EiR-Bericht Nr. 259 EIR-Bericht Nr. 259 Eidg. Institut flr Reaktorforschung Wirentingen Schweiz A High-Flux Fast Molten Salt Reactor for the Transmutation of Caesium-137 and Strontium-80 M. Taube, E.H. Ottewitte, J. Ligou l__l Wiuarenlingen, September 1975 [ I EIR-Bericht Nr. 259 A High-Flux Fast Molten Salt Reactor for the Transmutation of Caesiumn-13%7 and Strontium-90 M, Taube, E,H. Ottewitte, J. Ligou September 1975 sSummary 1. Introduction 2., rormulation of Reactor Requlirements 2.1 Minimal System Doubling Time Requirements 2.2 Heed for Central Flux-Trap 2.5 Determination of Flux-Trap Size from Destruc- tlon-Production Balance Reguilrements Weutronic Consideration 5.1 Burner Reactor Calculations 5.2 IPoderator Requirements 4.4 Motilvations for Molten Salt Fuel 5.4 Outer-Reflector Zone Consideration Tnermohydraulic Considerations Murther Parameters of the Burner Reactors Effect of Keplacing Chlorine with Fluorine Remarks about Transmutation and Hazard Coefficients Conclusicns Acknowledgment neferences page Summary A high flux molten salt (plutonium chlorides) fast reactor (7 GWth) with internal thermal zone for transmutation of Sr-=90 and Cs-137 is here discussed. These fission products have been produced by breeder reactors with total power of 23 GWth and by the the 7 GWth fast burner reactor. For the case when the power breeder reactors achieve a breeding gain G >0.2 the doubling time for whole system, including the high-flux burner reactor, equals ~30 years, The transmutation of 3r-90 and Cs-137 in a total flux of 3,8-1016n cm_gs_l, and thermal flux ~2,O'1016n cm_gs"l can achleve the steady-state which corresponds to an effective half-life of 1,8 year for Sr-90 and of 8,9 years for Cs-137. In terms of hazard coefficient the transmutation system gives an improvement of 14 times. The impact of numerous parameters is discussed, e.g., chemical form of both nuclides, nature of moderator in thermal zone, moderator, layer, fuel composition, radius of thermal zone, ratio of burner/breeder reactors etc, The effects of replacing chlorides with fluorides as fuel for fast core is also dis=- cussed. 1. Introduction The alm of this work was to study the concepfi of transmuting Sr-90 and Cs-137 in a very high-flux reactor. These two isotopes are of particular interest because of their high yield in fission, their longevity and their biological hazard. The rather high yields (Y: mol %) from fission are U-233 U-235 Pu-239 Sr-90 ~6,2 ~5,1 2,2 Cs=137 6,6 ~6,0 ~6,7 Assuming each of the above fissile fuels to be equally prevalent in the future, the mean yield is 4.1% for Sr-90 and 6,4% for Cs=137. The half-lives (28 years for Sr-90 and 30 years for Cs-137) indicate the longevity of the isotopes. To "significantly" reduce their activity (by a factor of 1000) through natural decay would take 300 years, The blological hazard 1s reflected in the maximum permissible concentrations (IAEA, 1973): Nuclide in air3 on water (uCi/em”) (uCi/em?) Sr-90 1.207° 1.107° Cs=-137 6-10"8 4-10'Ll This 1indilcates that Sr-90 is roughly 50 times more hazardous than Cs=13%7 and warrants prime attention. These facts have understandably prompted numerous transmutation studies (Schneider 1974; see also Taube, 1975). The most opti- mistic has been ftransmutation in a controlled thermonuclear reactor CTR (Wolkenhauer, 1973). However, even unrealistic flux levels in CTR could only reduce the transmutation half-1life from 30 years down to 5-15 years. The time required for "significant" reduction is five to ten times longer than indi- cated above, That would be longer than the usual life of a power plant. Furthermore, Lidsky (1975) has pointed out that for any reasonable ratio of fusion burners toc fission reactors, the burners soon contain a much higher radicactive burden than the fission reactors themselves. This is certainly undesirable, and probabvly intolerable. A motivation for alternate solutions remains. This work examines the basic reguirements to obtain a system cf reactors which provide adequate breeding and self-destruction of figsion products. These requirements lead towards the inclusion in the system of a molten salt burner reactor with a central flux-trap (Taube 1974). 2. Formulation of Reactor Requirements 2.1 Minimal System Doubling Time Requirements 1t was assumed at the outsef that a very-high flux reactor for fisslon product transmutation would be special, only one being bullt for every L (breeder) power reactors. Such a system of L+l reactors should provide a doubling time T <30 years; this 1s estimated to satisfy the needs of future world power development. Doubling time i1s defined by 1 GWatt thermal operating day 1 year 1.1 kg fuel fissioned 365 days 1 kg fissiles destroyed (absorbtion) InZ doubling time X X G kg fissiles net galn per e-folding time e-folding time (I+F) kg fuel fissioned 1 kg fissile fission 1 kg fissile fissioned 1+a kg fissile destroyed (abs) 1 calendar day SI kg fissile s in systen C operating day MWatt thermal )5S L . I+ 1 (I+1) Isys 1ne 1.73 (I+F)S Sys (1+u)CuSyS x 1.1 x %65 (l+a)CGSyS N where fuel = fertiles and fissiles fissiles = Pu-239, U-235 and U-233 SIi = speclfic fissiles inventory in the reactor sub- system 1 = I./P., i" 71 Ii = fissiles inventory in reactor subsystem i (inclu- ding fuel cycle). Pi = thermal power level 1n reactor type 1 I = fertile to fissile fission rate =~.2 Q = capture to fission cross section ratio ~.2 C = fraction of time at full power ~.8 i = I oI bu/Ibr bu = burner reactor subsystem br = breeder rcactor subsystem T) - “p Pbu/Pbr shen 1+G. P+ .7 LG, + B_G a - br “br bu ~bu _ br P bu sys : - L+ s Pb bu b BP T + + o _ 8ys _ ler Ibu ~ (L BI) a7 - s5ys P ) LP + P ) (L + B.) br SyS br ou P It wetting (conservatively) Gbu -1, and dropping "br" subscripts one nas Tnen, assumlng the previous nominal values for a, C, and F, the doubling time reqgulrement 15 2.16 (LB, T F < 30 Solving for X = L/BP, the ratio of thermal power in the L bree- ders to that consumed in the burner, the requirement becomes B /B, + 13.89/81 X > (13,89/8I) ¢ - 1 In Table 1 this relation is studied over reasonable ranges of values for G, SI and BI/BP' As one mlght expect, the possible variations in G produce the largest changes in the requirements on X, Furthermore, Table 1 suggests a minimum requirement that X 23, with nearterm practical requirements approaching X = 10. Table 1 Minimum "X" Requirements to Accomplish a Burner/Breeder System with 30 Years Doubling Time G = Breeding Gain = BR-1 MwWwatt thermal Advanced Current speclific fissile Breeder Breeder inventory Art Art kg fisslile in system ST BI/BP G = .4 G = .2 1 3.09 7. U3 0.8 2 3,26 7.83 3 3. 42 8.24 1 5,27 &.37 1.0 2 5,49 8,94 3 3,71 9.50 1 3,47 9,57 1.2 g 3,74 10.3% 2.2 Need for Central Flux Trap To accomplish a significant fission product transmutation rate will reqguire high neutron absorption rates. For both Sr-90 and Us=1357, the thermal absorption cross section (in barn) is one or two orders of magnitude greater than in the fast neutron region: EP o(.0253% eV),b o(fast reactor),b c(.,025% eV)/ofast Sr=50 0.8 0.0076 ~100 Cs=-137 O.11 0.0137 ' ~ 8 Consequently, the transmutation might be best accomplished in a central thermal flux trap surrounded by a fast fuel region (for high flux levels), If o¢(total spectrum, flux trap) = 1/2 « o(E = .0253% eV) the estimated total flux étot just to match (K=2) the natural decay rates is ffor Cs=1357: hgd3T 16 —2 -1 . 137 1.%*107 " n ecm s ("R = 7.35-10_103_1 o 11 . MO it ) 7.88+10 1% "1y HH H : N I ro = o O @) 3 —~ w I 2.3 Determination of Flux-Trap Size from Destruction-Production Balance Reguirement Another important requirement will be that the FP transmutation rate 1n the burner flux-trap (FT) exceed the production rate in the power and burner fuels: flux - Mtransm. Velux L N br bu trap trap ° _ dNi _ . JFP e1~10 . where Ni = 3T Pi Y 3,110 (fiss/Ws) where Ni = the number of F.P. atoms in subsystem 1. FP _ _ ,FP. FP transm, (00 + kB>FP =g using Pbu = BP Pbr and X = L/BP - get FP Y ©P NFP ____bu_ (X + 1) - 3.1_1010 . FP, FP fine K AB trap (Any additional transmutation beyond this would help to remove FP inveatory from reactors outside this system. However, for trans- mutation to be a significant improvement over tne 1000-fold decrease 1in 300 years by natural beta decay, it should then be accomplished 1n a much shorter time. A reasonable goal 1s 30 years (K=10), the lifetime of a power plant.) Assuming a central thermal flux trap in spherical geometry, the racgius 1s tnen | | T1/3 R (cm) I 3.1'1010(X+1)109 P {(gigawatts) . YFP ; A1 ux | Hm 5,023 - 1025 atoms KFPQFPA FP, - mole B trap an P . . - : where pi = density of FP nuclei (moles cm 3) 1n the flux trap reglion, 10 3. Neutronic Consideration 3.1 Burner Reactor Calculations To analyze the burner reactor, calculations were made by ANISN- code in the Su transport-theory approximation and checked in 88 with 23 energy groups and approx. 100 spatial positions (checked by 160 spatial positions). Pl—approximation cross sections were produced mostly with the GGC-3% code which utilizes ENDF/B-1 and -2 and GAM data. Sets for Cs=-137, Sr-90 and F were made in less exact fashion from AAEA and Hansen-Roach data, Fig., 1 shows the group structure,. A reference burner reactor model is shown in Figure 2, The flux trap 1s surrounded by a BeO spectrum-converter, a critical fuel thickness, and an outer wall {(see Table 2). Figure 3 shows the calculated flux distributions. The total flux in the fuel is similar to that in the flux trap. The calculated fluxes lead to the conclusion (used in section) that o(total spectrum, flux trap) = 1/2 « o(E = ,025% eV), 10 12 J 11 - 11 Pig., 1 Neutron spectrum in the core - -0 - Total mean flux H.OB'lOlO 27 Specific power 10.1 kwem Total power 7 GWth arbitrary units Hedytrom energy group: 21 19 17 15 N 15 12 11 10 9 3 ! 6 5 Lethargy Flg. 2 12 High flux burner reactor Reflector Target J ob g 100 " 2 - ] Total fluxX ———jéi T Thermal flux B \ / o (2% groups) \ /// o ‘/ _ Va7 ,/" 1// //”” *// ] Fast I'Lux " O. group .~ 4 7 (1,35 —/?.,Kfew )/ 4 ‘/' / ./ 1 ..~ / EEE 13 Table 2 High-flux burner reactor with chloride fuel Total power 7 GWtn Zone Components Neutron flux |Specific power o 24, -3 16 -2 -1 -3 Radius (cm) atom 10~ /cm 107" n em “s (kW cm ) Volume (cmB) total transmutation thermal rate (s“l) I 0 + 78.5 cm Ce=-137 00,0116 w9 Target . Sr-90 0.0016 5,83 Cs=i50 ~ 1,7°10 Vol. 2+10 cm 8] 0.0145 2,05 -8 0 2 o1kLn Sr-90 1,110 11 78.5 - 88 cm Be 0.060 %l%% Moderator 0 0,060 ’ with thin 4,365 graphite layer 0,201 IIT 58.0 - 94,56 em| Pu-23% 0.0014 tuel Pu-240 0.0004 -5 * 0 Pu-241 0.0002 4,05 L0, 1 kiem ) 0,0156 Vol.6.9-107cm” | & v.lel e Cl 0.0180 IV 94,6 - 97.6 cm| Fe 0.08 4,00 Wall 5,10"5 V ¢7.6 - 200 cm Fe 0.083 5,39 Fellector 1,8‘10_5 o, 41077 10-12 14 5.2 Moderator Reguirements To accomplish a thermal neutron flux trap one must naturally employ neutron moderating materials in and about it. As is well known, light materials can scatter neutrons past the neutron- absorbing intermediate-energy resonance region. iH 1s the most efficient nuclide in this respect but also exhibits appreciable 2D 9 172 5 1s a blt heavy though frequently already present in a molecular thermal absorption. Be and 120 are usual alternatives. 80 combination. Other light nuclides have unacceptable nuclear or physical limitations, Considering chemical and physical properties, the logical mate- rials to be used inside the flux trap are hydroxide and/or deuteroxide compounds of the FP, Figure 4 shows that just a small proportion of H molar fraction has a large deleterious effect on the Cs=-137 transmutation rate., This is due to the H absorption cross section, Therefore, CsOD and SP(OD)2 are preferred. As Sr-90 and Cs-137 also have their fair share of resonances it 1s advantageous to thermalize the flux before reaching the flux trap region containing these targets. Therefore a spectrum con- verter between flux trap and fast fuel is needed. Bearing in mind the high temperatures to be obtained in this reactor and possible chemical reactions with molten salt, HEO and D20 are unacceptable. This leaves Be, BeO and graphite or some variant therefore for consideration., Be {(and D) compounds, of course, have also to their advantage a relatively low (n,2n) threshold (1.67 MeV),. Location next to a fast region can therefore produce considerable extra slow neutrons 1n the flux trap - which is a main objective of the burner reactor. Replacement of Be by C or Mo wall material should therefore lower the FP transmutation rate, and it does (Table 3)., Figure 5 indicates an optimum thickness of about 5 cm Be. For the sake of safety and higher melting temperature, BeO 1is preferred over Re, Fig., 4 Hydrogen versus deuterium as moderator Strontium-90 and Caesium-137 activity 120 %7 7 Caesium-13 110 4 100 % = Strontium-90 ‘_ * —9 90 5 - 80 % 0 0,1 0,2 0,% Hydrogen mol ratio b (@A Table 3 ffect on Replacing Be Converter upon the Relative FP Transmutatiocn Rates ! Be | Be ! ! nmolten-salt C : Be flux | ffast driver trap Be I'e /Mo . . A . A canse materials transm, (Cs=-137) transm., (Sr-90) 1 Be, Be 1.0 1.0 2 C, Be 0.72 0.68 3 Be, Fe/lMo 0.84 0.82 (Remark: transmutation rate in arbitrary units) 17 1 H. 072 > Thickness of Dberyllium moderator zone Caeslum—-13%7 and Stront tim=90 activity 1504 Caesium-137 110 = Strontium-90 L 9 1 2 3 I 5 6 7 hickness, cm 18 5.3 Motivations for Molten Salt Fuel In a fast plutonium reactor the average fission cross seetion is estimated to be 2.5 barns. The Pu atom density is ,002 atoms cm_lb—l. The total flux is of the same magnitude as in the flux trap. The specific power for the Cs-137 burnup is then (using ¢ from section 2.2) 15 N(Pu)o_, ©/3,1+10 = 2.1 (K=1) kW Cm_5 f The rise of the thermal flux in the fast fuel (Fig. 6) leads Lo an order of magnitude increase in the fission density at the Inner fuel boundary. For a solid fuel core the resulting peak power posltion would present extreme demands upon coolant veloclty and flow distribution. To minimize thilis, the addition of boron in and about the fast fuel was considered. Reactor calculations were made for 5 to 100 micron-thick intermediate walls as well as for distributed boron inside the fast region. As seen from Figure 7 the FP transmutation rate is always reduced; although the spectrum in the fuel region 1s harder, the loss of thermal neutrons to the fiux trap has a greater effect. From the above one sees that the fuel melting point and the thermohydraulic reguirements effect a loss in FP transmutation rate for solid fuels. One solution might be to use liquid fuel, cooled out-of-core, Turbulent flow in the core will alleriate the thermal peakling problem. Also, the melting point limitation 15 removed, }‘_l 09 107 o Fission density Cission D i cnorsec 19 O Fission density in the fuel zone seryllium / / 2z F¥ssiq Gensxty o te /. Wall 130 104,06 106,6 108 111,14 20 Fig., ¢ Impact of natural boron 0.5 10t \ Actavity \\\ o . Thermal of Thermal \t\ f1ux Cs-137 |[flux in N lozb/secfue}gzgfie \\‘\ dnem™ Pu density, pPu = ,002 atomns cm_lb_l 0.9 g em”” fuel density, pf = 2.35 g em” volumetric specific heat, Cp = 1.95 Joules cm—BK temperature rise across core, At = 250 + 500 K length of channel 1in the core, Lch = 80 + 120 cm The cruclal parameter here 1s core power density. The given value is nigh but still near the present state-of-the-art (Table 4). Table 4 Feinberg, research reactor Melekeg CHM-2 FPETH Lane (cnhlorides) AFIR mean max Phicenix 250 Chilorotrans (here) kW/cm5 in the core Coclant only in the core 25 Fig, 10 Cooling of the burner reactor Total power of reactor 11 GW Heat excnanger 2,75 GW 4 units & s S S S S S S volume of fuel in tubes 0.825 m 5 24/' Z, 1 1] . 1N g :b < 17 q % # L/ % 1 4 Partial g ; ; Specific power % power 5 %% ] |1 Kwem™? o 5l 9 ; Total volume rar % =5 e ; ; # 2.75 m VO lur % 1 ¢ ‘ % 0. ; [ % L/ % “ ;)E S % L S LA TR / 1.08 m 26 Using the above one gets power rating, .07 g Pu / MWth (2.2 + l.l)'lO7 emos L H volumetric flow rate of fuel residence time in core 0.045 = 0.091 s . . -1 velocity of the fuel in the core = 26,7 + 3.8 m s The deduced fuel velocity is similar to that postulated by Lane (1969) for the HFIR reactor: 21 m s_l. (The HFIR is also a high flux irradiation facility). The specific power in the coolant is about % times higher for the burner than for the indicated breeder concept. Furthermore, 1t must be increased by BI to account for the increased fuel inventory in the burner reactor cycle. The crucial problem will be the efficiency of the external heat exchanger. In the follo- wing we use some typical heat exchanger characteristics to examine the possibilities., Specific power for heat exchanger ~1 kW/cm (rather conservative data) Total volume of heat exchangers for 11 GW(th) ~11 n° Ratlo of fuel, volumetric 0,3 Fuel volume 1n heat exchanger 3,3 m5 Fuel 1n the pipes core-heat exchanger 1,0 m5 Total fuel out of core b,3 m> Fuel in core 1,0 m3 l'otal fuel 1in whole system 5,5 m5 flean specific power of the fuel in whole system E£%—%¥— = 2,07 KW/cm3 ’ 3 Plutonium amount, in the fuel 0,8 gPu/cm 27 Plutonium inventory in whole system hoao kg Power rating in whole system 0,385 kgPu/MWth The postulated power rating for whole system 1 kgPu/MWth In this calculated case the power rating in the breeder power reactors 1,15 kgPu/MWth The above indicates tnat the achievement of a total specific power rating of about 1 kgPu/MWth may be feasible. 28 5. Further Parameters of the Burner Reactors Parametric studies were made as variations about a reference system which assumed P = 11 GWatts, (X=2.9, K=4.2) and RFT = 78.5 cm, The flux trap is surrounded by 5 cm EeQ converter, a critical fuel thickness of 6.6 cm, and an outer wall, Figure 11 shows the calcu- lated flux distributions for such a burner reactor, Note that the t.otal flux in the fuel is similar to that ir the flux trap. The calculated fluxes lead to the conclusion that o (total spectrum, flux trap) = 1/2 +* o(E=0.0253). Figure 12 shows the dependence of Cs-137 transmutation rate upon the burner power level, The higher the power, the greater the rate. However, power must, of course, be subject to thermohy- draulic restraints, such as pump and heat exchanger capabillities. Figure 13 indicates the relative effect of X upon the ratio R of FP transmutation rate to FP production rate for the reactor system, One observes the need to keep X low here. Absolute results will depend upon the Cs and Sr densities 1n the flux trap. An actual case of R = 1 for both FP nuclides was achieved at X = 4,5, The FP atom ratio there was (Cs-137)/(Sr-90) = 7.25, Another important problem is the relative high flux 1n the outer zone, the leakage from the core. This flux can be used for two purposes: 1) for transmutation of other fission products which have rather high absorption cross sectlon e.g. Tc-99 oth = 22 barn; tg 1/2 1-129 Oth = 28 barn; tp 1/2 2,1°10° a 1l 1,7-107 a 29 Radius, cm Fig, 11 Neutron flux in burner il ux 1. Power 11 GWth "%, _, 2. Power 7.5 Gwth \ 174em “s § Tnerral Ll § ‘ — — S ey — N N - % = §/ Thermal flux \\\ NN 16 ; \1 = 4/ \ 15_] V///// e " — 4 /’a" Fast neutpd R rlux ~ id 7/ (6. group) 14 7 Reflector N N : -// ii Wall 1 Efl%‘/ A Beryll u( i Oxiae s;/ L e \/ | \// Target zone _ \/1/ § */ 4 s/ /u/ M 81 86 103 for 11 GWth 73 76 92 for 7.5 GWth 30 Fig., 12 Impact of the power of burner for breeder/burner ratio ~3 Caesium=-137 activity in relativ units 124 - 120 - ~3 CO el Ne) }_l @) 11 12 13 Power, GWth 31 Fig, 15 Caesium-137 and Strontium-90 activity in the target zone Activity - Caesium o 110 Jesired level 709 Strowgium=90 - T ! | ' 1 | T 1 1 2 3 b 5 6 atlo of breeder/burner power 3 In both cases a flux of 1014-1015 n cm_2s L (see fig.ll) can result 1n a rather effective transmutation rate of Tc=-99 o = 7-10_95—1 t 1/2 1% i a I-129 o = 7-10'95'l £ 1/2 12 Mo o For improvement of this possibility use of a beryllium moderator in form of 5 cm wall in the outer region of core have been calcu- lated. This gave an improvement of transformation rate of Cs-137 in the innter target region but alsc a very significant increase of specific power, due to the softening of the neutron flux in the fuel region has been obtalined (Fig. 11). The possipility of transformation of these two longliving fission products will be further discussed elsewhere, The neutron flux outside of the core can be used for breeding in an uranium blanket., The breeding ratio can be really very good, higher than 1, but the decrease of the transmutation ratio is critical, and seems to be too low for a reasonable burner reac- tor, In spite of this, in a further study the possibilities must be checked of using part of a neutron for the breeding process in an external uranium blanket region, 53 6. Effect of Replacing Chlorire with Fluorine Is a fluoride-fuelled burrner, Instead of one with chloride-fuel, possibley (Table 5, Fig. 14, 1%) For fuel 1n tnils reactor we nave used an arpbitrarily chosen mixture of Pur, «%,54«ar-2,4 ZPFM. The calculaticons have been done for a pigger burner reactor of 11 GWth and, the whole T system totaling -h UWon, The neutron flux 1n the target region was in the case of fluoride fuel appr. 1.05 times higher than in the reference case with chloride fuel, The elfective nall-1life of both the fission product nuclides was (in years): In fluoride- In chioride~fuelled ffuelled reactor reactor (reference) Cs=137 &.57 6.973 Sr=90 175 1.83 But all these benefits must be paid by a twice higher specific 5 power: 1n [luoride-fuelled core of 19.9 kWem 7, and 10,1 chmB in criloride-contalining core. A1ls0o because the possibility to use graphite, instead of part of ceryllium oxlde as moderator, which seems to be possible from the point of view of neutronics, a rather important improvement of the coerroslicn problems can be achieved., We must remember that the com- pativility of molten fluoride fuel and graphite had been proved by the excellent experience of Oak Ridge Nat. Laboratory (with one difference: in ORNL experiments the fuel was LiF-Bel —ThFu-UFM). 2 34 Table 5 High-flux burner reactor with fluoride fuel Total power 11 GWth Power of total system 55 GWth Zone Components Neutron flux Specific power ) 24 =3 16 -2 -1 -3 Radius (cm) atom 10° cm 10 cmo s (KWem 7) Volume (cmB) total transmutation thermal rate (s_l) il 0 - 98.8 Cs-137 0.0116 _ 2 n=9 Target 6 z|Sr=90 0.0016 4,01 Cs=137 1,8 10—8 Vol.~4,1+10 7 em” |0 0.0145 —— Sr=90 1,2+10 D 0.0145 2,21 1T 38.8 - 109 em Be 0,060 5,08 Moderator 0 0.060 1,83 with thin 5. 25 3 —_— graphite layer 0,239 111 109 - 112.6 em |Pu-239 0.0017 Fuel 5 3 Pu-240 0.00042 E -3 Vol, 5.5+.107cm” |Pu-241 0,00021 TO5T 15,9 kWem Na 0.0075 2V 4r 0.0051 B 0.0340 IR 112,6-118.6 cm |Be 0.060 4,87 Wall 0 0.060 0,034 5,9 0,001 Vv 118.6-218.0 cm |Fe 0.08 0,0023 RFeflector 2,8-10“9 55 Fig. 14 Burner reactor with molten fluoride fuel Fuel: PuFB—B.SM NaF-2.41 ZrFu Specific power ~20 chm_3 Total power 11 GW Breeder/burner ratio: 4 Reryllium Reflector (Fe) 1de | | LI 96,8 109 112,6 Radlus, cm 36 Fig. 15 Burner reactor with molten fluyoride fuel Transmutation Specific rate of nucleil power - — 20 : kWem 110 = =15 100 _] Desired level =10 80 = 0 without with beryllium EeQ :one in outer in outer core core region region 7 57 Remarks about transmutation and hazard coefficients It seems to be worthwnile to make some estimation of the advantage given by this type of fission-product management. To operate for sale with only one figure for both nuclides: Sr-90 and Cs-137 the following mean values have Dbeen arbitrary postulated here: Ratio of Sr=90 Ca=137 hazard Sr=90 Cs=137 Maximum permissible (uCi/CmB) water 1-107° he1o™" 40 air 1-1077 61070 60 mean value 50 vield from fission 7 U-233% 6,2 6,6 U-23%5 5,1 5,99 Pu-23%9 2,18 6,69 mean value for power production 1:1:1 for U-23%3:U=-235:Pu-239 4,1% 6,4% relative hazard coefficlent 50 1 nazard coeff, X yield 2,050 0,064 total hazard coefficlent 2,114 in transmutation the effective half-1life 2 years 6 years hazard coeff, in transmutation 0,144 0,013 38 i 16 Inventory of Strontium-90 in the power reactors and in the burner reactor Power of total system 55 GWth Yield of Sr-90: 4,1 % Number | of Sr-40 atoms Inventory in the steady state o due to beta-decay only 1043 - (eg, In salt Ll s - . 28 ’////' 8.96+10 atoms 4 28 10 - nventory power reactors, mean " 6.6-1027 atoms Inventory in burner reactor 6.&6'1027 atoms o 10‘7 - ol 102¢ Charging, discharging Production rate of in power Sr-=90 reactors 19 . ~7+1077 atoms/sec 1092 = v | | | | 0,01 0,1 1 10 100 1000 Time, years 59 Thus 1n the steady state of transmutation the amount of hazardous substances 1s reduced by factor ~15 in relation to the steady- state of beta-decay. The amount of strontium, the most hazardous nuclide, is in a high-flux reactor burner approximately the same as in the power reactors after 3 years of fuel burning (fig. 16), But the most impressive result comes from the considerations about the "end of the fission power era". The storage (without transmutation) Caesium-137 and Strontium-90 decayed within half-1ife of 39 years, core after ~300 years the amount of these nuclide will be reduced by factor 1000 (fig. 17). In a high-flux transmutation the reduction with factor 1000 could be achlieved after approximately 26 years, (fig. 17) i.e. in the lifetime of one reactor generation (for more exact consideration see (Taube, Ligou, Bucher, 1975). Fig, 17 Hazard of the both fission products: 4o Sr-90 and Cs-137 1 Spa ~‘~_ be -~ - ~§~-~ Reactor S - hazard T 1071 Relativ i Total hazard hazard during \ transmutation LY Ts-1 Hurin _ L ransmudatio 10 ° . 7] 1 100 \ time, years - \ N 107° = \ Sr=9 durin * T transm‘fation -y \ 10 | } | I 0 10 20 30 ho 50 60 Time, years 41 8., Conclusions These consliderations show that a high flux burner reactor for the transmutation of selected most hazardous fission products as Sr-90 and Cs-137 may be achieved for reasonable assumptions: 1) the whole system 1s a steady state fransmutation system 2) 1in the steady=-state the amount of transmutated nuclides is more than one order of magnitude lower than in a spontaneously beta-decay steady-state. %) the ratio of conventional breeder power reactors to the burner is to about 4, making it possible to organize the fission power industry. )Y the whole system 1s still breeding, with a relatively long doubling time of 30 years. The relative high specific power- rating of 1 kgPu/MWth, inclusive of the amount of plutonium out of core (cooling transport, reprocessing) is rather opftimistic but not impossible, especially if a more sophisti- cated system will be used e.g. GCFBR or MSFBR (Molten Salt Fast Breeder Reactor) with a quasi-continuously pyrochemical reprocessing plant 42 9. Acknowledgement The authors wish to express their best thanks to Mr. Stepanek for helping in some data processing, to Mr, K.H. Bucher for discussing some thermo-hydraulic problems and Mr., S. Padiyath for realizing the computer calculations, b3 10. References - International Atomic Energy Agency Safe Handling of Radionuclides, Vienna, 1973 Lane J.,A, Test-reactor perspectives React. Fuel. Proces. Tech. 12, 1, 1 (1969) Lidsky L.M, Fission~Fusion Systems: Hybrid, Symbiotic and Augean Nuclear Fusion, 15, 151 (1975) Schneider K.J,, Advanced Waste Management Studies, High Level Radiocactive Waste Disposal alternatives Platt A.M, USAEC, BNWL-1900, Richland 1974 Taube M., The Transmutation of Fission Products Ligou J. , (Cs-1%7, Sr-90) in a Liquid Fuelled Fast Bucher K.H. Fission Reactor with Thermal Column EIR-Report 270, Wirenlingen 1975 Taube M. Steady~-State Burning of Fission Products in a Fast/Thermal Molten Salt Breeding Power Reactor Ann. Nucl. Sci. Engin. 7, 283 (1974) Wolkenhauer W.C. The Controlled Thermonuclear Reactor as a Fission Product Burner BNWL-4232 (1973)