rosenthal M.W., Haubenreich P.N., Briggs R.B. Taube M. Taube M., Ligou J. Taube M., Ligou J., Bucher K.H. Compatibility and reprocessing in use of molten UCly alcalichlorides mixtures as reactor fuel in "Sympos. Reprocessing Nuclear Fuels" Ed. P. Chiotti, USAEC, Con-690801 (1969) MSBR- a review of its status and future Nuclear News. Sept. 1974 The development status of molten-salt breeder reactor. Oak Ridge, ORNL-4812, 1972 Steady-state burning of fission products in a fast thermal molten salt breeding power reactor. Ann.Nucl.Sci. Engin. 1, 283 (1974) Molten plutonium chlorides fast breeder cooled by molten uranium chlorides. Ann.Nucl.Seci. Engin. 1, 277 (1974) The transmutation of fission products (Cs-137, Sr-90) in a liquid fuelled fast fission reactor with thermal column. EIR-Report Nr. 270, Feb. 1975