EIR-Bericht Nr. 249 EIR-Bericht Nr. 249 Eidg. Institut fir Reaktorforschung Wirenlingen Schweiz A Molten Salt Fast Thermal Reactor System with no Waste M. Taube il Wurenlingen, Januar 1974 EIR-Bericht Nr. 249 A Molten Salt Fast Thermal Reactor System with no Waste M. Taube January 1974 Abstract A Power system of 1 GW(e) containing a fast breeder reactor (2GW(t)) with a molten chloride fuel and a thermal burner reac- tor (~ 0,5 GW(t)) with molten fluoride fuel resulting in a total breeding system with a doubling time of 30 years 1s des- cribed. The fast breeder works on a U-238/Pu-239 cycle, the thermal burner on Pu-23%9. The thermal burner ‘'incinerates' (by neutron transformation) most of the long lived fission products. Kr=-85, Sr-90, Te-99, Cs-135 and other. The actinides (Np, Am, Cm) might be recovered from the molten salt with a high efficiency and then burnt up in the fast reactor. The entire system pro- duces no long lived radioactive wastes with the exception of T and partially Cs-137. The fast reactor has an inherent stability against power excur- sions, and loss of coolant accidents because of the two or more independant cooling circuits and in which the fertile material (molten uranium chloride) plays the role of cooclant. A very high efficiency aluminium-chloride secondary coolant and turbine working agent is proposed. Together with a district heating system practically no thermal pollution occurs. A very low I-131 and Xe-133 and low Cs=-137 concentration, and very low CsCl volability in the steady state core due to con- tinuous gas extraction guarantees the high safety level of this system. I. Why a new reactor system? The present generation of power reactors both thermal and fast both liquid—cooied (water or sodium) and gas-cooled, use solid fuel. The single reactor type with positive experience of liquid fuel is the Molten Salt Breeder Reactor (thermal reactor with molten fluoride and with breeding ratio ~1.05) (Rosenthal, 1972). Some other molten salt fast reactors have also been dlscussed in the literature over several years (Taube, 1961, 1967, Nelson 1967). The liquid fuel molten salt reactors have some specific charac- teristics which make them more attractive for future use, espe- cially in form of a coupled system of two molten salt power reactors as 1s proposed here 1) a fast breeder reactor with molten chloride fuel 2) a thermal burner with molten fluoride fuel. This system appears to have the following special features which results in a reactor unit much more fitted to the future pattern of energy production than are the present day solid fuel reac- tors: 1) high inherent stability 2) much smaller environmental hazard %) elimination of long lived [ission product wastes and other radioactive wastes 4) better suited to district heating systems with a rela- tively high thermal efficiency for electriclity generation. A The "fast breeder - thermal burner" system proposed here has the following specific characteristics 1) inherent stability against power excursions which arise out of the following chain of events: criticality increase~ power excursion - temperature increase - density decrease » movement of part of the fuel out of the core » very strong negative influence on the criticality. 2) 1in the fast breeder reactor the loss of coolant 1s at the same time the loss of blanket which results in a very high negative coefficient of reactivity in this "Loss of Coolant Accident™. The reactor thus shuts down without an engineered scram. 3) continuous purging of the liquid fuel Iin the core by means of a hydrogen/helium stream for extraction of the vola- tile fission products especially the most hazardous I-131 and Xe-133% and also I1-137, Xe-137, the volatile precursors of Cs-137 which have a large effect on the environmental impact in any reactor accident. 4y continuous reprocessing of liquid fuel with extraction of the long lived fission products (e.g. Kr-85, Sr-90, Zr-93 Tc-99, I-129, Cs-135) and the burning (neutron transfor- mation) of them in a thermal reactor. 5) continuous reprocessing (with high yield) of the actinides (Np, Am, Cm) and continuous burning 1in the fast core. 6) the possibility of having a combined breeding ratioc for both the fast and thermal reactors coupled together, greater than 1, achieving a doubling time of approx. 30 years. 7) with the exception of tritium, virtually no nett rejec- tion of radiocactive wastes to the environment, 1f Cs-137 has been stored practically all the waste 1is transformed in the form of stable or semi-stable nuclides. 8) some of the corrosion processes of molybdenum (structural material) occuring in the core and blanket may be control- led by the continuous gas purging system. 9) the high thermal efficiency for electrical energy produc- tion of the total power system (55 % for only electrical, or 40 % for combined electrical and district heating system) results from the use of a chemically dissociating medium as secondary coolant and turbine working agent (AlClB). 10) the system would seem to be suitable for an underground and fully automatic power plant of high safety. 11) this type of reactor gives, as do other molten salt systems, a complete independance from foreign supplies (no fuel- element manufacture, no external reprocessing plant) and has the ability to 'burn-rocks' - i.e. low grade uranium "ores" from granites. (see Fig. 1) I-131, Xe-135% Heat energy . Granite Cores W fi‘\\ Granite waste ~100 tonfday preparatioij - 441-7 ~100 ton/day U, VTh Gaseous /f reprocessing Y liquid fuel N U233 0,%kg/d process and —— > (238 0,3kg/d ~ blanket fuel dP———g— preparation A Np s ¢ —1 N\ o 1000 MW(e) Fast Electricity very stable \ breeder V i%g;“St ~2000 MW (t) no thermal polution in steady Stgfifi// Heat excnan A very small ] ger < P amount of [~ coolant/A1{] 1100 MW(th) \ Cs-137 i Thermal burrier ~500 AW (L)) F1ss ) . Producks reprocessing (low temperature) ) Cs-157 angw N T storageJ stable Nuclides (with some amounts of long lived) II. The fast breeder - thermal burner coupled system The total power system of approx 1 GW(e) contains: 1) fast breeder reactor with molten plutonium-chlorides fuel cooled in core by means of fertile material (uranium-238 chloride as fertile and coolant) The thermal power 1is approx 2000 MW. In the coupled fuel cycle the breeder with a breeding ratio of ~ 1.4 form the basis of the breeding ratio of the entire system of 1.1 giving a doubling time of approx 30 years. In the fast reactor are also some of the long lived fission products being held under the neutron flux. 2) thermal burner reactor with molten plutonium fluoride fuel, cooled externally (out of core). The thermal power is approx 500 MW. In the coupled fuel cycle the burner (with breeding ratio equal to zero; no fertile material) is responsible for the burning of the longest lived fission products coming from both reactors fast and thermal that 1s from total power of 2500 MW(t). 3) an appropriate power and heat generating system with the following circuits reactor fuel - in core cooling agent (molten uranium Salt) secondary coolant out of core (aluminium trichloride) tertiary coolant (e.g. nitrogen dioxide) hot water for district heating. 4y multi-stage multicomponent reprocessing system with the following units. - reprocessing of the fast core fuel, including prepara- tion of fresh fuel - reprocessing of the blanket material from the blanket of the fast reactor - preparation of uranium chloride from natural uranium oxide - reprocessing of the thermal core fuel - reprocessing of the irradiated solid or molten long lived fission products - reprocessing of the in-core gas purge. (see Fig. 2) h:] Since the system contains both breeder and burner reactors and since a combined breeding ratio greater than 1 is required the relationship of the power of the fast breeder and thermal burner 1s the highest importance. The problem of the achievement of the breeding ratio for the total system can be solved by the following simplified calcula- tions concerning the ratio between the respective powers. Nuclides Fast/Thermal Reactors System CHEMICAL l NUCLEAR THERMAL/ELECTRICAL ~3.0kg/d 0.3%kg/d v 4 STABLE F.P. Pu Long Fiss = lived {Prod rad oac%eplflO Fessipg lHeat [€{Heat p—— J % Y — % Ex. Fuel > > JCoo1lingy > | [he rm{*< Zone Puel THERAAL] BURNER i Yo v ] e D00 MW (1) T ing urbin r_,\__,\__;fiepro ) 1C1 WO & Drainage ] < 1000MW \fi -< - —— > (e) Fast rf”’ r/”' 11 é—fl\—— Fuel Fuel @} S TRepro[Z 1> *\\\\~ (__/‘\__q\__> A K__J (:) Y 4y e Heat Heat bs—P—F: lankpt Fxchnd |kxch ot 8 i3 L L 1P %nankde i) fio, fwater [~ ! = 1| ° leccwi L____fh__n\__a - > COOLINY) | FAST HRELDLR Gas 2000 MW (L) \ prain Tank Stable Tritium Table 1 Calculation of the fast breeder/thermal burner ratio Unit Fast breeder Thermal burner Total power MW(t) X 1 Specific power gPu/MW(t) ln core Pf = 1000 Pt = 20 in total system including repro-| gPu/MW(t) Pf = 1100 Pt = 80 cessing Plutonium burning gPu/MwWd(t)| F = 1.1 F = 1.1 Breeding ratio gPu/MwWd (t) Bf = 1.35 Bt = 0 Losses 1n the re- processing system gPu/MWd(t)| V = 0.05 V. = 0.05 and uncertainties Breeding gain gPu/Mwd (t) G=F.(B-1-V) F.(1.35-1.0-0.05)=| F.(0-1.0-0.05)= = + F, 0.30 = ~F, 1.05 Postulated doubling time: T2 years/days = 30 years = 10 days leeded breeding gtotal _ yx p 4+ p 107 : f t . galn = — = 0.1 I T2 10 Effective (total breeding gain for fast/thermal system: ¢ °% - x-F 03 - F-1.05 = 0. The fast/thermal power ratio: X 1.00 + 0.1 5 E 0.5 10 For a self-breeding system containing fast breeder and thermal burner the following power distribution is needed (B = breeding ratio) ffast - fast breeder reactor (B = 1.,35) b MW(t) - thermal burner (Bther = 0) 1 MW(t) - total coupled system (BSySt 1.1) 5 MW(t) or for a total power of 2500 MW(t): - fast breeder: 2000 MW(t) - thermal burner: 500 MW(t) A very rough estimation of the characteristics of both reactors is given in Table 2. More details about the fast reactor are given in Chapter 3 and the thermal reactor in Chapter 4. Brief characteristics of the 11 Table 2 "Fast breeder/thermal burner" system Total electrical power Total efficiency (arbitrary) Total thermal power Reactor, total power Specific power (per unit volume) in core in reactor with coo- ling region but with- out blanket Core volume Cooling region volume Neutron flux Core volume composition Fuel-liguid Moderator Coolant Fission products Tubes MW(t) MW{(t)/1litre =1 GW(e) = 40 ¢ = 2.5 GW(t) Fast Breeder Thermal Burner 2000 8750 same as core T X 1015* 38.6 none 55.5 (in this case the fer- tile material) no special vo- lume for F.P. 5.9 500 1.1 homogeneous fuel; BeF2 as moderator no coolant in the core; external cooling 84 in the core Blanket Role in fuel cycle Breeding ratio Fuel components fissionable diluent moderator Irradiation target (fission products) Literature 2 47.85 (UCl3 as fer- tile“material) Breeding of Pu in U-23%8/Pu=-239 cycle ~ 1.35 Molten chloride Pu-239 + other actinides (Am, Cm NaCl Nno Zr isotopes (?2) (Taube, Ligou 1973) no blanket Burning of long lived F.P. 0 Molten fluoride Pu-23%9 Zr Fu BeF2 Kr-85, Cs-135, Sr-90, Tc-99. I-129 For thermal bree- der reactor with molten fluorides ffuel see (Rosenthal, 1972) * here for calculation purpose a neutron flux of 8 x 10 lated, however a flux of 10 n 15 1s postu- _2_. . cm S seens reallsable., 13 I1f. The Fast Breeder Reactor In this study a 2000 MW(t) fast breeder molten chlorides reactor was used. (Fig. 3 and 4) The detailed describtion of the reactor type was recently given (Taube, Ligou, 1973). The most important characteristics are summarised here: (Table 3 and 4) 1) the liquid fuel contains only plutonium chlorides diluted by other metal chlorides. The 1n core breeding ratio is realised by the presence of the uranium-2338 chloride (diluted by other metal chlorides) which acts as the coo- ling agent flowing in the tubes. 2) the fertile material (233 UClB/NaCl) is at the same time the cooling material and the blanket material. A very important and desirable feature of this fast reactor can be seen when tne "maximal design accident'" 1s discussed. Thils appears to be the Loss of Cooling Accident (LOCA) (Fig. 5) "he LOCA in fast reactor arises from two events. 1) the increase of criticality due to each coolant (even helium) being a neutron absorber and the loss of coolant results in an increace of neutron flux in the core (a trivial polint 1s that 1In a thermal water cooled reactor "thne Loss of Coolant" 1s also a "lLoss of Moderator'" whilch reduces the criticality); thus in a (ast reactor the LOCA must be accompanled by an engineered shut-down device. 14 . 3 CHLOROPHIL / EIR 2 GW(E) > Sl fi%ooc 790° AN | | AR e NV, A qulOm S;Z )W WS OVAV ) 3 | Coreo /A/ /{/* ‘ ~9g90 " C | fuel TN, KRR ninm 2.5b 15 EIR CHLOROPHIL 4 Fi AlCl 5 Blanket Material Cooling Agent 3 independent cooling circuits (including 3 blanket regions) Core (not divided) with 3 pumps Heat Exchanger SO0 4,26 m 16 Table 3 Molten Chloride Fast Breeder Reactor (MCFBR) "CHLOROPHIL" Electrical power, approx Thermal power, total/in core Core volume Specific power Core geometry Fuel: liquid PuClB/NaCl Liguidus/boiling point Fuel mean temperature Fuel volume fractlion 1in the core Power form-factors radial/axial Fast flux, mean across core Fuel density at 984 °c Heat capacity " Viscosity " Thermal conductivity, at 750 °c Fuel salt 1n core Total plutonium in core/ in system Plutonium in salt Mean plutonium specific power Mean plutonium specific power in entire system kg welght % MW(t) ke 1 ~ 800 2050/1940 8.75 220 height 2.0/ 2.36 diameter 16/18 .685/~1500 (approx) G814 0.386 0.60/0.78 ' 1015 2344 0.95 0.0217 0.007 7900 2900/3150 36 .4 0.07 0.62 17 Table 3 Coolant 1liguid U-238 ClB/NaCl Mol % 65/35 Liquidus / Boiling point °¢c ~710/~1700 Coolant temperature inlet/outlet °c 750/793 Coolant volume fraction in the 0.555 core ) Coolant density kg m™° 4010 Coolant salt in core/ in \ . blanket kg 19'500/165'000 Thermohydraulics Fuel (shell side, pumped), -1 . ms 2 velocity Coolant, velocity msml 9 Number of coolant tubes 23'000 Tubes inner/outer diameter cm 1.20/1.26 Tubes pitch cm 1.3%8 ?reedlng ratio, internal/ 0.716/1.386 votal Doubling time, load factor 1.0/0.8 years 8.5/10.5 18 Table 4 Fast breeder neutron balance Region Atoms Absorption Leakage | Production e -1021 % % yA - (n,y) 22.51 U-238 3.5629 25°5O(n,f) 2.99 8.2% - (n,y) 5.58 Pu-239 0.66796 3M.56(n,f) 28.98 85.55 _ (n,vy) 2.24 Pu-240 0.16699 3.78(n,f) 1.5y h.72 Na 6.3017 0.26 -.- [ 0% in fuel 1.10 o 1 19.495 5 in c00l.2.06 e 5.978 1.30 -.- Mo 0.7386 2.04 -.- ¥.P 0.0679 0.50 -, - Total - Corpe 71.10 27.40 98.50 - (n,y) 23.15 U-23%8 6,42 23'7O(n,f) 0.55 1.50 = @ Na 3,457 0.08 - - et E c1 20.72 .00 .- 2a) Total 26.00 2.9 1.50 Blanket 19 fi NN NN NN NN NN N NN N NN e —— fi.um o NN N\ \ \ N N (s} <~ N\ N\ Wi N /// N Z LN S N S N = N O S // < HM N N o N\ N TS AN N < F o< N N 2 e O R AN N o AD D // // | [ | N 23 B = a0 m AN // N = N \ = N N g N = // NN © N / = Q AN / // = < N AN — AN N N / / = = A = A ) o\ @ // 52 ~ N\ N\ 2 S SEIEIRIREIICHEREKL N = = h 0O R RRRREIELLRRL N CRRERIREGRENRZX S \ SRR o N N\ ~ L N N LRI pm N b 3 WN@"OMOMOWO % hflflo & N\ X X 0., g AT(TC)=950-800=150 ; NENAN AN N i m \ % ‘/|. /n el ——— ] [ N | NI N N -~ il % P (UG ER ER eI NAB M @_ - DR g E // 4/ 4/ - SN ,,/ ~ NN ~ @ v %q g e — I - .M > e ) P — D o ~ ¢ ) — : ‘/ N\fi P ~aam— ——— “ N N % N N // VA_ AN t ‘j TS b,/ /\h v N N ™y N N Y ™~ ! ~ A\ X e —— = | o~ 1 AN X — R q \VU ALY AY N ) - A - ] ‘ “ N N //// // // N // //v/// s - AN AN NANANANANANAN AN N NNANENANAYANANANANAN // N < — // < T < < < < T < S \ N T ,V/ A‘O\/ RN NG NN N h SN N VA N \ - —— ' q/ N N « N // // N N // N N, ~ N < f/ N // N N N N // // N N // // // I/ // I/ // ~ // // N // q N N NN NN, N N N N N NN N N RES NN N, N N N N N N N NN NN N S N N N N NN All N NN N\ N\ N\ AN NN N NN N\ N N N \ NN N AN \ \ N NN\ //r BeTdoY SUTTOL— 26 Table 5 First approximations of the thermal burner reactor neutron balance Total power: 500 MW Specific power: MW/1 Volume: 250 litre Thermal neutron flux: wth = 6 X 1015 n cmmzs_l 2 X 106 p) 3 2 MW/1 gives ~ = 2 x 107 W/cm 1077 Fisslon rate 2 X 103 Xx 3.1 x 1010 fiss/W = 6.2 x lOl3 fiss/cm5 . PU-239 6.2 x 10°° 6.2 x 1000 Plutonium atoms: N 3 —>> T G el cm (7.2 x 10 )x(6 x 1077) 4.3 x .0 = 1.45 x 1019 Atom Pu/cm3 For 250 litres: NPu—239 (1.45 x 1019)x(2.5 X 105) = 3.6 X 102” = tot = 6 Mol Pu-239 Table 5 Amount in Cross Neutron Volume Fuel steady state Section Balance (litre) (mol/in core) (octherm,barn) Absorption Production 1 Pu-239/241 6 Mol (80%) Of = 720 4320 36 100 o = 250 1500 | 12.5 Pu-240/242 1.2 Mol (20%) GC = 250 300 2.5 Irradiated Fission products 11 Sr-50 350 Mol 1.20 420 3.5 2.5 Tc-99 57 Mol 22 .0 1260 10.5 7 Cs—-1535 185 Mol 8.7 1600 13.3 1-129 7 Mol A35.0 245 2.0 0.5 : I-127 14 Mol 6.0 85 0.7 30 Kr-85 68 Mol 1. 120 1.0 Diluent /Moderator BeF2 5100 Mol 0.01 150 1.25 ZrFu 1100 Mol 0.03 100 0.8 Structure 15 (tubes 7Zr) 1100 Mol 0.3 300 2.5 250 litre Leakage 1600 13.3 Totals 12000] 100% 100 Lc Fig. 8 coolant ‘ pump coolant D < s 28 Fast/thermal system (G coolant productecy G D | Pump fuel (therm) pump ‘F&ssion T fuel 0 B Rl e :1:17 R A -} ! < e 2 AN \C\\\\\\\\ 5 AR TRNI NN AN N oo \\ AT AN 7-L_{,;Y- /8/ \\\ )}fiezrlii '/;y/ \\\\\\ 1 ‘D’l”,& 11 &) \\ 1 AV N ) f p ~§\ ‘/[ U £ N coolant - (e AN N R NN N \\\\ Fagt NN A W Nase NIN NGOV NN ) SN NN | SN NN NN \\\_ Blanket o ) © >~ E‘ ——{E; © 5 [ 29 V. The in-core continuous gas purging V.a) General remarks In this reactor an in-core continuous gas purging of the molten fuel is postulated which significantly improves the safety of this reactor 1n an 1in-core accident. A mixture of hydrogen-helium gas is continously bubbled through the liquid fuel in the core. The mean dwell time of the gas-bubb- les needs to be controlled and the mean transport time of the molten components to these bubbles must also be controlled (e.g. if speed-up 1is desired-intensive mixing, if delay-local addition of a further gas stream). The aim of the gas stripping is as follows: 1) 2) 3) i) to remove the volatile fission products which in the case of an accident control the environmental hazard. (I-131, Xe-133%, Kr-85 and precusors of Cs-137 and at the same time for the thermal reactor, removal of the I-135, pre- cursor of Xe-135, improves the neutron balance. to control the production of delayed neutrons since most of the precursors and nuclides of this group are very volatile, e.g.: Br-I-isotopes, removal of oxygen and sulphur, continously (see Chapter VIII) in situ control of corrosion problems on structural ma- terials (see Chapter VIII). For the sake of a first approximation a gas flux of 30 cm3 per sec. (normal state) of H2/He is arbitrarily assumed. At 20 bar pressure 30 and with a dwelling time in core of 20 seconds, the gas bubbles 5 will only occupy a fraction of the core equal to 10 - of its volu- me and has little influence on the criticality, (but the collap- sing of bubbles results in a positiv criticality coefficient). The system proposed for continuous removal of the volatile fission product from the core itself has a retention time of some hundreds cf seconds only. Each reprocessing mechanlism which operates out of core is limited by the amount of molten fuel being pumped from the core to the reprocessing plant. This amount, due to the high capital cost of the fuel and high operation costs cannot be greater than that which gives a fuel in-core dwell time of about one week. Even with a 1 day dwell time, that is, 1f a after one day the fuel goes through the reprocessing plant, no acceptable solution to the I-131 problem is obtained since the activity of this nuclide 1s only diminished by one order of magnitude (see Fig. 9). Only a direct in-core removal gives the dwell time in core as low as some hindreds of seconds. V.b) Delayed neutrons emitters The principal question arise out of the fact that some of the short lived iodine and bromine (perhaps also arsenic, tellurium) 1sotopes are the precursors of the delayed neutrons. 51 Table 6 Precursors of delayed neutrons for Pu-239 fast fission Group Half 1life Firaction ! Probable t 1/2 (seconds) % i Nuclide 1 52.75 3.8 Br-87 2 22.79 28.0 I-137, Br-86 3 5.19 21.6 I1-138, Br-89 I 2.09 32.8 T-139, Br-90 5 0.549 10.3 6 0.216 3.5 As other possible nuclides the following can be considered: As-85; Kr-92,-93; Rb-92,-93,-94; Sr-97,-98; Te-136,-137; Cs-142,-143. The removal of these delayed-neutron precursors from the core re- duces the value of B, which is lower for Pu-239 than U-235. Thus we have a problem of reaching a compromlse between a removal as rapid as possible of the hazardous I-131, and as long a dwell time in the core for the delayed neutron precursors: I-140, I-13%9, I-1%8, I-137 and the appropriate bromine isotopes. The compromise is given by a quasi-steady state regime in the core in which the amounts of isotopes present are given as an example in Fig. 10. 32 Fig. 9 I-131 in CHLOROPHIL I-131 Radioactive decay only 10 10 10 Extraction O Extraction 10-3 ) -2 Extraction 10 Farmers safe level of I-131 1 min 10 min 1lh 10h hq hoa l¥ 1Qy | | | | | ] | | [ lu T l6 ] |8 10 lO2 lO3 10 lO5 10 107 10 Time, second 33 Fig. 10 Removal of I-131 in CHLOROPHIL I and the problem of the precursors of the delayed Neutrons 2 GW = 6.2 x 1019 fiss/sec. I-138 (precusor) Production Rate t 1/2 = 5s A = 0.14 s . N * 69 Atoms/s 1%o 3 y/ H2 extractio / p ’ &L Y _3 a Ex = 10 x10 = Tx10 At /s = 0.7% of I-13 Neutron loss = 0.7% I-131 P = 1xlOl8 Atom/sec. . = 101) dis/sec. a = 30k Curi ~— /3/{> 3 urie One thousand times smallel I-131 activity than in stationary case without purging 2 1.65 mili Mol 2 Without extraction: after approx 2 months 50 Mega curies I-131 (lxlOl8 dis/sec) 34 In this case the mean dwell time of iodine in the steady state reactor is about 100 seconds. From the curve in Fig. 9 1t can be seen that the activity of iodine for a 2.5 GW(t) reactor is of the order of only 10 k curies (for 100 seconds) instead of ~ 100 M curies in the steady state, a decrease of activity of 5 approx. ].OLl (or 10° for 1000 second extraction rate). V. ¢) Gas-extraction and other fission products The gas-extraction also influence the other volatile nuclides.,. From a very rough estimation for these molten salts (with a small excess of free hydrogen) the following fission products and their associated precursors of ilodine and bromine can be volatile at ~ 1000 °c. In elementary form: Xe, Kr, Te (2) In simple volatile hydrides: Bri, IH In simple volatile chlorides: Sang, SbClB, NbClB, CdClg. This amount of finally veoclatile components including 1, Br, Xe, Kr amounts to approximately halfl the total fission products (i.e. 100 micromoles per second). In addition there 1s tne corresponding amount of tritium (from ternary fission). This amount of all fis- 5/s or 10 times smaller than the postulated amount of hydrogen f{low at 30 cm5 197%; Beattie, 1973). sion products corresponds to a gas volume ratic of about 2 cm /s . For volatile fission products release: see (I'armer, 55 The extraction removes all short 1lived fission products which are volatile under these conditions. Thus not only is the removal of the iodine isotopes and the consequent reduction 1n production of xenon (e.g. for the atom number: A = 135, 136, 137, 138, 139) achieved but it slows down the in-core production of Cs-137, Cs-138, Cs-139 and then also Barium-139 (Fig. 11). The gas extraction will also cause the vaporization of the higher components of the liquid fuel: PuCl3 and NaCl. The fuel consists of: - 15 mol% PuCl boiling point 2040°K 33 -~ 85 mol% WaCl ; boiling point 1738°k (see Fig. 12). One can as a first approximation say that it would have to follo- wing composition in the wvapour phase: 5 mol% PuCl, - 95 mol% NaCl. 5 The order of magnitude of vapour for pure components at a tempera- ture of about 1250 K is NaCl ~5 x 107° bar; PuCl, ~10 ' bar For the PuClB~WaCl system one assumes here a lowering of the vapour pressure (thermodynamic activity coefficient approx 0.1). 5 At the postulated volumetric flow rate of 30 cm H2 normal) per second, the vapourized amount of plutonium is given by: 30 cm3 . 10-4 bar - 10_1 = 10_8 mol Pu/s 5 22000 cm”/mol A Mass No. 132 133 134 135 - 136 — 137 - 138 - 139 — 36 FPig. 11 Precursors of delayed neutrons C) Radioactive Nuclide Volatile elements in this <> Precursor of delayed neutrons fuel N\ N @® stable nuclide - B decay (V,Probabl‘/ volatile direction T Sb 131 sn N, L) | | ] <10 102 103 10% =105 Half life t 1/2 seconds Pressure mm Hg 760 — 10 very volatile S| ’—l- 0Q noble gases 37 12 Vapour pressure of Fission Products and Fuel compenents (in pure state) Fuel Temp in + the core The main fission products are circled The main fuel components 100 i | ! T | T T T T T 1000 1500 2000 Temperature K 38 This amount of plutonium is of the order of lO—u relative to the amount of plutonium fissioned in the same time (approx 10-4 mol Pu/s). However, it still has to be recovered which makes the reprocessing unfortunately more complicated. Last but not least is the in-core gas extraction of two other elements - oxygen in the form of H20 : oxygen from impurities (i.e. PuOC1l) - sulphur in the form of st: sulphur from the nuclear reac- tion: 35 354 Cl (n,p) (see chapter VII) VI. The neutron transformation of the fission products VI. Principle of neutron transformations Some of long lived radioactive nuclides are shown in Fig. 13 and 14, The aim of this part of the paper is firstly to discuss the possi- pbility of the steady state nuclear transformation (steady state burning) in the thermal neutron region, of the larger part of the long lived fission products. A first approximation for the solutions of this problem is given here and a further discussion initiated. 39 Fig. 13 Long lived radioactive wastes Mega Curies for USA Mega Curies for O Burning in CHLOROPHIL I/EIR Europe A ¥ Burning in CHLOROPHIL II/EIR 10 —}03 * Nuclide which cannot be burnt here s-137 103—\ ¥ 10 S Vg Cm-244 \ =) (2] 3. o« ¥ Activity - Mega Curies A\ 014 cs-135 % l | T I 0 100 200 300 400 Time from unloading from Reactor (at the year 2000) Potential Hazard Index 4o Fig. 14 Postulated hazard index for long lived fission products and actinides 10%__ Ingestion 10fl__ Sr-90 I 0 10 100 200 Pu-239 19 107 Am 241, oh2 Inhalation Pu-288 T I I 0 10 100 200 According to 41 The idea of destroying the beta-active long lived radionuclides is based on the following: A g~ AE FP = Fission product, ZFP —5- (Z+1) (stable) . primary tl/2 ~10 years FP = Fisslion product after irradiation (n,y) neutron irradiation in high A = Atomic mass flux reactor = Atomlic number Y E = Stable nuclide (A+1) . (A+1)E FP - (stable) 2 tl/2 very short (2+1) In Fig. 15 are given the nuclide transformations under a high fflux discussed here. This problem of nuclear transformation of fission and non-fissio- nable actinides, being the irradiation products, has been discussed for many years. The most recent remarks and calculatlions con- cerning this problem made the following claim: "It 1s not practi- cal to burn fission product wastes in power reactors because the reutron fluxes are too low. Developing special burner reactors Wwith the required (thermal) neutron flux of the order of I 10 n oecm s or burning in the blankets of thermonuclear reac- tors 1s beyond the limits of current technology. It seems reaso- nable to extract tnree actinides (Np, Am, Cm, (M.T.)) for separate storage or for recycling through the reactors that produce them" (Claiborne, 1972). 42 Also a controlled thermonuclear reactor CTR was used for calcu- lation of neutron transformations. In those calculations a neutron flux of 5 x lOl5n c:m.-gs_l was used with remark that this flux level 1s somewhat higher than that usually associated with CTR power plants (Wolkenhauer, 1972, 1973). A three year irradiation was proposed but it would seem that for the most important fission products the neutron build-up in relation to the radiocactive decay plays a relatively small role; e.g. for Strontium-90 after 3 years irradiation all neutron reaction products from (n,y) and (n, 2n) are lower than 10% of those from the radicactive decay products Sr-90 and for Cs-137 only (n, 2n) products result in ~ L0% of the radiocactive decay products. But (n, 2n) reactions need a rather hard neutron spectrum (Gore, Leonard, 1973). VI. b Cross sections and neutron fluxes for nuclear transformations In this system the following simple evaluations are made 1) the amount of fission pr-ducts coming from both the fast breeder (2000 MW) and the thermal burner (500 MW) 2) the cores are purged by means of gas (He + H2 stream) for ecvacuation of some volatile nuclides (Kr, Xe, Br, 1) 3 %) the fuels and blanket materials are continually reprocessed ) the irradiated fission products are continuously {(or perio- dically) separated for the sake of eliminating the daughter stable nuclides (e.g. “r-90 and Zr-91 rfrom the decay and burning of 3Sr-90) 5) the amounts C in steady state (ss) burning are calculated by the obvious relationship for the i-th nuclide. 43 Fig. 15 Burning of selected Nuclides 140_ Cs-135/137 139 - 94— Sr-3g0 2x10 y _ Zr-93 138 : 93 32m 137 ° 92- 136 91 — 135 90— La Ce Atomic Number Z Sr Y Zr Nb Mo 7 | ] i | | | ] | | 55 56 57 58 38 39 40 IA 42 103 - ®Long lived fission product 1024 @ sStavle Huclide OBeta unstable intermediate nuclide 101 A , . . @Long lived nuclide 100 - 99 — T¢ Ry Rh L i px(3.1 x 1010) fiss.s_lw_l X ni C = (mol) %) (1 i 2 + o.9.) x (6 x 10 R 3 J) ( = thermal power (watts) = yield (relative amount per 100 fissioned Pu-239/241 atoms) radioactive decay constant (s-l) = c¢ross section (cm_g) = mean neutron flux (n cm_zs—l) o Q > 3w " appropriate neutron spectrum (j = fast or thermal) . il The most critical values 0§®3 are given in Fig. 16 for some selected nuclides for thermal and fast neutron fluxes. The calculation given above results in the following: Burning in a thermal high flux power reactor # = 6 x 1015 n cm_gs”l Cs - 1535 Sr - 90 (FPig. 17) Kr - 85 Te - 99 I - 129/127 (Fig. 18) ) . _ 16 -2 -1 Burning in fast power breeder reactor: @ = ~ 10" n.cm s Cs = 137 (Fig. 17/1) (other option: only storage but of core!) Zr - 93 (Fig. 18/II1) - 2 CTIONS, BARN (10 2ucm ) OSH-SECTIHE CR NonfIssionable THERMAL FLUX 17 2 5 18 1016 10 Neutron Flux (n cm_2s-l) 46 A more complicated problem is the need for the separation of the two caesium isotopes Cs-135 for thermal burning and Cs-137 for fast burning or out of core storage. From Fig. 11 it can be seen that the in-core gas extraction takes out the precursors of Cs-135, that is the long lived volatile I-135 and Xe-135, but the short lived precursors of Cs-137 remain in the fuel, so the sepa- ration of these isotopes 1s possible (see Chapter VII), Among the most dangerous long lived radioactive wastes are some of the actinides (see Fig. 13, 14). The problem of the "destruction" of these nuclides: Np-237, Pu-238, Am-241, Am-243, Cm=-244, Cm-245, is ‘simple’. In any high flux fast or thermal reactor because of the relatively large cross sections these actinides are burned up, the problem is only limited by the present state of the reprocessing technology of irradiated fuel. In the most obvious agueous processes e.g. the aqueous phase - organic phase extraction, these elements: Np, Am, Cm being hydrolysed, because of a relatively low concen- tration, and in the hydrolysed form are not extractable by many organic extractants. Of course a more refined aqueous process could give a significant improvement in the recovery of these actinides making it possible to recycle in the fuel and eventual- ly to burn-up (Fig. 20 and 21). In the fused salt fast-thermal system proposed here the recovery of the actinides seems much easier and simpler due to the fact that 1) the reprocessing occurs not in an aqueous but in a molten salt medium; the hydrolysis of the low concentratilon actinides is not possible; the recovery 1s very high. MOL IN CORE MOL IN CORE b7 Fig. 17 _— Only radioactive decay —_—— Burning in the Reactors 1(Cs 137) 1260 Mol, Amounts in steady-state Fast Flux V = Volume 1n steady-state, Thermal flux Oth = 8.7 barn ! ¥ ] T | .1 1 10 32 100 1000 1 10 32 100 1000 YEARS YEARS I1(Sr 90) 102 4 1V (Kr 85) Thermal flux Thermal flux L. 1,2 barn " ot - 1,8 barn 10 I T I | 1 T | ] | 1 1 1 10 32 100 1000 1 10 32 100 1000 YEARS YEARS MOL IN CORE CORE MOL IN 10 10 YEARS Fig. 18 J1(Tc99) 102 1 11(1129,1127) Thermal Flux Thermal Flux oth = 22 barn - 1014 N 23000 Mol 4700 Mol in 100 years in 100 years 3 3] - mlO - o (@] E - =2 010 - 10 4 - | T ] T | S T 1 0.1 1 10 32 100 1000 1 10 32 100 1000 YEARS YEARS N e O BETA UNSTABLE 11(Zr 93) I;i/ . )@ LONG-LIVING 7 (Zr ‘ . Past Flux E‘] BETA-STABLE Of - 100 mb SEPARATION POSSIBLE (2) T 4200 Mol C) CJ [] ————— A - 2 - O—-O0—0—0—L{ O + | ~o—o0—O = _ = < _.O-—‘Q———D < - S—0—O T T | ! ] ! T T T 1 1 1 10 32 100 1000 Sr Y zr Nb Mo ATOMIC NUMBER INCORE MOL, 10 10 10 10 10 Fig. 19 REACTOR: 49 2,5 GW(t). ao— FISSION-PRODUCTS IN STEADY-STATE POWER Cs 137, fast Sr 90, thermal Cs 135, thermal 57 Mol Te 99, thermal / S o A// I I 129, thermal //’;;7 o ~ d A 4 A // S P T o wu o) £ PERN Y = . o A O e © 3 » 0 v so) 0 | I | 1 1 1 10 32 o 100 1000 TIME OF IRRADIATION (YEARS) 50 i i i 244Cm Fig. 20 Burning of the long lived alpha . emitters dc = 0.4 70 Mol 243 rast . gt2%Y = 8x10 5ncm 242 241 — 240- @ 800 Mol 5 e 0.33 B 8~ 239 @ @ % 6000 Mol se /T (n,y) S 0.21 . 238 B 238 U @ @ 100 Mol (n,pn) 51 2.0 h = o Oj (n, ()) 1or1 X ot (D @ U Np Pu Am MOL IN CORE 10 10 10 10 10 51 Pig. 21 // O o /o / ¥ Y ) B Y S | QO ) b /// o d L P ju o] o wn q [ [ & o (ORI Q e [ gV % h ] | 10 32 o4 100 TIME OF IRRADIATION (YEARS) 1000 steady-state core in 52 2) the continuous remotely operated extraction processes and preparation of the liquid fuel, and the absence of fuel elements etc. make this recycling of the actinides (very hazardous alpha and partly neutron emitters) possible and desirable. The high grade recovery of the actinides postulated here gives the steady state results for the fast reactor shown in Fig. 20 and 21. The reason for burning in the fast and not in the ther- mal reactor is based on the fact that the relation ég/dzh is greater in the fast reactor than in the thermal. It must be remembered that the relatively high neutron flux results not only in the irradiation of the non desirable radio- active nuclides but also the stable fission products, not to mention the stable components of the fuel and the structural material. A very short review of these partial processes 1is given here 1) chlorine contains two stable isotopes (underlined) (see Fig. 22). - 3 559l (n,Yy) 3501 —E—m-36§£ volatile 17 17 years 108 [ - 38 ey () o AT 12 12 but not only (n,y) reactions are important here. Much more impor- tant is the following reaction: 559l (n,p) 358 B o BSQl 17 16 8.7 days 17 40 - = Fig. 22 Chlorine Burn-up é:éHazardous Nuclide 39 — table Nuclide Eventualy enriched 38 — chlorine isotope and product. 37 1 Cross section for fission neutrons millibarns 36 — (n,p) (n,a) 35 Cl1 1.6 ) 37 Cl 0.24 ? 35 - 34 — Cl(natural) Cl(natural) 0.7 mb 33__ A5 C1 0.7 mb 37 C1 0.03% mb 32 31 — P S Cl Ar K l | l 1 L 30 | r | | t 7 15 16 17 18 9 54 The presence of sulphur influences the chemistry of the molten fuel. 2) Sodium has only one stable isotope Na (n,v) 2Ma Lo 2“@_5_ (n,y) 11 11 12 2 26 51\_4_% (nsY) Mg = —= Mg (n,y) ‘Mg -£= °Tn This chaln seems to result in increasing amounts of stable magnesium and in longer time also stable aluminium. A short summary of the "history" of the long lived radioactive nuclides proposed in this paper is given in Table 7. Table 7T Nuclide Sr-90 Zr-93 Tc-99 I-129 Cs—-135 Cs=137 Actinides 55 Problems in fission-products management Separation from in core gas purging diluted with H from in core gas purging after decay of other Kr isotopes: from fuel reprocessing: separation Sr/Ba needed from fuel reprocessing: the pre- liminary separation of Y-GQ1, precursor of Zr-91 i1s needed Mixture of Zr-93, 94, g5 from fuel reprocessing from gas purging, after decay time of I-131, that 1s ~ 160 days from gas purging, after the decay of Xe-13%5, that is after ~ 5 days from fuel reprocessing: separa- tion of Cs/Rb 1is needed from fuel reprocessing together with plutonium; very high effi- ciency of separation processes is needed. Transformation not possible, waste only burning in thermal reactor (under pressure or as fluoride com- pend ) burning in thermal reactor burning in fast reac- tor; not possible 1n thermal reactor due to too large a volume. The radioactive Zr-93 is diluted by Zr-94, 95 partially used as a dilutent for the thermal reactor 7ZrP or as structural ma- terial & fission pro- ducts cladding in thermal reactor. burned in thermal reactor burning in thermal reactor burning in thermal reactor quasi-burning (storage) In fast breeder core- (In solid form) or out of core fissioning 1in fast reactor VII. Reprocessing 56 The system proposed here is a rather complicated and sophisti- cated 'chemical machine'! and not only a nuclear energy thermal energy transformer as are other classical reactors. The entire reprocessing scheme includes: Table 8 Continuou§ Process Gas extraction Fast chloride fuel reprocessing Thermal fluoride fuel reprocessing Fertile chloride, material reprocessing Fission products, mostly fluoride reprocessing Scheme of entire reprocessing system Short description Hydrogen-helium gas stream Reprocessing of the 2 GW(t) fast breeder Remark: only plutonium! No uranium! Reprogessing of the 0.5 GW(t) thermal only plu- tonium! No uranium! Reprocessing of the fast breeder blanket material Mostly uranium, small amounts of plutonium. Reprocessing of the- long lived fission products, in the thermal high- flux. (Fig. 23) Principle action irradiated in Secondary removal of I-131 and Xe separation of Cs=-135/Cs-137 removal of stable nuclides from Zr 57 The effective burning of the most important long lived nuclides needs a rather complicated reprocessing plant with high tempera- ture pyrochemical separation methods, both in gaseocus and liquid salt or/and metallic phases. In some cases also a retention volume is necessary (for spontaneus decay of short-lived nuclides). This reprocessing philosophy arises from the following assumptions: The caesium nuclides are very hazardous but there are at least two isotopes Cs-135 (t 1/2 = 2 x lO6a) and Cs-137 (t 1/2 2 30 a). The first one has a reasonable cross section from the point of view of neutron transformation (Oth = 8 barn), the second a very small cross section in both fast and thermal reactors of = 100 mb or less, oth = 110 mb). The separation of two flows e.g. Cs-135 for thermal burning and Cs-137 for quasi burning (rather self- decaying) could help solve this problem. In this system the continuous in core gas purgling system is expec- ted to extract some of the volatile fission products especlally those with. longer half life period. In the case of caesium the separation of both isotopes is possible due to the different time chains 20.8 h 9.1 h 2 X 106 year I-13%5 - Xe-155 - Cs=-135 - 1137 _ 2S5 xe-137 _ 2% ™ (s5-137 . 30 years ——— The extractable volatile and relatively long lived nuclides 1in short time gas purging are underlined. The Cs-135 could be extracted in the form of the precursors. Cs-137 1is retained in the fuel and must be separated by more conventional (but not so simple) methods. 58 Another similar problem arises in the case of zirconium isotopes (see Fig. 18/IV) and iodine (Fig. 18/II). Part of the zirconium could be used as a diluent of the homogenous fuel for the thermal core (e.g. 20 mol% ZPFM + 80 mol% FeF2) part as component of the molten fuel for the fast core (?) and part in metallic form as structural (but radioactive) material when in the final period of fast reactor development when the Z2r-93 will be reduced to less than 1 mass % of the total zirconlium Some further similar questions will have to be discussed, in some a separation of isotopes is possible by means of rapid in core gas extraction or slow out of core (liquid phase) separation process. A preliminary draft of the possible reprocessing scheme is given in Fig. 23. VIII. Molybdenum and corrosion processes Tne corroslon processes in this type of reactor are of crucial importance. In the cores of these reactors are approx 950 kg molybdenum or approx 10'000 mol. We postulate here, arbitrarily, a relatively high corrosion rate equivalent to a full or 100% corrosion of the complete amount of molybdenum 1in approx 10 years. This means a corrosion rate of: 10'000 mol molybdenum 5 = 3 x 10 ° mol/s = 450 um/s 10 years x (3%.1 x 1O7s/year) Pu 59 Fig. 23 F,P. = FPFISSION PRODUCTS REPRO = REPROCESSING SEPA = SEPARATION "THERMAL" FUEL . ~ ) - AN REP HERMAL, Tc-99 UE Sr-9p — AN > P 2 L N > \ | THERMAq | CORE | = rHERMAL | | > |[BURNER | . i Fiss l I I \ Prod e | l \\\:—/ r—flr - BLANKET l v \ ) Kr,I,5r,Tc,Cs [ o \\_,_/ N2 AN ) - N\ > N— - L LANK - — MATERT - > A > v SS‘J.B? s ZP_9 ~ N Pu ~ 4 +ACTINIDES A v URANIUM PLUTONIUM (DEPLETED) FOR SALE TRITIUM STABLE NUCLIDES Br Xe Te (?) Xe- Xe-130 Zr-91 Ru-100 Rb-86 Ba-13%6 Zr-91 RARE EARTH NOBLE METALS SEMINOBLE METALS OTHER ELEMENTS ALKALI AND ALKALI EARTH METALS ) -1 of formation (KJ.mol energy Free -50 -100+ -150- -200+ 60 MoC | 1000 Temperature K 61 Thus 30 micromoles molybdenum per second (or approx 3% mg Mo/s). In addition to this we have molybdenum as fission product. In the 2000 MW(t) fast reactor the amount of fissioned plutonium equals: 2000 MW(t) x 1 day x (1.06 g Pu/MWd(t)) % 2150 g Pu/day or: 25 mg Pu/s: that is; 100 yu mol Pu/s. The production side of fission products equals 200 u mol/s. Molybdenum is present at about 18 mol% of fission products which means a 36 u mol/s, or almost half the amount of corroding ma- terial. From both sources, corrosion and fissioning the amount of molybdenum is about 50 u mol/s. The corrosion processes can go in three directions 1) Chloride reaction Mo + 2 C1 + Me+ +~ MoCl. + Me met 2 met 2) Oxygen reaction Momet + 02 > M002 3) Tellurium reaction Momet + Te2 -> MoTe2 These and eventually other molybdenum corrosion processes are potentially dangerous and should when possible be prevented in the core itself. This possibility exists due to the reaction with the gaseous hydrogen (see also Chapter IV). 1) MOCl2 + H2 > Momet + 2HCL 2) M002 + H2 - Momet + 2H20 3) MoTe2 + 2H2 - Momet + 2T€H2 - Te2 + 2H2 (see Fig. 24) 62 Penneability Solubility and 26 met gas wae Wo*ww (pJIBPUBAS) WD 2/T- 3 Hlflml (pJIBPUEBIS) ¢ d — 7.. Q_.. ! = = o = l | Temperature 7___4 J._ /.__._ R_u_ o o o o - . | - uge woqly oy /woqBR H A3TTTANTOS ¢/ T~ 63 In the extreme case the molar amount of hydrogen is twice as big as that of molybdenum giving 120 u mol H2/S. Under normal conditions (1 bar, 0°c) this amount of hydrogen is acceptable: ~120 umol x (2.2 X ZLOLl cmB/mol) X 10_6 = 3.0 em For further calculations we arbitrarily assume ten times more, 3 approx 30 cm Hz/s. (see also Chapter V) Here is a major question concerning the reaction of metallic molybdenum with hydrogen. From about 100 elements there are only twelve elements which form no chemical binding with hydrogen. c’lLl d5 d6 d7 d8 5d - Electrons Mn Fe Co ia - Electrons Mo Te Ru Rh 5d - Electrons W Re Os Ir Pt Thus molybdenum fulfills the minimum requirements: 1t has no reaction with hydrogen. Also manganese-cobalt alloys are suitable for this reactor (chromium and nickel not!). In addition the solubility of hydrogen in molybdenum is relatively small. The dependence of the solubility of H2 in Mo-metal on tem- perature and pressure is known since molybdenum is a possible constructional material for containing plasma (hydrogen) in fusion reactors. Also the diffusion of hydrogen through metallic molyb- denum 1s relatively very small (see Fig. 26). {form oU Fig. 25 CHLORIDES-OXIDES EQUILIBRIUM DIAGRAM AT 1000K Ba Cs . - 400 °® CHLORIDES MUCH MORE STABLE THAN OXIDES Ca CHLORIDES IN FQUILIBRIUM WITH OXIDES Na @ ‘300‘~ Pr () [ ¥ Zr e U Pu -2004 T OXIDES MUCH MORI @ Al STABLE THAN CHLORIDES Cr .Sl ' il \ . il Feg . ra -100 1 ta ¢ ® Mo T ] T T T I T -100 -200 - 300 aGEOUOE oY TDES (KJ.1/2 mol 10) form 65 The diffusion rate of hydrogen from the melted fuel to coolant and blanket (here also UCl, - NaCl) needs to be discussed. b One can assume, however, that also this melt with hydrogen is saturated so that the porosity of the wall (molybdenum) will play a minor role. The most important: the variation in mechani- cal properties of the molybdenum caused by the dissolved hydro- gen. The problem of the corrosion of molybdenum in chlorine containing media is particularly complicated by the numerous moclybdenum i : M . chlorides 0012, MoClB, MoClu, M0015 Also the oxXygen chlorine systems for molybdenum and some selec- ted fuel components including hydrogen are rather complex (Fig. 24). In the region of the external heat exchanger the main corrosion process results from the action of gaseous aluminium chloride (the secondary working agent) AlCl3 + % Mo - % MoCl2X + AlCl( ) () () & This reaction has been discussed in earlier publications (Blander 1957) but unfortunately not all the thermodynamic data 1s known, The stability of the molybdenum chlorines 1is of course strongly influenced by the concentration of free chlorine and also by the temperature,. 66 Fig. 27 Iron and Molybdenum burn-up (Core materials 60 @ — b0, oy B H9 Stable Nuclide 59__ Co 454 (:> Beta- Unstable Nuclide 58 ] Fe 58 Al T&A .19% 57 ‘ Iron main reaction VA Molybdenum A = 10712571 n = fission yield Z . — — 952r 9)Nb DS Mg 67 A more detailed calculation of metallic molybdenum corrosion in the aluminium trichloride is needed. These calculations are very sensitive to the vapour pressure of chlorides. More detailed calculations of the corrosion in this system have been given earlier (Taube, Schumacher 1969). IX. Molybdenum and iron irradiated in a fast high flux reactor IX a) Nuclear effects The high neutron flux irradiation causes physical and chemical changes in structural materials. Molybdenum 1s a mixture of stable isotopes (Fig. 27). The most important by-product of neutron irradiation is the Tc-99 beta- emitter with t 1/2 2 2.1 x lO5 year and which belongs to the chain (stable nuclides underlined) Mo-98 (n,y) Mo-99 B o me-g9 B — 239 66 n 2.1 x 10° a (n,y) Te-100 B Ru-100 17 s for approx 1000 kg. Mo in core or about 10'000 moles the Mo-98 gives 2300 mol. The irradiation rate equals: y'1e=39 - (2.3 x 10°) x (6 x 10°7y x (10 x 10797y «x 1010 Atom/s pro sec = 1.2 X 1017 Atom/s Total Power Plant thermodynamic efficiency 68 Fig. 28 Approximate thermal efficiency using a turbine cycle with various working fluids 754 70- Two circuits with 65+ dissociating fluids in gas turbine 90 bar 60- e 7 (N,0,, /8O, )+ (AL,C1 /A1C1 ) 50- 454 404 354 (%c) 30 T T T T T 400 500 600 700 800 S00 Temperature 69 After 700 hrs the steady state concentration of Mo-99 (t 1/2 = 66 hrs 2 2.4 x 1053: A= 3 X 10'—5 s-l) equals: 17 Mo-99 _ 1.2 x 10 _ 21 steady 5 = 5 x 10 Atom The activity of the Te-99 (t 1/2 = 2.1 X 105a = 6.2 x 10125: A E lO_l3 s-l) after 3% years irradiation of 100 kg of molybdenum in the fast reactor core: Activity?g_izar) = 1.2 x 1017 Atom/s x (3 x 3.1 x lO7 s/year) 10717 X T 2 3% Curle/l tonne of Mo 3.7 x 10 IX b) Radiolysis of molten chlorides in very high flux of fast neutron The nigh flux of neutrons and especially of the fission fragments results in a high damage in the f{uel material and cladding. In the type of reactor under discussicn thils problem seems to be much easier to solve because 1) the molten chlorides arec fluilds with structure that is with a very low activation energy for recombination of the radiolysis product and therefore with a very high recombination velocity. 2) in core there are no structural materials (tube, clalding ete, ) 70 X. Aluminium trichloride as a secondary coolant and working agen® In this consideration for the power production a non conventional gas-turbine cycle is postulated: we have chosen aluminium tri- chloride, AlClB. Tt is now clear that this chosen secondary coolant-working fluicd substance has specific additional properties, which make 1t particularly intersting in the present application. The reason is the following: it is well known that real progress in 1im- provement of the thermal efficiency of turbines is limited by two parameters: 1) for steam turbines, corrosion effects with non austenic steel above 580°C on the one hand, and on the other hand the dramatic increase of costs when an austenic steel is used. 2) for gas turbines (e.g. helium), the operation of the gas- compressor consumes more tnan half of the energy produced by the power turbines. A number of methods are available for overcoming these limitations, one of these proceeds as follows: the working agent in the turbine, when the temperature decreases undergoes a chemical reaction which decreases the volume. This result is most conveniently obtained when the working substance polymerises at the lower temperature. Of course, this polymersiation must be completely reversible, that is to say that, at the higher temperature, the depolymerisation process 15 effectively complete. One of the most promising working agent 1s aluminium trichloride, which, at the lower 71 temperature dimerises by the following mechanism: (2 A1c1,) B2° o~ (Al.c1,)895 L (A1, C1,)t1a 27767 27767 . mononer dimer dimer higher temp. lower temp. lower temp. When the temperature increases a monomerisation process takes place. Already in 1959 Blander and co-workers discussed the prob- lem of AlCl3 as working agent. They wrote: "A number of typical application (including a gas turbine cycle employing AlCl3 as the working fluild and a binary vapour cycle employing water vapour for the lower temperature: M.T.) have been considered but 1in none of these has the aluminium chloride shown outstanding advantages over more conventional media. However, it is believed that for some specilal applications it may well prove to have outstanding advantages, where the characteristics of the other system com- ponents are such as to make 1t possible to exploit to the fullest - the unique characteristics of aluminium chloride" (Blander, 1959). Much more optimistic conclusions could be seen in the papers of Krasin and Nesterenko (1972). These authors maintain that the overall thermal efficiency of a dual cycle with aluminium chlo- ride as working fluid in the high temperature cycle and N02 in the low temperature cycle promises to increase up to 60% for 500 °C and 90 ata (see also: Krasin, 1971). But with aluminium chloride and with nitrogen dioxide the chemi=- cal industry has considerable experience, including experience with corrosion problems, in the high-temperature region. Unfor- tunately there seems to be a lack of data concerning the erosion and corrosion behaviour of these substances in turbines and on their behaviour under neutron irradlation in reactor conditions. 72 Aluminium chloride as a working fluid in a condensation turbine has the following advantages: 1) the efficiency of the power generation is about 30-50% higher than for the "classical" working fluids; helium or steam. The quantity of fuel used 1is correspondingly smaller. For a country in which the sources of uranium an and plutonium are limited this advantage may be decisive. 2) the waste heat is smaller an the temperature of the hea- ting water is lower. In a country in which the only sources of cooling water are the rivers, such an advantage 1is of importance. 3) the size of the turbine 1s approximately 5 times smaller than for a steam turbine of the same power. For a direct cycle (when the reactor coolant agent is the working fluid in the turbine) when the circult is radiocactive thewsmall size of turbine is an important factor, L) all these properties of an aluminium chloride c¢ircuit have a significant influence 1n reducing the capital cost of 2 power station. Let us repeat the advantageous properties: decreased fuel con- sumption, reduction of waste heat, decrease of turbine size. All these trends promise to give a significant reduction in power generating costs. The crucial problem of possible corrosion processes caused by AlCl3 may be controlled by the small addition of hydrogen. Acknowledgements For discussion parts of this paper the author thanks: J. Ligou (neutronics calculation), Dr. G. Markoczy (heat trans- fer), Dr. J. Peter (chemistry problems), G. Ullrich (metallurgy), Dr. G. Seifritz (neutronics). The author is also indebted for useful discussions and criticism concerning fast breeder molten chloride reactor to Dr's, Dawson, Long (A.E.R.E. Harwell) and Smith (A.E.E. Winfrith). Also Dr. P. Tempus for his discussion, criticism and support and Prof. H. Gridnicher for his encouragment and support. Finally Mr. R.W. Stratton for preparation of the text, Literature Beattie J.R., A Possible Standard of Risk for Bell G.D., Large Accidental Releases. Blander M., Epel L.G., Aluminium Chloride as a Thermody- Faas A.P., Neuton R.F., namic Working Fluid and Heat Trans- fer Medium. ORNL-2677 (1959) Development of Adequate Risk Standards Blomeke J.P., Nichols J.P., Managing Radioactive Wastes. MeClain W., Physics Today, 8 36 (1973) Cheverton R.D., H.F.I.R. Core Nuclear Design. sims T.M., ORNL-4621 (1971) Claiborne H.C., Neutron Induced Transmutation of ’ High level Radioactive Waste, ORNL-TM-3964 (1972) Crouch E.A.C., Fission Products Chain A.E.R.E.-R-7394 (1973) Farmer F.R., Proceed. Principles and Standards of Reactor Safety. TAEA-SM-169/43 (1973) Gera F., Considerations in the Long Term Jacobs D.G., Management of High level Active Wastes. ORNL-4762 (1972) IAEA Proceed. Principles and Standards of Reactor Safety, TAEA-SM-169/43 (1973) OECD Report Disposal of Radioactive Waste, OECD, Nuclear Energy Agency, Paris 1972 Rosenthal M.W., The Development Status of the Molten Haubenreich P.N., Salt Breeder Reactor. Briggs R.B., ORNL-4812 (1972) Taube M., A Carbide-fuelled Fast Breeder Schumacher H., Reactor with Molten Chloride as Fuel Bonding Material and Chloride Vapour as Coolant. EIR-Report Nr. 167, (1969) Taube M., Molten Plutonium Chlorides Fast Ligou J., Breeder Reactor Cooled by Molten Uranium Chlorides. Ann.Nucl.Sci. Eng. (in press). Taube M., Steady-state burning of fission products in fast/thermal molten salt breeding power reactors. Ann.Nucl.Sci. Eng. (in press). Walker W.H., Fisslon Products Data, Part I ALE.C.L.-3037 (1968) Wolkenhauer W.C., The Controlled Thermonuclear Reactor as a Fission Product Burner. Am.Nucl.Soc. Meeting, BNWL-SA-4232 (1972) See also Am.Nucl.Soc. Winter Meeting 1973, p. 52