IR -Bericht Nr 332 EIR-Bericht Nr. 332 Eidg. Institut fur Reaktorforschung Wirenlingen Schweiz Fast Reactors Using Molten Chloride Salts as Fuel Final Report (1972-1977) M. Taube Sily Warenlingen, Januar 1978 FAST ~ REACTORS USING MOLTEN ~ CHLORIDE SALTS AS FUEL FINAL REPORT (1972 - 1977) | , [ - e are " -~ SWISS FEDERAL INSTITUTE FOR REACTCOR RESEARCH CH-5303 Wlrenlingen Januar 19/8 Co-authors and contributors to this work Physics, Neutronics J. Ligou Cross sections and codes E.H. Ottewitte J. Stepanek Thermohydraulics K.H. Bucher M. Dawudi Chemistry and experimental Dr. E. Janovici Ur. M. Furrer Other assistance Programming - 5. Padyiath Code operations -~ B. Mitterer Text preparation - R. Strattan Support and encouragement Prof. H. CGrénicher EIR, Wirenlingen Or. P. Tempus EIR, Wdrenlingen Or. J. Peter EIR, Wirenlingen Cr. H. Schumacher EIR, Wirenlingen Support and encouragement 7. Fougeras ‘Fontenay aux Roses J« Smith Winfrith U.K. C. Long Harwell U.K. L.E. McNeese Oak Ridge National Laboratory U France C e Do A Ssummary This report deals with a rather exotic "paper reactor” in which the fuel is in the form of molten chlorides. (a) Fast breeder reactor with a mixed fuel cycle of thorium/ uranium-233 and uranium 2538/plutonium in which all of the plutonium can be burned <4 44fu and in which a dena- tured mixture of uranium-233 and uranium-238 is used to supply further reactors. The breeding ratio is relatlively high, 1.583 and the specific power is 0.75 GW(th)/m® of core. (b) Fast breeder reactor with two and three zones (internal fertile zone, intermediate fuel zone, external fertile zone) with an extremely high breeding ratio of 1.75 and a specific power of 1.1 GW(th)/m’ of cocre. (c) Extremely high flux reactor for the transmutation of the fission products: strontium-90 and caesium-13/. The effi- ciency of transmutation is approximately 15 times greater than the spontaneous beta decay. This high flux burner reactor is intended as part of a complex breeder/burner system. (d) Internally cooled fast breeder in which the cooling agent is the molten fertile material, the same as 1n the blanket zone. This reactor has a moderate breeding ratio of 1.38, a specific power of 0.22 GW(th)/m® of core and very good inherent safety properties. A11 of these reactors have the fuel in the form of molten chlo- rides: PuCl3 as fissile, UCl4 as fertile (if needed) and NaCl as dilutent. The fertile material can be 2*®UCl1, as fertile and NaCl as dilutent. In mixed fuel cycles the **’UCly is also a fis- sile component with 232Th614 as the fertile constituent. In some special cases a hypothetical mclten fluoride has been checked using PuFs3 as the fissile, UF3 as the fertile and NaF and Zrl, as the dilutents components. In this case one obtains a lower but still respectable breeding ratio of 1.5, In all cases a directly coupled continuously operating roproc- essing plant is proposed. Some of the technological problems of reprocessing are discussed. Furthermore the report touches on some of the difficulties associated with corrosion arising from the use of these molten media coupled with the irradiation effects such as structural damage from fast neutrons. The thermohydraulic studies show that even under the extreme op- erating conditions of very high neutron fluxes and high specific power, cooling is possible, in most cases by out-of core cooling but also in one or two cases cooling internally in the core. Some molten salt reactor specific safety problems are discussed. The influence of fast neutrons on the chlorine, forming sulphur by the (n,p) reaction has been experimentally investigated and the results are reported briefly. With this report the work of several years at the Swiss Federal Institute for Reactor Research is brought to a conclusion. CONTENTS p3ge Forward 8 1. Molten Salt Reactors. General Description 9 1.1 Methods of classification 9 1.2 Method of cocoling. External: Internal 9 7.3 Intensity of Neutron Flux 10 T4 Number of core zones 12 1.5 Type of Fissile Nuclide. Plutaonium: Uranium 12 1.6 Neutron esnergy: Thermal and Fast 12 1.7 Purpose of the reactors 15 1.8 Fuel Components. For Molten Salt, Fluoride and Chloride 16 1. Short resume of the classification 16 1.170 Method of Neutronic Calculation 17 2.1 Breeder Reactor with Plutonium Burning 4n sLtu 19 2.7.7 Aim of this Concept 19 2.17.2 Reactor in the Build-up Phase 2 2.17.3 The Transient Reactor 29 2.17.4 The steady state reactor 29 2.17.5 Comparison of the three phases 35 2.2 Impact of some parameters on the mixed zone two zone fast breeder 43 2.3 Material balance of the steady state reactor 49 2.4 Conclusions 49 3.1 The Three Zone Reactor 51 3.7.7 Introduction 51 3.17.2 The three zone breeder with thorium/ uranium-233 57 3.2 A Three zone breeder reactor with a mixed tfuel cycle U-238/Pu-239 plus Th-232/U-233 52 3.3 The three zone reactor - uranium-plutonium fuel cycle 57 3.4 The three zone breeder reactor: Very high breeding gain 63 I> B A The Two zone fast breeder. Fuel of uranium plutonium fluorides 3.5.7 Introduction 3.5.2 Arbitrary assumptions and uncertainties High Flux Reactor with Fluoride Fuel High flux burner reactor for transmuation A o~ O Need for fission product transmutation — 1. Introduction 1.7 Why some opinions concerning trans- mutations in a fission reactor are rather pessimistic 4,1.3 Which fission products are suitable candidates for ftransmutation and in what quantities? 4.17.4 In what way could a burner reactor be coupled to a system of breeders: 4,17.5 1Is the rate of transmutation sufficient? In what reactors is transmutation possible? 4.1.7 What are the limitations of a solid fuelled reactor? 4.1.8 The liquid-fuelled fast reactor with central thermal zone 4, 4. N = RN ) The Neutron-physical aspects of the High Flux Reactor (according to Ligou, 1972) 1 Introduction 2 Neutronic calculations .3 Moderation requirements 4 Influence of other parameters B MM Thermohydraulic considerations Some results Comments on hazard coefficients Secondary processes Conclusions page /70 /0 /0 83 An internally cooled breeder with uranium- plutonium fuel g1 oy U U LN - Design features adn objectives The reference design Neutron physics Satfety problems, comments Chemical and Related Problems 6. 6. Experimental Work (according to Ianovici, 1976) 7 /. 1 7 B w 1 2 Physical and chemical criteria for slat components Corrosion of structural material 6.2.17 General criteria 6.2.2 Molybdenum as structural material 6.2.3 The irradiation of molybdenum and iron in a fast high flux reactor Fission product behaviour in the fuel Some comments on reprocessing In core continuous gas purging The proposal 6.5.1 6.5.2 Delayed neutron emitters Chemical behaviour of radiosulphur obtained by *°Cl{n,p)?°S during in-pile irradiation Temperature dependence of sulphur species (accaording to Furrer, 13/77) P3ge 137 137 147 147 154 177 wl page Thermohydraulics 185 .17 Introduction 185 8.2 High Flux reactor with the core as a spherical shell 135 5.3 Power Reactor with spherical core 169 8.4 The external heat exchanger 191 8.5 The internally cooled reactor 194 References 199 9.1 List of EIR publications used in this report (chronological order) 199 9.7 Former publications concerning molten chlorides fast breeders and the fluoride thermal breeder 200 3.3 Publications concerning transmutation 204 J.4 Publications concerning the thorium fuel cycle 206 9.5 References to the experimental work (chemistry) 208 5.6 References to the physics calculations 208 FORWARD The history of the development of fission reactor concepts using molten salt as fuel media is as old and as complex as the history of the development of nuclear power 1tself. The ups and downs have followed those of the parent technology but the swings have been if anything more violent. In 1976 for example molten salt technology all but died out but then in 1977 a new attempt at revival was bepun this time associa- ted with the new interest in proliferation-proof systems. The author of the present paper has a profound belief that the concept of molten slat rsactors coupled with continucus reprocessing and the associated waste management will become an important feature of nuclear strategy perhaps in 10 or 20 years time. In this report the efforts in this field over the last =six years are summarised. 1. MOLTEN SALT REACTORS. GENERAL DESCRIPTION 1.1 Methods of classification There are many ways of classifying a reactor type. One such pos- sibility 1s shown here. a) Method of cooling b) Flux intensity related also to specific power density c) Number of zones in the reactor d) Kind of fissile nuclides and fuel cycles e) Neutron energy f) Purpose of the reactor g) Dilutent for the molten salt It 1s clear that such an arbitrary classification is not neces- sarily internally compatible and not all reactor types fall easily into the scheme chosen. 1.2 Method of cooling. External: Internal Molten fuel reactors differ from the point of view of the cooling system. The following are three types of molten fuel reactors: to the external heat exchanger. In this type of reactor, only fuel and fertile material are present in the core (no cooclant). The large amount of molten fuel ouside the core does not of course contribute to the critical mass. This type of reactor has been discussed for example by Nelson, (Argonne 1967) and Lane (USA 1870) especially as a high flux ma- terials testing fast reactor. In externally cooled fast reactors the loss of a portion of the delayed neutrons could adversely affect reactor contrcl. Also the biological shielding outside the core is very expensive. In this paper most of the reactors discussed are externally cooled. Internally, direct cooled reactors: here the cooling agent is pumped directly into the core where, after mixing, the fuel in the lower part of the core is separated and pumped out of the core to the heat exchanger. The direct contact of molten fuel with molten coolant has several particular advantages! very good heat transfer, no coolant tubes (or cladding), possibility of transporting fisslon products. The disadvantages are unfortunately, also numercus: problems of mixing and separating the fuel and ccolant, corrosion, etc. This type of reactor has also beern studies, e.g. cooled by molten lead (Long, Harwell and Killingback, Winfrith 19679, cooled by boiling mercury (Taube, Warsaw 196E8) and ccooled by boiling aluminium chloride (Taube, Warsaw 1566). This type of reactor must bhe considered as an "extremely exotic type”, and only some references are given here. Internally indirectly cooled reactor: here the cooling agent flows through tubes in the core. Heat is transferred from fuel to cooclant across the tubes. No direct contact between molten fuel and liquid or gaseous coolant 1s permitted. These types have alsc been studied, 1In most cases using sodium as a coolant, (Nelson, Argonna 19687]) or molten chlorides of uranium (Taube, 1370}, See Fie. 11. 1.3 Intensity of neutron Flux The molten salt reactors discussed here can be used for two more or less guite different purposes. - power production and fissile breeding, which is self evident - neutron producticon for nuclear transmutation of the long radionuclides produced in power reactors. In this report both types have been considered - power breeding ractors with a mean power level of approx 3 GWith) and steam production with over critical parameters. - burner reactors with a very high neutron flux particularly in the internal zone for neutron moderaticn when the thermal flux reaches 3 x 10'% n cm™ %571, Fig. 1.1 TYPES OF REACTOR COOLING SYSTEMS TYPE SCHEME CHAPTER INTERNAL CHAPTER 8§ INDIRECT COOLING \. J ____> i fanuun EXTERNAL INDIRECT CHAPTER COOLING 2, 3, 4 9 J L > INTERNAL : ( I f DIRECT ) 1 B Ta Ve W W COOLING o OO (o) (BOILING) o O O o o © Qo ©O oo here not O discussed oo O . e J U —J 1.4 Number of core zones The division of the reactor into several zones must be consid- ered from the point of view of neutronics, thermohydraulics and safety. The organisation of multiple zones is easier in the case of molten fuel reactors than for solid fuel reactors. In this re- port two types are discussed - with two zones - wlth thrse zones including outer and inner fertile zone, (soe Fig. 1.7). 1.5 Type of Fissile Nuclide. Plutonium: Uranium The fast reactors show excellent neutron properties, not only tfor the fuel cycle: Uranium -238/Plutonium bhut also Thorlum -232/Uranium-233 Also a mixed fuel cycle of both types has some spectial advan- tages. Fig. 1.3 shows the nuclear properties of the fissile nu- clides. 1.6 MNeutron energy: Thermal and Fast The reactors discussed here are all fast reactors. Thermal reac- tors however have also been extensively and intensively investi- gated during the 1360's and 18/0's in Oak Ridee National Labora- tory USA. (Rosenthal at all, 19727 _"]3_ Fig. 1.2 TYPES OF REACTORS PROPERTIES NUMBER ZONE GEOMETRY POSITIVE NEGATIVE SIMLICITY OF RELATIV ONE TECHNOLOGY LOW BREEDING RATIO, BECAUSE OF TOO SOFT NEUTRON. FERTILE BLANKET TWO THE ORTIMAL DESIGN GODD USE OF THE GEOMETRY NEUTRONS IS VERY COMPLEX THE NEUTRON FLUX THREE BDISTRIBUTION IS DISTURB Thorium Uranium Cyecl /\ Stable in Nat () beta stable (Q beta unstable thermal a is ri \(n,y) 15 O e ure ’neutron capture —* heta decay ’[ fission elativ big Tt Fa S T T T BB a1 i z {ross section Cross Section of Pu-239 (simplified) 108 varns (13 1000 4 100 4 107 “fission oM, ) Gcapbure\ aln,y) \\ \ 1 \ \ \ \ \ \ AY N\ \\ 0.1 1 ~ ~ AN N\ \ \ 0.01 T T T T 1 1072 1 10° 10" 10° Neutron energy, eV Uranium-Plutonium Cycle - . 2 = cross section for neutron reaction (cm”) g = A $ = neutron flux (neutron cm'Zs—l) X = decay constant (s-l) all data for ¢ 2 10%%n en 57t 1.7x1077 241 | fast e e op :j107%? 240 4 A ) [} t 1o ofast is small af =12 * i 1 ! [} x oz o st €337 O 2.5 min O .7 days \’ o8 o 10740 238 A U Np Fu T T T 92 93 RE Z Value of n for fissile nuclides {simplified) n 34 Ve Pu-"f2 ///// —-—n minimun 2 1 Thermal Fast Reactors Reactors 0 2 T E T 4 LI 1 q - : 10 1 10 10 10” 107 Neutron energy, eV 1.7 Purpose of the reactors 15 The principle purposes of the large ractors proposed can be classified as follows. Table 1.7 (Table 1.1) Reactor type Primary Aim Secondary Aim Comments to be found in chapter Power Electrical Production of ch.5 energy . fissile nuclides T > 600 C 2R > 1 dreeder Production of Production of ch.3 fissile nuclide electrical B.R. Vv optimum ENergy High flux Neutron Flux Production of ch.4 d (n cm %5 1) > 1pt® for transmutation electrical energy High Temperature T > 850°C for chemical reactions Production of clectrical energy not discussed here Non-proliferating Maximum security. No plutonium output Production of electrical energy ch.? Propulsion Heat for steam turbine not discuseed here Space Heating Heat with 100 < T < 200%C not discussed here 1.8 Fuel Components. For Molten sal+, fluoride and chloride In the thermal molten salt reactor the best fuel compound 1is unduobtedly the flucrids. For fast reactors the use of chlorine as the compound seems to be preferable but the use fo fluorine (as zirconium and sodium fluoride) as dilutent is not excluded. 1.9 Short resumé of the classification Table 1.2 brings together all these characteristics 1n an at- tempt at classification. Table 1.2 This work: yea no Method of External X Cooling Internal Direct X Internal Indirect X Flux Intensity High X Low X Number of One X ZONRS Two X Three X fFissile Plutonium Uranium Mixed Lnergy of Thermal Neutrons Intermediate Fast X Aim Power X Hreeder X High Flux burner X Dilutents Fluoride Chloride 17.70 Method of Neutronic Calculatiaon Almost all results given here have been obtained using the following calculational method - the reactor code: ANISN - number of zones: 5, 6 or 7 - 40 - 100 spatial positions - order of quadrature S5, checked by Sg - neutron groups: Z2 or 23 groups including the thermal neutrons (see Tabic 1.3) - anisotropy by first order Legendre expansions - library ENDF/BI, BII and BIV processed by code GGC-3 and GGC-4 - the management of additional sub-routines have been realised by RSYST. Table 1.3 Relative Fluxes in Each Group 18 - Upper Mean Au Core o Centre boundary value (Lethargy) boundary AL_S 1 15 MeV 12.2 MeV 0.4 - 107" (.0002) L0010 2 10 8.18 0.4 0o (.0025) note L0045 3 6.7 5.4 0.4 0020 (.0118) oz 0112 4 4.5 3,687 0.4 L . (.0286) o7z 5 5.0 2.46 0.4 222 L .0146 U ‘é] nan7y ; 35 g 2.0 1.65 0.4 < naot) .0285 L0264 1.35 1.23 0.7 019 / (.0536) 3 . L0281 o a 1.11 1.00 0.7 Coean) 0215 .nazeg q n.9 78 0. o 0371 § e e . (.06837) | 10 .87 0,55 n.4 L2 0787 : : ‘ (.1373) | .45 0,37 0 He 007z 11 .45 . 37 L4 ( 4ona) 977 _ 1162 2 0.2 .25 . 1130 d 0.25 0.4 (.1047) 113 1044 3 20 N.165 0.4 ne4 q - (.0749) 1 10486 / 0,135 0.10 .45 A1 14 13 108 0.4 Cane 11118 _ i L1061 i 15 86.5 keV 50.5 kel 0.75 C esa) 1200 1043 6 4 5, i 6 16 10, 8 5.0 1.00 * ha04) 1262 L0589 15.0 2,0 .25 : n748 17 5 o 1.2 14 n7 0047 ' . 2.9 a. 008 18 4.3 2,94 0.75 Lo 1051 | .N148 _ 2.03 1,20 0,75 _ 0708 ‘ (.0025) e .0048 20 0.96 N.67 0.75 .0079 (.0004) o7 .0008 21 0.45 0.24 1.25 0017 l C10- 001 5.75 < 10-" | 22 0,13 kel 0.4 eV < 10 § < (107%) Total 0.4 eV - 15 MeV = 17.40 1.0000 1.0000 2.1 BREEDER REACTOR WITH PLUTONIUM BURNING IN SITU NI - [N Alm of this Concept The aim here is to demonstrate the possibility of using a molten chlorides fast breeder reactor with external cooling as a device for consuming all plutonium produced, 4n ALtu. At first the re- actor is fuelled with denatured uranium -233/uranium -238 and this 1s changed stepwise to a feed of thorium and depleted or natural uranium only. Such a reactor will have the following phases in its fuel cycle (Table 2.1) See Fie. 2.1 Table 2.7 Fuel input Fissile Fuel output Phase PUF”Fd Fertile fissile o setu Fissile Fertile start. Build-up| U-238: 70% {U-233: 30% U-233 none none nhase. Th (Fig. 2.1 A) Transient Phase| U-238: 70% [U-233: 30% Pu-239 none none (Fig. 2.1 B) Th U-233 Steady State U-238 none Pu-2393 mixed Th + other -233: 30% =738 Pu-isotopes 20 - Flg. 2.1 Two-zones Reactors with uranium-233/plutonium-239 A) BEGINING (NOPU-417) from the reactor B) TRANSIENT (NOPU-501) BLANKET C) STEADY STATE (UTMOST NOPU-302) \ V J U238 o> U233 for a new reactor j 2.17.2 Reactor in the Builild-up phase At the start of the cycle the reactor core is fuelled by ura- nium-233 denatured with uranium 238 {(see Fig. 2.1). [(Table 2.7] Table 2.3 gives information concerning - the method of calculation - densities of elements in each of the 5 zones (core, wall, blanket, wall, reflector) Table 2.4 shows the neutron balance in the core and blanket Table 2.5 gives the breeding ratio calculated by a microscopic metnod of the form: - and the macroscopic method by _ production rate of fissile nuclide macr rate of destruction of fissile nuclidse BR - and the maximum neutron flux which gives informatlicon on the flux spectrum in the core. Table 2.6 shows the geometry of this reactor e.g. see Figp. - radius: 0.955 m - volume: 3.65 m? and - specific power: 0.75 CW(therm)/m’ of core - total power: 2.8 GWltherm) and inventories of fissile and fertile materials. Table 2.7 gives some informatlon concerning - the material flux in this type of reactor. (for more ser sectiaon 2.3). OBJECT Thorium-uranivm Breeder with Plutoniom burnine in situy REACTOR TYPE : Power, Hreeder CEOMETRY : INTERNAL ZONE : Fuel WAL L Ser INTERMEDIATE ZONE . Fle WALL : P EXTERNAL ZOUNE : Fertile zone WALL, REFLECTOR POWER (GW thermal) a3 B FOWER DENSITY (GW therm/m3 core) @ 0,75 NEUTRCON FLUX, MZAN (n/cmzs] s 1.6 x10 FISSILE NUCLIDE . Pu-22¢/Pu-241 in core, fuol U233 in fertile FERTILE NUCLIDE - U238 in core, ThZ327 in fertiie DTILUTENT : Chloride COOLING SYSTEM Cuter BREEDING RATIO 1.58 PARAMETER STUDIED : Make up reactor with U233/0238 transient reactor with U233 + FHy23@ steady state reactor with PouZ3¢ METHOD OF NEUTRONIC : See table 1.10 CALCULATION — 23 - Fig. 2.2 Three zones reactor °.8 GW th (meter) TITLE 3 DATE ¢ NA PA UTMOST-417,NU2U. SRSV FIFFERRERL PR TR R SPHZRIC GcOM. C3/s709/77 INPUT OJATA (RSYST FORMALIS M) RN ERELEYE RRRERRRER RN R EEFEERD ORJ_R OF SCATTIRING QUL DRA TURE ORUER NO. OF ZONES NOe OF INTERVALS NO. OF GROUPS CI5_NVALUE MODIFICR -2 PR-CISION UESIRZD L NORYs FACTOR 2 MoW PARM. MOD. SRCH. 1 g e SO & DENSITIES(ATIMS®L L -24/0M3 FEFRERERELERFEFRRERRRFEXERDE CORE = 7.04E=-03 CL = 1,30c2=-02 = D.obE-U4 PU2C39 = 7,31c-1i7 = 7T.31t=-47 FP239 = 1.6lip=-4d7 Ui/ (PUTOT + U3) = Be37 FIRST HWALL = B.b2E~ud Fr = ¢J.88E-d¢ LA = SeklE=-ud oL = 1.u8-0¢ = 1,3092=0C5 u2a33 = 1,03E8-15 2=NDJ WALL = 1,37E~u?d F = 7e3lce=-ud REFLEETOR = .06 Fo = Babdp=-ud » ue23s PUC 44 THZ 32 4ol 5c=ud 7317 'D.:-}i\"_‘t43 TITLE ¢ DATE ¢ UTMOST=417 4 NGB U FPUREEFPEL IR XD ES R L BER SPAZRIC GoOM, Usrbari? FEFRFERFTRRL R FL B NR S IR B F R EXR FREFFF RN Y XL LA S X FE RN SR S F R L X E 5 ¥ ¥ 3 * » 4 S % * * 3 ¥ ¥ ¥ x® 3 ¥ & * 3 3 ¥ ¥+ ¥ *® £*3 ¥ %x x * 3 E'A 3 * ® "3 L 2 ' * cLEMENT b B B e B Ba B | CORL ue33 uc3s PuU23Y pu2ul PUZ4L CL FE NA G FP TOTAL CO-. BLANKLT TH23Z2 ue3l3 PA CL FE NA C TOT. BLAN. ENTHO * LEAKAGL T OT7 A o NEUTRONLCS FRRRFEEFFRRFER LY AYSORPTION R R R R T ] 03.992E+18 6E5«836+18 Fden7uls 32.,01B8E+15 B7.8495c+#.15 1l.638E+48 21le373cE4+17 10.552E+17 13.334E¢10 12:735F¢ 1L 15,1108413 (%] S0 .043E+18 1l.181zZ+17 59.619c+15 6CUbTE+LT LU+230EHLB 27 «54lceln 742516413 v3.1/72#18 13.285E+18 22,0E8E+19 MALR ltuaeuwl PRODUCTIUN M RN v w ww =y 16.939c+19 549.530E+18 13.217¢c+1p Jlec+ls e U716 o 0 J o « 3830417 S.ul7E4+17 . J £ ] s b 43,290 ¢ 17 2l lodt+17 1bdeuu P R I B IR B B B T T T S T T S S R S LR S O L B L IR R R e R R I XTSI NS R RN T R N R ST E YT ST I F A R THE REACTION (n,2n) UTH40ST-417,NO?U, FEREERFEEVFARRERRER Y TITLE 3 SPAZRIC GEOM, : Ugrs09rs77 R SIS R N P R e I TR R P S N YR RN TRy S e L L L IR BN B R BEE B 2NN JEE BN SRR IR BNC DK BER-EE TR JEE DK JNE BN BN BEE BEE BN SEE BN & MAX. NEUTRONIGS MICR FAR U R X NEN S R RN ENTITY SYM30L NIO NIl NIZ NI3 NI8 NI9 ALFHM FASHB ABPAR LEAG PRODUCTION FAST BONUS ABSOR.?ARS LEAKAGL BREDLRATIO MICR PU39 -PRODUC PU3S DLSTRO U33 PRODUC u33 DESTRO BALANCZ U233 BALA33 BALANC: PU239 BALA3Y TOTAL 3REZDING MACR BRTIT VALUE 3.183 3,028 2455 2550 2+945 € +990 « 485 319 173 58 1.357 e fc «159 137 1.633 W3 d7 b eddl 7.111 -2 4120 4501 le3uc FLUX IS ¢ 2.42E+15 FOR GROU? 11 IN INTERVAL | IR R TR Y R E T N Yy Y Ty e TRy *COMMENT ¥FASB, BONUS, (FAS3 ¥ (NIs - 1)) / ¥*ABSORPARS, ABPAR / (1 + ALF). * EAKAGE, LEAG®ZNIM / (L + ALF). ¥*PRTOT, *CAPT. (U3 + PUS + PUL), I R T R R I R R Y R Y Y S S N I IR PR T TEE (1L + ALF). N¥SIoMA FISS.FeRT / N*SIOMA FIS5.FISSIONASLE. *ABPAR, N¥SIGMA PARAS / N¥CAPT(FERT + FISS). ¥LEAGy LEAKAGE / TOTAL AJSOR(FROM TAJSLE NEUTRONICS) » *PRODUC, (ZNIM - 1 - ALFM) / (1 + A_F), *FAST N¥*SIGHMA TRANSMUTATION (TH + U8 + PUC) 7/ N¥3SIGMA. ¥ * % -3 & = x ¥* "3 ¥ "3 ¥ ¥ * ¥ ¥ .3 ¥ X ¥ 4 «> 08 ¥ 3 % ¥ * ¥ x * F-3 ) » * = '3 I3 ¥ 3 ¥ L & * ¥ SR TR I8 ¥ I LSRR RIEER LT DY B $E B E T T el e s BN R N TG LESFERALREANBF LY RERRBE XTI T PR x e VALUc ¥ v e g v oy ey gy - ww % ¥ *x o # . 355 % - ! t el ¥ y . S I 20 ¥ ’ o - o Ud * i : H i NETRTEY ¥ . - . sUUJ 5 L E W p¥E 59’355 ¥ : G e g il.0: ¥ £ ¥ * oM a7 33 * WM PR ¥ A G A 2a7 390 * e JHS IR RS o fDJ ¥ . R A Rl edu ¥ S R MR G 857,330 ¥ FS ¥ x & T LT - Lydpd ¥ oo Thivdeam b, AT a7 K3 3.135 ¥ S K5 Budsdlc ® N PO B Yo K3 470y ¥ : tonT, UsTOT K3 5854.340 @ : fn il Tt o7 Koo 23s13,349¢ # R ) L i VT - 2l43.0l1l ¥ ¥ £ -3 R - ! ¥ s e Yoo A S . ¥ ; NEUJT/SP(OME L, loiz+ ¥ % s B RRB B RS AL SR A RN R RN R ER KSR R AR K F L S & 8 FENFERFP IR PP L EE RS FIP VAR T PR IR PE R RRCEF FFIR L FERLZF RSP E SR 4 * ¥ ¥ TABLE CERBEEE MATZRIAL BALANCE OF THE 3RZcDER SYSTEM INCLUDING NEW REASTOR x Y ¥ ¥ P R R Y I P I I S Y N P T R R RN 3 * *ORS *SBL *F 8L *FTR *SCO *FFP *FCB *FLR *SRE *F WA “F NW PU3Q ¥ 2.03b+ucH ¥ * * l.dbe+ud¥* ¥ lsdobk=-Uo* l.GbE—US" 1odbt;“c4* ¥ THOR PA y233 uc3s FERFEEVEAREFERERNF RFIEEE IR ERREREFLFEFRIENFEFEXN X SREL XX R XXX FEFE L 2,33:+006 belUbE+UZ 5.37E4+02 =1,87£+U2 2.36_+04 5.69E+U1 L,74£+01 84842400 8.492+090 7+3%c4+32 Bslb+(2 5.37c#(2 S «69E+02 8 ,00:+32 6.86E+03 C 52E4UL2 HeiS5_+02 8.21E+02 Ce 30+ L0 5«63E-43 a7 4E-]I3 be8bE+0U 2e306-+#00 5.69c=03 B.,006-J1 6e80E+0U B.i5c-01 «2.,02¢-31 beBOE=Uuy 2.33_+0¢6 -2, 026402 -1,01F+03 Ced3btl g -2+0yt+du -1.026¢01 *SNW ¥ PYFFEPFENPLT X R R R P TEF L RN LR R FEX PSRN PP AT N EE NV ERFF FLEL PR RS R HHERE ORS SBL FBL FTR SCO FEP FCB FLR SRE FiA FNW SN W LT T T A 1 N A ) FLOW FROM THE ORE STLADY STATE IN T4z FLOW 155, OR SYNTH. IN KG /7 YtAR GLANKzT IN KGo + PARAS.IN JLANKET IN KG 7/ YzAR FLOW FOR TRANSMUTATION IN KG /7 YcAR STEADY FLOWTS STLADY STATE IN TH:z FLOW TUO THE #ASTE IN KG /7 YZIAR FLOW TO THE NtW IN KG / YzZAR I TH:Z NEW AFTER JTIME STATE IN THE CORE IN K5 FLOW FIS55.¢ PARASITIC IN FLOW TO THE CORc SORL LN KG I K6 J2 BLANKzZT IN KOG / YcAR RZPROCESS AND RETURN IN KU 7/ YEAR RE-PRICLSSINGL IN KG / YrAR - 29 - The data given indicates that in the given geometry the nominal power and all related values: temperature, temperature gradient, velocity of circulating fuel, heat exchange etc. are as for the steady state reactor with only plutonium fuel in the core (see section Z2.1.4). 2.17.3 The Transient Reactor Very shortly after the start up of the reactor, a significant amount of plutonium has been produced in the core. The total amount of plutonium chloride, after having been sepa- rated from the fission products but mixed with the uranium chlo- ride and sodium chloride is circulated back into the core. The amount of fresh U-233 required is correspondingly smaller. The reactor now burns two fissile nuclides, the uranium-233 and the reprocessed plutonium. The data given below refers to the case where approximately half of the fissile uranium-233 is re- placed by plutonium-239 and plutonium-241. Based on rather simplified assumptions concerning the isotopic composition of the plutonium it can be shown that the same core cesign is capable of burning the mixture of fissile materials maintaining approximately the same power level. Tables 2.8, 2.9, 2.10, 2.11% and 2.12 give the equivalent values previously shown for the build-up phase. These confirm the sulitablility of the core design for both the build-up phase and the transient phase. 2.1.4 The steady state reactor This reactor phase 1s in fact the main object of the work covered in this report, the steady state fast breeder reactor having the following featurses. TITLE DATE - o0 HEA Ue3s Pligal RATIO HA P A R EE TSR FSEERE LY FEEYE ] SPHERIC GEOM, 16713977 /¢ PUsi), INPUT NATA (RSYST rFURMALISH) I ZETEIAEEEFEFSENERSREENEEENE N RN R NrDER Or SCATTEMING NUADRATURE QRIEFP My OF ¢ulN[S N, OF INTERVALS Niis UOF unrOUPS CIGENVALUE MnoDIFIER PRECISIUN DESIRED NORM, FAUTOR MEeW PARMe MOD. CRCH., i =2.,unplery 1.0000"C3 2,47350+20 1,U0000«C0 DENSITIES(ATOMNS*1E="4/CM3) T ELEEESEEERE RS R ERERERSERS S CNREL = 7,59c="3 L 2 2,88c=04 rg23g US/(PUTUT &+ UJ) = EIRIIWALL 2 A,062e="2 FE DLANKLT d 5.,41-"73 mL 1, 30g="8 11233 2=NdomALL = 1,372="2 FE CeEELELICE LI rr Le30E="D PeS4ceng FeB0E=" 8B o0 NeB8RE~rD e3e PUR4AL 5,A41F =8 TITLE DATE SPHERIC 1672977/ UTH0ST=-501, t B AR EEEREREERANES R Y NE wEOoM, PLasU. T RS EEEER SRR EEE AR R E RS R R R R R AR R R AR R R R R R R R YR SR ERR R R R ENE S . * = . . * * * * - * *« & " " * . » '] » & * » » - - » - " - E » » . . - * » E ELEMLCUT CORL ued3 UpglAa PU239Q PUZ41 PL2AY ClL FE NA C P TOTAL CURE BLAMKEL TH232 U233z PA CL FE N A C TOT, PLAN, BT TWO LEAKATE T 07T AL NEYTROMICS MACRE Acodr e o W N gk ok Bk ABSQREIIOYN 350U77E+18 71.,188E+18 27.072F+18 33,v93c417 4700596*&7 12.,950F+18 23.2B8E+17 11.,¢275F+*7 1S5.0RFE+16 4d01086F+1L 0 1615364179 58.878F418 11./51E+17 785,403R416 650UO0E+17 1y,a50:+18 28 e¥131°416 764,5295+13 72.528F+18 14.126FK+18 23,4n5m 419 [; . ( Q tyc.r’l PRCDUCIIAN 89,A79E+18 62.6498E+18 “B.323E+18 61.250E+17 12.348E+48 23,917E+10 18,379E+17 26.710E+17 o) o 45, "QRE+17 44 ,A6FE+17 "4, THAEH10 36,80 25,73 28,04 2.51 5.07 w00 « 00 «0C w00 «On 98410 75 1.10 00 v0C « 00N + 00 « 00 1.85 1.8°7 100.0°0 FYS R FETEE R A ENS R R RS BWETEEEREFEEEEIFEELEESNNEEER SR SR EREREEE R RE] o . ] » " » » » » * * o * * ¥ * * *® = » " " » * " * L » » * * » ] » * * » " UTMOST-501, PU+U, L EEEFEREEEREBRELE R ENEEYE] TITLE ¢ SPHERIC GEQM, DATE : 16/ 1397147 BT EEEEREREERE BN R BN EEEFFEREEEE R R RSN YRR R N FE SR SR SRS N R EE ] * » * MEUTRCNICS MICR » * I EEEE LA SN RS RS " » CALLTY singol VALUE » * * . NG 3,182 * t NIi 30022 * * NT2 2,456 » . NTT 2.554 * L | NIQ 2.946 * * NIQ 794 * » At FH .136 * . FASP L3800 * * ARPAR o172 - * LFAR M58 * » PPONUCTION 1,426 - » FAST a30ONUS 571 » * ARSCOR.PARS L1582 » " T AKAGE o141 * » ERENGKATIO ajCR 1.784 . » PLIZO PRODUC 4,991 . ® PUZTY DESTRO 2,767 * » Ul PRODUC Hel6 3 » » U3T DesSTRO 3,025 * » DALANCE J23¢ RALAZZ 1,338 * . BALANCE PuU2sy BALAZO D.223 * * TOTAL BREEUING MACR ARTOT 1,451 * - » » MAX, FLUX IS & 2,58M«45 FOR GPOUP 11 IN INTERVAL 1 » ™ ] at*t*t*t*im*&ttt:tt*t*t*n**mt:tt*w***tit*tt#*ttttttttatttt SCHMUENT * *F 238, M*SIGMA FISS,FERT / 1=*=SI0GHMA FI3R,FISSIONARLE. * *ArPAR, H*STGMA PARAS / 1 «CAPT, (FERT + F1S0). * sLEAG, LFARAGE 7/ 10TAL APSORL(IPIM Tadr NOCUTRONICS), » #PLODUC, (INIM = 1 « ALFU'Y s (1 + ALF), * #F LST POMUS, (Faso = (MNIF - 1))y 7 (4 + pi P, » s ASCR,PARD, ABRAK / (1 + ALF), * aL-AKAGE, LEAGeZNIM , (31 + ALF), * *BeTOT, M*STGHA ITRANSMUTATION (TH + Us + PLUQR)Y / MNeSTGMA, *C:/PT, (UJ + PUY + PUL)Y., * I E R R R SR EREREE R R R R RS R R N R RN R R R R R EE R R R R E N E R R R R R TITLE DATE UTHOST=5(1, lable Z.11 PUsu, IR P R R R E R R E R A NS N ERE SPHERIC GEOM, 16/09/,77 - e IR T L R e Y N Y Rl i A I It I # £ & # & % & & # B X & B B % € £ & # ¥ X & B & B ¥ £ B & #H ¥ B 4 X R & ¥ X % & EXIIIY GEOMETRY CORE RaDIUS HALL 1 THICKw~NESS BLANKET THICANESS WALL 2 THICKNESS HEIGHT OF CryuliMDER HETGHT/DIAM. RKATIOQ VO! UME OF CuUre VOLUME OF HULANKET FOWER POWER IN POWER 1IN TOTAL FOWER POVER POWER SORE HLANAET FOWRER JENS, ‘«JENSC AT ING GUPE uf Py INVENTORY TOTAL PU239 TOTALL PU L2333 1N CURE o3 It BLANKET TOTAL 0238 InNvENT, TOTAL IH INvesNT, PATIO 938/{(Pru+tiz3)y INVENT. INVENT, OTHEw UATA PUPH, ~«ATIU PO TG HEAM FLUX (P2 Ty IMPLL T SLANKET SYHEOL PC1) THICK(D) THICK(Z) THICK(4) HEIGH! HPRAT VOLCOR VO RL POTNm POINA POV TN ronee FODECH COWRAT roTnT PHITOT L2 his HATOT THTNOT FATUR CURHR TTING FLMPAN U@il MALUE & 0963 M «040 M 5590 M 040 M +C00 - OOUO M nd J.740 Ma oyl 11,461 Gw 2,749 G v « 059 G 2.8(8 Gu/Men « 735 GW/Mee? 005 MN /KRG 5.091 KG 377n326 K 530,940 KG 400,894 K 47.984 KG 6978.166 Ka 23880,212 - 10,924 - «B26 YEARS 1.201 NEYT/SxCH2 1,334E+16 PR AE R R TR EE R EE R A RS R E RS R E AT SR AR ST TR R PSR EE Ry * » » » * » » » » % L * L * * * »* * *® L ] » » L * » * * » % » » » * » " ] * » Table 2,12 T IS B R RS ERLE SRR REE R EENRE SRR EEESEE AL SRR R E R R R E SRS R R RN ERR FU " tapyf * . KKK R % * L] MATERLAL BALAMCE nf 117 HpeFEDRER SYSTEM . . INCLUDING NEW PRACTUR . PR IR R R R R FE R E R SRR SRR N TR R RS R RS ERE SRS SN EE SRR LR EERE R NN * * THUR PA « URIY » uZg o« PUZO e T T L I T I T Y *ORS 1.92E+0% 6,39Een2 ©,37T4nn 3,83k«03 O, P2Fe02« +S8L 2.39E+14 5 ,7A4r1 4,8 T+iy - oF oL 9 IVE+Y( R,Q5F+0 4, 44F+02 » «F TR fo39ESD 6,305+ 2 be20 402 . S0 4,1rF*02 6eOBE+DT Se77F*02 sFEP P.tEF«02 3e46F+00= «F (8 6.49E+32 R E7E402 * sFLR 2,3%E+u0 5,7AF-"3 4,80 «n3 £,080+00 3,770=01+ «S!E 2.39E+ 0 5,7AC="3 4,40 M=y 6 SBE+CO 3.77C0=01" *F A 5.49F=01 1,800=01 £ 4CBCen4 3,77 =07 wF il 100E+i14 1.,8C0+05 Coedd4l+02 * w«ShW PD.3UE+ 14 2,447+ > 1e820+03 . e Y L L T T T T s R IS N R R R WHI RE PR = FLOW FRO2M [WE ORE CF SYNTH, I8 K6 / YEAR S8, = STCAJY STaje IN THT fpanNkDT Iw FRL. = FLOW FISS. * PARAS,IM BLANKET [N KG / YEAR FTw = FLNW FOR TwaMSMUTATINON [H KGO 7/ YEAF SCrr = STEANY STAtTe [N THI CoRE N K FFP = FILOW FISSe+ PARASITIN [N CORE 1w #C 7 YEAR FCi = FLOW TO THe CORE OP PLANKET IN k0 / YEAR FLe = DUl KEPRUCESS AND RETURN TN / YEAR SRE m STEAULY STafe [N THM PEPROMNFSSING IN KG FWwa = FLOY TO THe WASTE IMN pG / YPAR FNW = [LOW Ty THe NEE M KGO 7/ YO AP SN. = It Tri& GEA AFTER DT IMp 1 00 - all the freshly bred plutonium can be burned in the same re- actor (the history of the transplutonium elements is neg- lected here). - to achieve this 4n s4fu burning of plutonium the reprocess- ing of the irradiated fuel 1is limited to the separation of all or just the most neutron absorbing fission products and directly coupled to the reactor - for the next generation of reactors the fuel is produced in the form of denatured uranium, that is an isotopic mixture of 10.5% uranium -233 (produced in this breeder) 89.5% uranium -238 (from the mine cor depleted uranium stock- piles) - for over 80% of the lifetime of this plant, that is 30 years, the reactor burns its own plutonium and produces the uranium -233 for the next breeder generation. Tables 2.13 to 2.17 glve the corresponding data for this reactor phase as before. 2.17.5 Comparison of the three phases In spite of this rather over simplification it seems that the praoposed reactor design 1s suited for burning: uranium -233 in th first phase uranium -233 and plutonium in the second phase plutonium in the steady state phase. (see Fig. 2.3) As a first approximation the timetable of such a reactor will be as follows - doubling time approximately 4 years (including the out of core inventoryl. - transient pericd (going from U-233 of plutonium] approximately 2 years. - steady state period approx 30 years including shutdown periods for the exchange of the gpre vessel etc. See Table 2.18 TITLE GATE : OROER % 3 QUADRATUME SO MU, I N"} ® ¥ » fr) aF GF LN o Wl £ Ut m R AR a#)?’f"’"n.”.’ RS 2o . T PRCER o o 1 : = o e ! M 5. i d W col A ‘ - e SN AR L e Y s, k] . s : N » H W o oE b s " ] ¢ e [ g ERVIRE R s . St S tu&wmwa-&*;._"“ . - ESPEC U S o 56 “.m,.’_fi'_")s:c Tl h‘fii%fifi?’fi*fig%‘fi*fi* ‘qah :f’(‘ &, r‘?"_&i,., @ om ok om m W W E o oeowm 7 - e T e | 3 o = ' A e - ti o s o S r * : £ E T e § o -y T - e - é(s:wtl nt Som * = am T4 2 e 3 ! - R o & e : A = . 3 N = e Rl n R PR = <, ORe e z T.X e = - Cm r N A4Are n e P el b I % L4 S’ TH232 5041["‘ S TITLE DATE o aw Table 214 UTHOST=302,MH0P, L B EEEAERE R R RN BN gy SPHERIC GWEOM, 03/0977¢ R R RN ECEREEER R RN R R g R I B o R g e s Y S B O#E # B # % % % # B = E oz ¥ o ¥ ¥y ¥ o ELEMEMT CORE Uegdd edn PUR39 PUZ 40 FilZ2 414 CL o BlLALKET TLT, F‘;LANO VR ATy, Lo A AGE T O 7T A L NEUTRON]CS MACR PR R R GNEER gy ABSQREIIoN 10.990F+16 760015E*18 58,061F+1 8 691115417 11, 8880418 i\stURSE"'la 29 .411¢c«27 11,833 +27 14,2016 10.9R75+27 %/ UEJ 12080 7941574 o c o L ™~ (TS ) & YO MY MM t 4 ¢ @ kl}\yl}lbl ~N O N J o 30.4485*16 /7.95{_6#13 n + 76,v9CEesd jobel - 14.536:_’-0':,_8 24.0":6&-&:‘_9 o4 J0,¢64 23,65 2079 4,59 5s 8 1,02 48 R A2 6R,G6 23.(1% . VY ce6Q 4,4C 12 oC< 31,04 He6T 10C,00 pReUCIIAN :’4.(336E4‘1() 65.720E+18 ' 40513E"'1Q ‘2.726E+18 ?n.t5obe1B - - Fm.14QE+10Q 4‘:0‘;'196*17 P 2. EE+L1D *r 8 & B B ¥ % B R N ¥ & € 4 X € ¥ « ¥ &£ ¥ K ¥ 14 - AR R RS R R R R R R R A T RS TR R N e S L R R T - 38 - Table Z2.15 UTMOST=35L,H0P, . EETETREEREE R LR JR Py TITLL ¢ SPHERIC wEUM, DATC U3/snGs 7y I ER R R E TR EE RS RE R R R EE RS AR FRREER SR AN S F IS F S S SN FEERRER EER NN * CNILIY - N1 - MIa . ¥ 1S B '.;Q - AL M - FASE - APt AR “ “tAG TRODUCTION TAST BONUS a0 «PARYS L AKAGE PUZG PRODUC FUZ9® DESTRD U3z PRODUC U33 DESTRO PALANCE U233 BALANCE PU24Y %+ # # ¥ % &£ # #£ # *® BN g RS R K WS W e Wk K ok ok W W o kW b e g o R ok o R R K «COMMEMT *FASE, N*SIGMA FISS,FERT / N*SIGMA F1SS,r ISSIONARLE, *AHPAR, N«SIGMA PAPAS / lNeCAPT (FFRT + r15S), TOTAL BREEDING HeyTRONICS MIcP S 2 N TR R R RN SXUR0E RPENLRATIO MICD BALA3S BALAJ® MACR 8BRTOTY *| EAG, LEAKAGE / ¢PRODUG, sFAST RONUS, ‘AHSOR.PARS, *LFAKAGE, LEAG#ZINIM / *RRTOT, M«SIGMA THRAMSMUTATION (TH =+ *CAPT, ERAKE R IR R E N R AN TR R g g R R B RN Aoy BN RN R kRSN & k% IOTAL AFSQR.(FROM (FASH = ABPAR / (1 « ALF), (1 + ALF), (113 « PU9 « Pyt), e 4+ PLD) VAaLUE T.187 3,020 2e454 2,515 2947 2,993 200 L 455 1173 » COH 1,501 AT REY. 144 1,951 5,371 5,666 630 137 5,402 -, 495 1.581 MaX, FLUX IS 1 2,84C+15 FOR GPOUP 14 N INTERVAL Tar| g NEUTROMICS)., (INIM e 1 « ALFM) ,/ (1 + AlLr), (MNIE = 1)) / (1 +« ALF), / NeSTGHMA, 1 < n » L] - " % L] | » ¥ * L w » * » » * * = *® ] * * HTHNgT=302,N0P;, L ERFEREEE N D R R QS gy TITLE DATE SPHERIC GEOM, R N B *m ®wa T T T N » & . cnrrzy SXHBOL UNTX MALUE * * . * GEOMETRY * | » * CORp RADIUS Bl M 2954 " * WALL 41 THICKNESS THICK {2y M 0410 . . PLANKET THICANFSS THIcK(Z)y M e 550 * » FALL 2 THICKNESS THICK{4)y M e 040 * * HETGHT 77 SYLINDER HEIGHT M w000 * » HETGHT/ZD 0y wATIg YDRAT - 000 * * VOL UMD OF (Owe VOLCOR Mw g 3 T.633 * . VOIL UME OF RLanNF[T VoL el Moyl 14.299 * b L * POMIR ® " » POWER 1M 10w AR IS Gw 2.722 » POWrR In HLANKET POTNK Gw 063 * » TOTAL Powee FOWRTO G 2.765 % * FOWER JENS, QURD rorco OW /My e 752 * “ POWED OBRENS, =LANKET TODEH CW/Muncz 2006 * * POWER RATING @ pyy "OLRAT MW /K0 2.570 * L » * THYENTORY * % * " TOTAL PUZ3¢ INVENT, "STQOT KG 736,542 * " TOTAL PU INVENT, cyTe T Kg 1063,452 * e (223 IN CORE var KG 1.035 * * U232 IN RLANRKRET v Keg 47 .3C7 » . TOTAL U238 INvEINT, VRTOT Kn 6763.900 - » TOTAL TH INVenT, THTAT K 23551 ,7E9 » * FPATIO U/ (Pu=ti33y TATUS - 6362 " L3 * . THEN GaTa * ] * * PR . AT IO fwildy T URMR - 50963 * ® LU TN T Mk T Ivk YE ARG 1.984 * . PEAN FLUX COKRE fLMEAN NEHT/S«CH2 1,460E+16 * "« * L EEREEEEELEEEEREEREEEEEESENEREEEEEREE RS FWCECENEE RS FEEEE R R R IS RN FFTEEAEEEENEER RS S EEE RN REEESER AR N FFESEASSENEEEEERE RS B BV * TARLFE * * "R RN R N N * * MATERIAL RALANCE OF THE BnEFNnFR SYSTEM * * INCLUDING NEW REACTUR * PRI YR E AR R R R R R E SRR TR SRR E RS R TS SRR SRR RE SRR R S * " THOR . pA - uesad * ur3e - PUZ® * A L I A E F E F E N T e A R R S Y R LA RS2SR X R TR I M) *0OKS3 1.26E+04 AeBICT2 6,7 E+nn Ae04F+03 6,/ OF+(0>* * Sl 2. IBE+N4 R,680+"1 4,730+ " *FHL Q.55E+00 P417E+70 1,540+04 " *FTR 6.83E+12 Ay7Al+r2 6e6OFE+02 . »S0 1,03E+0n e 76E+03 7e37E+0n ¥FEP De78E+02 7.34E+00w *FCR 6eP4E+2 Ged47E+02 * *F LR 2e30E+ D B e68F=r 2 4,73C=n" 6e76E+00 7¢37E=04" aSii 2.30E+0)0 5 6B =t ] 1.03F =07 te76E+00 7e37E=04 = *FwA feF4E= 31 6,500 ™04 Hel6E=04 7:37E=0>2* *E AW {.19E+;4 6,56 +n> 2.c90+02 * #S MW DG 30E+i1 4 1,010+ et 3E+NT " g L R s A A T T T R S IR A R T IE E Y WHFRE ORrRS SHL FRL FTR SC- FF# FOw F ik Skt FuA F M SN W " 8 R 3 0N 1 n STRADY FLOW FLOW TO FLOW FROM STEADY STATE IN THI FLOW FISS. STFADY STATE IN FLOW FISS.+ PARASITIC FLOW TO THE CORE OF [FLANKET IN g , YEAR FLOWTO REPWUCESS AMD PETURE IN STATE IN THI TO THE THE I THE I QRFE THI NASTD 1IN MEW TN KG NEW AFTER DT Inr I OR SYNTH, IM GLANKET IN «n, + PARAS.IN BLANKECT FLOW FOR TRaMSMUTATION o / YEAR KG 7/ YEAR I FG 7/ YpaPp CorRE IM Kg N CNRRE FC=PROCESS1Ne K G / YEAP / YLAR It TN OKG / YEAR / YEAP IN KE Fig. 2.3 Fissile atom concentretion in the core R = constant P * constant (:E:) (:E:> } nn*%&?&1w' Yy \ NOPU (V17] B 5.73 - ___—_—__—_—_ 5.68 . s - H3 2 1 T T T T T T T T T 0 0.5 1.0 Pu23§ Pu239 430 kg Pu241 U233 499 kg U733_800 kg I« 843 kg T+ 839 kg L - 800 kg Fig. 2.4 Thorium and uranium-236 breeding {NOPU-21) P R g L : ///// Bybedi in -23 % 1 /] / U'I i < r U-238/Pu-2%9 ratio in core Table 2.18 Core Diameter = 0.95 m Power = 2.8 GWl{th) Build-up Transient Steady State A B C Figsile - U233 U233, Pu233 | o 5ag 4 o4y + 247 Core inventory ke 800 409, 430 840 Fertile - U238 U238 U238 Core inventory kg 6864 6978 6763 Fiasi] 2227 (enrich- % 10.4 1.7 11.0 Fertile ment ) Hreedi i reeding ratio AR 1.34 1.45 1,58 total tot Mean reutron Flux [1nte DELErons 1.26 1.33 1,06 cm2s GW(th _ Power density (Lherm) 0.75 0.745 0.75 m core It should be emphasised that the era when fusion reactors are raeplaced by other energy sources and hence the plutonium burners are shut-down, has not been dealt with. bven here it is conceivable that a reactor design could be procoe- cd which only consumes and thus be used in this shutdown phase. This however has not been calculated. 2.2 Impact of some parameters on the mixed zone two zone fast breeder In this section some intermediate results are presented concern- ing the influence of selected parameters in this design, and giv- ing the breeding capability of a two zone fast breeder reactor. (Table 2.19) The influence of mutual displacement. T(Pu -239) + T(U-233) = 9 x 10°" (10°*atom. cm~ ) by five steps from A to £ 1s given in Table 2.20 and Fig. 2.5 and Z2.6. The influence of the thickness of the external blanket, which contains thorium, on the total breeding ratio and the volume of the core for a given case of the following concentrations of fissile nuclides PU-239 3.5 x 10°% (10%%atcom. cm™?) U-233 10.0 x 10°% (10%“atom. om-?) is given in Table 2.20. For the same case the partial BR's and BR total are shown in i) Fig. 2.7 and Fig. 2.8 In addition the followling 4 reactor designs have been calculated having the following arrangement Core Pu-239 1.1 x 10% —> 1.8 x 10" (x10°*atom. cm™ ) U-233 3.4 x 107 %— 5.5 x 107" (x10%*atom. cm ?) 1-233/Pu-239 = 3.1 Blanket thickness 120 cm. The data are given in Tab. 2.217. Other results are found in Fig. 2.9, 2.170 and Z2.117. Tahle 2.189 OBJECT: Plutonium burning «4n s4tu: some parameters. REACTOR TYPE : Power, Breeder, GEOMETRY : INTERNAL ZONE : Spherical core, WALL : v 4 om INTERMEDTIATE ZONE - WALL : v 4 cm EXTERNAL ZONE : Blanket: 100 cm WALL, REFLECTOR : Iron,.... POWER (CW thermal) 3 GW POWER DENSITY (GW therm/m®CORE) NEUTRON FLUX, MEAN (n/cm?s) .2 FISSTLE NUCLTIDE - Core - PuZz39d : Out blanket - U-233 FERTILE NUCLIDE Core - UZ38 0Out, Blanket ThZ32 DILUTENT : Chlorides COOLING SYSTEM : External to core BREEDING RATIO : PARAMETER STUDIED : Ratio UZ2368/PUZ3Y in core : Ratio PUZ38/0U233 FF Concentration Wall (Molybdenum) thickness Hlanket thickness METHOD OF NEUTRONIC CALCULATION : see 1.10 (page 17) Two zones Nvobh o em Table 2.20 Mixture of Pu-238 and U-233 1in the core Pu/U3 Ratio Py gnly PU g U3 | o 5 U3 Pu S U3 3 Enly Fission/sec 2.74x10%0 2.6x1029% | 2.51x10%°% | 2.41x10%°%} 2.32x207°0 Pu-238, core, 3 9,107" 8.10:2 4.10:: 2.10:: 0.0 3y U-233" x10%%at/cm®)| 0.0 3.10 5.10 7.10 0.10 AR total 1.76% 1.64 1.56 1.49 1.42 Partial BR U/Pu 0.353 0.312 0.289 D.271 n.253 Partial BR Th/U-233 1.408 1.33 1,27 1.21 1.16 Flux tot. (ncm ?s ') | 2.05x10'¢® 1.99x101§ 1.94x10'% ] 1.86x101%) 1.76x101° Radius, (cm) 67.28 56.22 65.72 65.5 £65.8 Volume, (cm) 1.28 1.22 1.19 1.18 1.0 Spec. power (GW m~3) | 2.34 .48 2.52 2.54 2,57 Ratio of partial 3.99 4,76 4,39 4.48 4.48 breeding ratios “ission ratios %igggfi 0 0.89 2.28 5,40 = Fig. 2.5 Thorium and uraniue-238 breeding (NOPU~-22) BR 1 n ., w W NV 27/SNNN\ 1%, NNNINNY, NN - SN \ N Ny N E 0.0 P T Pu-239 9 8 7 6 2 1 0x10"*x10*%/cm’ U=-233 0 1 2 3 L] 5 7 8 9x10"*x10%*/cm?® Approxunatelly Fig. 2. Radius of core versus radius of blanket (NOPU-23) Py = 3.5x10°° 2 , T 20.0“0..) x10?atom/em’® 13 p 1.3 4 Breeding 1.2 ratio 4 1.1 4 BH tot 1.0 ¢-9 A Radius of core, cm 0.6 - 4 50 e 8 - < "1 Radiusjof core LT3 v.7 - LL) < N2 U.6 40 T 9 20 40 00 8o Radius of blanket, cm Fig. 2.6 PRatio of breeding ratios (NOPU-22) Ratio Pu-239: ) ] | | / G Fig. 2.8 Partial breeding ratio (NOPU-23) Pu = 3,5x10"° uzs = Ratio of Breeding m]tio I ! T h-233/0}238 { Ratio qf Burni ratio | ¢ e Ul-233/Pu-2 -1 Ratip of ’ Breeding 4 ratio / U-233/Pu-239 e ! " B [+ / - * T T L 1 T T 9 8 7T 6 5 L} 3 2 1 10 OXIO"} xlO"/om/cm’ o/ - . E BR Uranjum 4 = T 0 20 40 60 80 Radius of blanket, cm Table 2.21 47 - Influence of Pu/U33 ratio FarameLers A B C D U-238 1.0 0.85 0.85 0.85 PU-739 - 3 1.8x10"" 1.5x10"" 1.3x10-" 1.4x10°" 0%%at | pu-24g X107 atom/cm 2.5%10- % 2.1x10- 5 1.8x10-5 1.56x10-5 U-233 5.5x10°" 4,7x107" 4.0x40"" 3.4x40°" U3/Pug ratio 3.08 3.13 3.08 3.09 BR total 1.5 1.51 1.52 1.528 Partial BR thorium 1.12 1.04 0.85 0.852(see fig.2.9 Partial BR uranium 0.39 0.47 .564 0.6/77 o breeding ratio 2.37 2.21 1.68 1.26(see fig.2.10) Flux tot (nem-%s-1) 1.3x10%°8 0.92x1018 N.62x10t6 0.463x101° Radius of core, cm 78,06 4941.4 10/7.7 122.0 Volume of core, m3 Z2.03 3.7 5,23 7.6 Spec. power (GW m™3) 1.48 0.938 0.574 0.39 Thickness of blanket m 1.0 1.0 1,0 1.0 Volume of blanket m3 80.6 93, 4 109.7 124.0 PuU-39 in core, atoms 3.65x10%° 4,8x10%°8 £.80x10%¢ 8.37x10°%° mol 509 800 1130 1400 kg 146 191 271 333 J.33 in core, atoms 1.12x10%7 1.5x10%7 2.09x1027 3.50x102%7 mol 1860 2510 3490 431G kg 434 584 512 1000 Th in blanket, atoms 1.01x10°°2 1.21x10%° 1.49x10°° 1.806x10%° ' mol 169000 202000 248000 203000 g 39200 45800 57600 58000 J-238 in core, atoms q,54%x10%7 1.5x10%8 2.46x10n%8 3.568%x10°%8 mol 15300 25100 41000 596000 kg 3780 5970 a750 14200 Th-32 : T inventory ratio 10.6 8.1 B, 0 4.4 Jurning rate Py-239 in 107 atom/s 1.44 1.39 1,39 1.35 U-233 in 10'7atom/s 5.08 5.1 6.08 6.09 U3 Ratio-ga burning 1.30 Fig. 2.9 Partial dreeding ratio (NOPU-24) -238 Concentration ratio Y-233 5 3.1 . BR total Pu-2139 1.5 1.0 7 4 BR 4 0.5 =1 4 4 p 0.0 v ’ ’ v 4 & i 7 : 20 -3 concentravion of fissile, 10 atom cm Fig. 2.17 Ratio of burning rate snd ratio of breeding rate 5" U-~2 Ratic of burning rate Pu-239 4 od 37 A Hatio off r;me-ding“M - PR g L | Fo=23g . Fu-2 3¢ PRt - ! o | i ! 3 | i 1 !C B A i 0 T T T 4 5 T 20 -3 concentration of rissile, 10 atom cm —4 Fig. 2.11 Impact of aome selected parameters (NIOPU-26 reference) 2 em wall S Mo + Pe 1.u5 = referengs 1.60 < smaller 1.50 — RS blanket 80 cm 20 times more ¥.P. 2.3 Material balance of the state reactor The most important feature of this reactor is the (n s4fu consumption of plutonium. The studies show this to be feasable. Fig. 2.12 gives the resulting material balance for the steady state reactor. 2.4 Conclusions The molten chlorides fast breeder is ideally suited to the (- s4tu burning of plutonium matching the current requirement for a "proliferation safe” concept. _50_ Fig. 2.12 S S ORES L L L 12600 4240 kg N * N \ 23600 |» N\ ' ‘ Th 9; 388 BLANKR ? o @ o N L6760 560 Pa) 9.17 & tfpisgf : 27 N 734 U3)15.1 N\ 757 |- — RPN N I P B :: | i ) t: chE AN : N\ N WASTES Th: 23600 kg U3: 1310 kg Ug: 6530 kg AFTER DOUBLING PERICD 3.7 THE THREE ZONE REACTOR 3.1.1 Introduction The reactor discussed now is rather unconventional because of its three zones (see Fig. 1.2]) - internal blanket zone - fuel zone - external blanket zone This concept has been compared with the more conventional type. Holding most of the parameters the same the breeding gain comes out about equal. However, one large difference is that although the total power 1s the same the specific power changes by more than one magnitude being higher in the conventicnal central fuel zone reactor. Also the mean neutron flux increases from 1.7 x 10'%n em™%s”™! for the non-conventicnal central blanket zone to 2 x 10'"’n em™?s~! for the conventional central fuel zone. Since the specific power and intensity of neutron flux 1s clearly a major pronlem form the point of view of the englineering design of the reactor (cooling, radiation damage of structural materiasl and fuel), both systems have been studied, that 1s without a cen- tral blanket region and with a radius up to 110 cm. The results are given below for fuel without uranium and with uranium fuel for both cases: no internal blanket zone and with an inter- nal blanket zone. 3.1.2 The three zone breeder with thorium/uranium-233 The first case is the fast breeder molten salt reactor with uranium-233 as fissile and thorium-232 as fertile material with a three region layout. - internal fertile zone 1, 50, 90, 110 ©m - wall 2 cm - spherical shell core 19 cm - wall 3 cm - external fertile zone 110 cm - reflector 40 cm Table 3.1 gives the main details of the reactor. L) .2 A Three zone breeder reactor with a mixed fuel cycle U-238/Pu-239 plus Th-232/U-233 The next step is a study of a mixed fuel cycle in the three zange fast breeder reactor. This problem has received much attention and the published papers given in list D should be referred to. A three zone fast breeder reactor with the geometry shown in table 3.1 has been calculated. The range of variations covered include Pu-239 or U-233 in the core U--38 Th-232 in the blanket For a given geometry and concentration of ertile and fissile nu- clides the influence of the fission products, when the concentra- tlons are increased by a factor 10 results in a reduction in the breeding ratio by 5%. An increase by a factor 10 in proactinium decreases the breeding ratio by only 2% (Fig. 3.4). For the three zone reactor calculations have also been made far mixed fuel cycles. For the fissile materials Pu-239 and U-233 and for fertile materials with U-238 in the core only and Th-232 in the blanket only. Table 3.1 OBJECT: Three zones thorium cycle REACTOR TYPE : Power, Breeder GCREUOMETRY « INTERNAL ZONE : Spherical, fertile material (™) WALL : Metallic INTERMEDIATE ZONE “hell; fissile material, active core WALL : Metallic EXTERNAL ZOMNE : “hell, fertile material WALL, REFLECTOR : Metallic 6 A6 - 9.hY 10 POWER [(GW thermal) 2. POWER DENSITY (GW therm/micore) : 0. NEUTRON FLUX, MEAN (n/cm?s) v FISSILE NUCLIDE : Internal Blanket : Core - Pu + UZ33 Outer Blanket U233 FERTILE NUCLIDE : Internal blanket Core U238 OQuter Blanket - Th 237 OILUTENT : Chlorides CODLING SYSTEM : External BRECDOING RATIO < 1.08 - 1.14 PARAMETER STUDIED Thickness of blanket Thorium concentration in balnket material FP and PA concentration METHOD UF NEUTRONIC @ ANISN, 54 23 Groups CALCULATION : 060 spatial positions P1 approxim. GGC-3 code ENDF/B-1 and B-2 Dats Table 3.2 (THOC-300) Three zone reactor. Volumes and breeding ratio Geometry: internal blanket, radius 110 cm wall, width 3 cm core, width 27 Cm wall 3 cm external blarket, width 100 cm wall reflector 140 cm /ariable Results Internal | Internal | U-233 concen. Yolume Specific Oreeding Case fertile wall in core of core DowWer ratio, (cm) (cm) (107" /em?) (m?) (GW m~3) total B 1.0 1.0 0.0012 0.783 3.36 1.08 C 50 2.0 0.0012 1.66 1.57 1.10 D 890.0 3.0 0.0017 3.73 0.70 1.13 E 110 3.0 0.0017 5.67 0.46 1.11 A 3.0 roference 110 0.0C18 3.13 0.80 1.16 f: 110 3 00,0003 9,57 N.28 1.4 see Fig. 3.7 (THOCL-300) Three zone reactor: ataomic composition (atomic concentration x 102%) Internal External fortile zone Wall Core Wall Dlanket Reflector o 4.5x107° Fe 7402 | Th 2.5x107° the same Fe /x107° L 1x10-" Mo 1x10-7 Pa 1x10°" as inter- Mo A1x10-7 =33 1x107" =23 1x1n-? nal c1 2.2x1077 -2 axant® nlanket (remark: Ha 4.0x107 3 wrong variable variable Fe 7x1n~7° reflector radius 1-233 con- Mo 7x10° 2 with Mo!} (see Fig. 3.1) centration (see Table abovel. .1 Impact of the radius of thre internal 55 olanket (ThCL-300) on the breeding ratio 1,10 - ¥ | BR * breeding ' ratio o il ] y 1,05 "/c i C 1,0 T T T 1 T T [+ 20 &0 100{ 120 internal blanket rgdius, cm & calculated cases: A, B, C, D, E Tpecific 9 rower (cwem™ ) 1 c 5 — 4 2 - E (more U-33) 4 J 1A l A (reference) 1 ! (less U-33) [¢] Pip., .3 Uranium concentrations for given geometry (THOCL-301) for A, B, C, D, E, F, G, H, T, 1.164 Th mol ratio p 0.6 4 o 1 0.5 1.15 D r 1 A g pe BR 0.4 < - 1.10 Power rating 1.5 J - - 0.60 1.3 > specific specific - power, P e ] | 0.50 = kg U-2 | MW tot - - owe 1.0 JNFning L ob e 4 - 0.8 4 I 0.7 — ——r—r 0.3 0.17 0.21 0.25% U-233 mol ratio of the Rate of breedins Pig. 3.2 Concentratign of thorium and uranium-233 (in 102:7“3) {THOCL-301) 0.0048 ] 4 H mol ratio T™ = 0.6 1 - 4 0.0040] P— E T™h d ol ratio \ ’ a ™ * 0.5 b 0.0035 4 1 - mol ratio Th = 0.4 - \{ 6.0032 T T T T vr% 0.0010 0.0015 0.0019 J.0021 Uranium in core Pig. 3.4 Three zones, mixed fuel cycle Impact of fission products and protactinium {THOCL-305) Rate of breeding 4 1.55 4 J ~ ratio 4 1.50 + st Paom 1.45 ~ = 0.18 1.%0 1.35 o 4 1.30 10 times reference less PP FP = 1 Pa = 1 Pa = 1 Table 3.3 - 56 - (THOCL-302) Influence of FP and Pa concentration gecmetry internal wall core wall external blanket balnket 110 cm 3 cm 19 cm 3 cm 100 cm reflector 40 cm Fission product Protactinium concentration Breeding ratio content corresponds to dwelling time, Case days. 1024 -3 * e 2 5 18 A 3 x 10°°, 1 x 107° 1.11¢ B dwelling time = 3 x 1077 1.124 C 3 days 9 x 10°° 1,123 5 x 1077, 0 dwelling time = 1 x 107° 1.1238 10 days fon) S also Fig. 3.4 Remark Tabhle 3.4 cance led | [ ~J4 I ihe influence of parameter variations 1s given in Fig. 3.5. The following results can be noted: the breeding ratio increases (under the given conditions) when - the concentration of U-233 in the core decreases and U-238 increases - when the internal wall thickness 1s halved (note this is very sensitive due to the presence of 15 atom% of molybdenum) - when the outer fertile radius increases - when the inner fertile zone radius 1ncreases. 3.3 The three zone reactor - uranium-plutonium fuel cycle The reference reactor i1is described in table 3.5 and 3.5. The thermal flux in all three zanes, the external breeding zone, the fuel and the internal breeding zone is aonly 10°°% of the total flux and in the external hlanket reaches 10°% of the to- tal flux. The total flux has a relatively flat distribution and even in the fuel region the max. to mean ratio 1s only about .13 (Fig. 3.6 and Fig. 3.7). The neutron flux 1s rather hard and the mean neutron energy (calculated as the mean of the no. of fissions) is around 370 keV {see Fig. 3.3). In a typical LMFBR and in a gas cooled fast breeder this value is 120 keV and 176 keV respectively. A pood illustration of the influence of the most iImportant papameters aon the breeding ratio is given in table 3.7. The differences between these calculations and the computur cut- put is approx 8%. Fig. 3.5 Three zones; mixed fuel cycle Impact of selected parameters on the breeding ratio {NOPU-1) wall 1.5 cm h K ——— 1.7 o greater ouf. D gblanket 125 om BR ] greater inner covmm— p P4 vlanket 150 cm ) T™/U in more U-238 in core, o _ 1.6 inner blanket no U-233 more U-238 1 C erm—— half of i u-233 reference 1A out blanket 100 cm wall 3 cm 1.5 M inn. blanket 110 cm Fig. 3.6 Three zones: uranium-plutonium fuel cycle Neutron Flux in reference reactor fotal neutron llux L~y 10t d 1647 Fast ne (~370 KeV) 101"_ (Neutraqs) cm< sef 1015.' 1012 Internal fertile zone wall 11 10%% ] el pond 1010_ WLl External fertile zone Reflectdr 1Y ‘ b ‘nermal neutrons o flux IUU - 10 T ég.r Tt , 20 40 60 B0 11)" 14C 160 180 200 260 Radius {(cm) - 59 - Table 3.5 OBJECT : Uranium-plutonium cycle - Maximum breeding ratio in three zones REACTOR TYPE : Optimised breeder CEOMETRY : INTERNAL ZONE : Spherical fertilile zone (M) WALL i Iron, Maolybdenum INTERMEDIATE ZONE : Core, fFuel WALL : Iron, Molybdenum EXTERNAL ZONE : Fertile zone WALL, REFLECTOR ¢ Iron POWER (GW thermal] 1 b POWER DENSITY (GW therm/m® CORE) : 1.1 NEUTRON FLUX, MEAN {n/cm®s) couinte - oty FISSILE NUCLIDE Internal Blanket: small amounts of FPuZ3Y, Core Pu 238, PuZ40, PuZd1, O.7: /.2: 0.7: External Blanket: small amounts of PuZ3%, FERTILE NUCLIDE : Internal Blanket UZ38 DILUTENT : Chlorides, Sodium COCLING SYSTEM : QOuter HREEDING RATIO . 1.864 - 1,044 PARAMETER STUDIED : Flutonium-uranium ratio With and without uranium Refloctor, Fe, Pb MEHTOD OF NEUTRONIC : see chapter 1 CALCULATION : Table 3.6 Three /ones reactor: (200/C) uranium-plutonium fuel cycle. Radius Width /one Compoailion Flux Specific cm of atoms/10%“cm’ thermal; power Z0ne total, GW/ma, cm Sreeding Temperature: roa-io 0 I 1-238 6.4x10°° et Pu-239 B6.0x10-° 1.05%x1016 lnlpt—7naoc - ) h*S el 0.0 fortile F.P. 2.0x10 3.7x10° . " one Ma 3.4x10°° 1o SB00°C : C1 2.27x10°3 BR 0.490 OULLE 110.0 - IT Fo 7x10° 7 1.10x100° 2e0°c o Wall Mo 1xq0-2 Ix107 R 113,10 I11 Pu-233 1.3x107° g ny-240 4.2x107" 1. 1GW/m 17.9 Fuel cy-241 2.9x107" 1,02x10L® _ =750°C o 3 —— i inlet zone =238 4.7x10°° B,.Ex10 A=E 2.0x10° T . _=1050°0 N 3.4x107 3 RR 0,22 outlet ] 2.6x10°7 130. 9 TV Fo 7.0x1077 8.24x10%° .0 - r 3 Wall Mo 1.0x10° 7 2.4x10° SR 133.9 v the same as 3.6x101" T, 7ot Uxternal central fertile 1.0x107 HhLet 1000 fertile “one, 1 _ ZSOUOC ~one AR: 1.040 cutlet 233.9 VI e 8.0x10°7° 5.2x10%7? 40 TrITETT Hx10 273.0 Fig. 3.7 ‘hree zones: uranjum-plutenium fuel cycle Total flux in the ruel zone Internal fertile \j Fuel zone \* zone i\ 1 1,0 . \\ '\ 1,1 = Neutron flux 1oke .2 - nem s \ 1,0 = :::: - \s [T \ ~ Fig. 3.8 Three zones: uranium-plutonium fuel Spectrum in fuel zone lo(é_ T 1041 1029 | - 4 1019 N Median < < 10i5 L It I i Y " 4 T T L) . T T 1 102 103 10 10° 10 107 Neutron Energy (eV) Table 3.7 Three Zane reactor: uranium-plutonium fuel Simplified calculation of the breeding ratio and neutron balance. Component of Parameter . . breeding ratio Median energy Kol 370 Pu-239 of{barn) 1.83 (from computer output) 0. {barn) G.180 V 2.95 o 0.0984 Nreooding potential (n-1) 1.6857 1.685 Fission ratio S 0.37 fertile/fissile (v-1) Fast bonus 0.539 0.538 1+ Total positive 2.275 Losses (sbsorption in 0.160 P, C1, Na, Mo, Fe) ‘ Leackage (arbitrarily]) 0.10 Loss+o Total losses = 0.32 0.320 1+0 Calculated BR {(micro) 1.830 Computed BR (macro) 1.752 - 63 - 3.4 The three zone breeder reactor: Very high bresding gain Cne of the most important factors in achieving a very high breed- ing ratio i1s the hardness of the neutron flux. This is strongly influenced by the fuel compositicon. In this case the fuel is postulated to be a mixture of a PuCly - b NalCll - ¢ UCL where a = 0.1 - 0.2, b = 0.7 - 0.8, c = 0.1 - 0.2 Untortunately not all data are available for this system. (see Fig. 06.9) The rough calculations on changing the concentration of PuCl, in the melt of NaCl (Fig. 3.9) shows a rather sharp decrease of breeding gain (BG) for decreasing plutonium concentration, espe- cially when the plutonium molar ratio to sodium is lower than 025, In spite of these uncertainties of the PuCla-NaCl-UCly system, the influence of the U-238 in the fuel has also been calculated. For a constant PuCly concentration with a simplified assumpticn for the NaCl concentration the results are given in Fig. 3.10 and 3.11. Increasing the ratio of uranium to plutonium in the fuel from 0 to 3 causes the total breeding gain to increase from 0.55 o .95, This 1s rather clear and thus the reference reactor con- cept includes uranium In the fuel in a ratio of 2 : 1 to pluto- nium. Such a high breeding gain is a special feature of this type of reactor for producing large quantities of fissile material. Fig. 3.12 and 3.13 give the results of calculation when the ra- dius of the central fertile zone is varied. Table 3.8 shows the simplified calculation of central and external fertile zone brecding ratins. - 4 - fig. 3. Impact of Plutonium Concentration Fig. 3.10 Three zones: uranium-plutonium fuel cycle Impact of U-232 concentration in the fuel Pu-Concentration: 0.0021x102* atom/cm’ ouD g Pucla 1 - NacCl 700 - Tempegature ompeyatu 0.9 (c) 600 - Breeding gain 500 0.8 4 400 NS N S MO SEN SR 1.0 0.7 0.5 0.3 0.2 0.1 0.0 Mol ratio of plutonium 0.7 0.6 0.5 = Specific 0.6 breedfng power gain (kW/cm”) ). o o 1 U/Pu ratio 2 3 L 1 i L 0.8 . 4 —i L— T— 0.7 ) 1 P4 3 L 5 6 *-¢ 24 Atom U in fuel / (0.001:10 /cm,) 0.2 0.5 0.7 0.5 0.3 mol ratio of plutonium Pig. .11 Three zones: uranium-plutonium fuel cycle Fig. 3.12 Three zones: ursnium—n%utonium rue} gele " Impact of Uranium concentration in the fuel Impact of central fertile zone radius (202/201) versus central fertile zone (No U in fuel) 1.0 o Uranium 0.y Concentration x10*!* at/em? o Breeging ‘vith u "5_3 103 gainjtotal 0.5 o Core 0.9 vl 1 p 4.2 10 Core 28.0 0.7 ] 133 . Breeding Gain ore ° L 16 0.6 =15 0.8 o 3,15 16° 1 Breeding gain -13 -2 ¢ -3 2,1 10 <11 Spec., 4 power 0.7 Kw/cm™3 ) 0.0 ‘Mm ¢ 0.6 Core 200 5 Core Ve o 110 cm T T T M M T v T M Y T 9 10 20 30 40 50 60 70 89 30 100 110 Central fertile zone, ratios (cm) Central fertile zone, radius {cm) - bh - From these results the following conclusions can be drawn - 1increasing the raius of the internal fertile zone up to 110 cm increases tLhe breeding gain for a given type of fuel. The ef- fect of wall and fertile material changes are insignificant - at the same time the specific power decreases dramatically - increasing the U/Pu ratic from 2 to 3.6 does not influence the total breeding gain (see Fig. 3.13) - the FP concentrations play a rather significant role (Fig. 3.74) Table 3.8 T i Internal fertile External fertile OLa? Fuel zone breecing zone zone , ratio Case Pu-239 Uf, qU Pu—239jc Pu—241F Pu-2349 U{_ ] (Number) iss cap f fiss cap OXV OxV Ox\V OxV OxXV Oxw three zones | U.14 10.308 0.8 3.05 0.351 0.30 0.47 1.83 1.70 (200) two ZOnes - 3.63 0,367 .04 0.46 2.50 1.53 (180) The influence of the 40 cm reflector if changed from iraon to lead is not very great as shown in table 3.0 Fission product concenfiration however plays a very important role. For a given reactor design and given fuel and fertile compositions, increasing the concentratlion of fission products (simulated here with Cs-133 only) from 2x107° to 2x10°% (in 10%%/cm?®) decreases the breeding gain from 0.85 to 0.38 when the specitic power docreases less than a factor two. Table 3.9 Central fuel (Core 180) (wall 2.5 cm Pu = Three zone reactor: — 6 6 — uranium-plutonium fuel 2.1x10°3% x atom/10%%cm?) Case A B C Uranium 238 in fuel no VeSS no 4,2x107° x10%%atam/cm’ Reflector 40 cm: material Fe Fe Pb Volume fuel x 10°cm’ 2.95 Z2.40 2.97 Spec. power in fuel kW/cm® 18.4 23.0 18.3 Hreeding ratio total 1.54 1.394 1.66 Tot T Z20Ne 2 y 2l Tl of fuel zone 1.18x1017 1.25x1917 1.187x107 n/cm’ s Three zoves: ararium-plutonium fuel Breediry ratio versus radius of internal fertile zone Fig. *.1° Breeding T 50 100 Radius of internal rertile zone, (cm) Fig. 3.14 Three zones: uranium-plutonium fuel gele Impact of Fission Products Concentration in fuel (very simplirfied, from different calculations) J,7 - breefing gain B.G. B U,b S.T7. (kw/em?) =3 0.5 -2 S.P. 0.b dwelling time of fuel in ¢ore (days) i v.3 t-2 ~10 "20 T . 107> 127 Concentration of F,P, atoms -« 10° /cm - 6/ - In the steady state reactor a concentration of 2x10°° atoms F.P. x 10%%/cm?® for a fuel having 2.1x1073 atoms Puxi02%/cm? is reached for a specific power of 2 GW/m® after a time t of _(2x107°) x 102" ) . b STy (5onio) w2 C 1-61 x 107 s that 1s after 1.87 days. The higher fission product concentra- tion - that is 2x10-* corresponds to 18.7 days of mean dwell time f or the fuel in the reactor, The influence of chlorine-3/ separation may now be looked at. The influence of each absorber on the breeding ratio is given by - A+ D+ L + T+ o B = decrement of breeding ratio A = absorbtion rate in a given absorber absorbtiaon rate in rest of absorbers J 1 L = leakage o = Oc/ojc It can be postulated that for a strong absaorber in a hard [(fast) spectrum that A - 0.15 0 + L - UJ.15 o = 0.15 The relative influence on the rather high breeding ratio of 1.6 results in a case where the profit of the separaticn fac- tor will be for example 0.2 then and In relation to the breeding gain AG —4—7 = 0.20 - 606/63 - which results in an increase in doubling time of T2 ) NI = .83 It can be seen that introducing three zones does not result in any significant increase in the breeding gain (table 3.10}). Therefore the two zone reactor must be preferred for its simpler layout. Table 3.10 Fuel in central zone (2 zone reactor) versus fuel in middle zone (3 zone reactor) iwo zones Three zones Core, Case Nymber {conventional) ‘nonconventional (1230 (2nn) eometry Central Zone Fuel 100 cm Slanket 110 cm Middle Zone -—- Fuel V18 om Outer Zone Blanket 100 cm Thermal power, GW 6 B PLU/FP ratio 2.1x1073/2x10°°% | 2.1x10-3/2x10° ¢ Spec. power, kW/cm® 17 .7 1.41 Power in fuel, % 90.9% /B.2% Flux total left 2.04x10"7 1.2x101° ‘ , boundary ———— in fuel right 1.15x1017 1.08x1016 Flux in left 8.99x10'°® 9,7x10'° outer } boundary blanket right 2. 16x101° 1.5x101" Breeding gain 0.63 0.70 Median energy (group) g 10 3.5 The two zaone fast breeder. Fuel of uranium plutonium flucrides 3.5.1 Introduction The aim of this section is to give a rough idea of a fast breeder power reactor having the fuel in form of plutonium trifluoride in the molten state instead of molten chloride. (Table 3.11) The earier suggestions for a reactor of this type came from A.M. Weinberg. The first attempts at carrying out calculations on a reactor of this type were not successful because a fuel was chosen having a high concentration of light metals, lithium and be- ryllium. A very rough attempt by J. Ligou and the author (1372) shows the possibility for a fast breeder reactor with molten pluto- nium fluoride where the light metals were eliminated and the melting point increased. F. Faugeras (Fautenay aux Roses) claimed that the fluoride of U-233 and Th-237 can be used for non-thermal reactors. Some preliminary results for a three zones reactor are eciven in a short form in Table 32.12. The neutron flux remains rathe- hard (Fig. 2.15]. 3.5.2 Arbitrary assumptions and uncertainties The fuel composition has been arbitrarily chosen since the appropriate data is lacking in the literature. In most cases the following fuel composition has bean used 1 PuF, / 1.2 NaF (T i © 727°0) + 2.4 NaF /1 Ir F, (T 1 7 510°C) another alternative would be PuF, / 2 Naf CaF. / 3 NaF (T = 820°C) % melt Table 3.11 OBJECT Fast Breeder, Molten Fluoride, REACTOR TYPE FPower GEOMETRY INTERNAL ZONFE (M) WALL INTERMEDIATE ZONE WALL EXTERNAL ZONE WALL, REFLECTROR POWER (GW thermal) : 6 POWER DENSITY (GW therm/m’CORE) : 3.5 NEUTRON FLUX, MEAN (n/cm?s) : o 1nte FISSILE NUCLIOE Flutonium. FERTILE NUCLIDE U-2368 DILUTENT NaF, 7rF, COOLING SYSTENM Cbornal HREEDING RATIO 1.38 (up to 1.0 PARAMETER STUDIED Wall: beryllium, Fission product METHOD ON NEUTRONIC CALCULATION FLUDRTOE+ Fission product only as Cs-133 Three zones Fuel caomposition PuF wall structural material: Iron, recalculated from Hansen-Roach Fertile zone Iron, Graphite Fuel, fluoride Iron, Mo, Craphite Fertile zone Iron a7 Naf, 4 with higher specific power) graphite thickness concentration ANISN Regions O Meshes 110 Order of quadrature 84 Anlstropy P1 23 neutron groups incl. thermal ENDF B/111 -~ /7 Table 3.12 Design of fast breeder molten fluoride reactor (Three zone: 6 GCWithermal]}) Power rating, total = 1 kg Pu/MwWwth, Doubling time = 5.5 years. Radius Zone Components total Broedine cm molecules Flux ————— L per cmix102Y thermal ratio 0 I | Internal | UF, 5x107° 1.6x10'° 0.42 Blanket NaF 5x10°° 5.8x1012 ) liquid Pufg Bx10°° T = 800C state 80.0 I Wall Fe 7x107 2 1.86x1018 - - Mo 1x19°2 2.2x1017? (graphite is also possible) 81.C 11 Fuel Pu-239 F3 1.47x107% | 1.57x10"° “ertile liguid Pu-240 T3 4,2x107" 1.00x1016 material state PU-241 F1 2.1x107" . B N R NaF 7.5x1073 "inlet O” Irf, 5.1x1077 T_. T 10s07C) 0,056 F.P. (Cs133)| 0.2x10-° ou 98,2 TV Wall Fe 7x1072 1.37x101 8 - Mo 1x10-2 5.2x10172 (graphite is also possible) 100. 2 v | External | U, 5x107° 8.8x10"° n.sa9 blanket NaF 5x10-3 2.5x107T liquid | Pufjg 6x107° T = 800°C state Mean 200.0 S 7.5x10%! - VB2 — . 5. 3% 106 240.0 Total TR = 1.38 Fig. 3.15 Neutron spectrum in molten fluoride fast breeder (roughly) arpitrary units 14 10°* 107" ————g» Median enargy. 60 kaV Lethargy The results obtained, in spite of the uncertainties are en- couraging. The simplified breeding ratio calculation gives a BRyy+ of 1.517 and the computed value is BRy,t = 1.465 (table 3.13). Table 3.14a shows that the influence of fission products 1is very significant (see also Fig. 3.16). However the change of structural material of the wall form beryllium to 1ron has ocnly a small effect on the breeding ratlio and specific power (table 3.14b and Fig. 3.17). In the calculations for three zones studies here, 1ncreasing the radius of the internal fertile zone from b0 cm to 80 cm, that is a volume increase of 2.3/ has little effect on the total breeding ratio in spite of changes in the regional hreeding ratios (table 3.15a and Fig. 3.18]). Altering the small amounts of plutonium 1n the fertile material as 3 result of reprocessing efficlences also has only a small effect on the breeding ratic and specific power (tablie 3.75hb and Fig. 3.19). Becouse of the good experience of American and French groups using graphite as a structural material for molten fluordie thermal breeders, calculations have been made using grapnite for gseparating walls for fast breeders. Graphite 2 cm thick as the wall material was chosen. The results are rather encouraging. The breeding ratio using graphite is still very hiph, even slightly higher and the specific power in the fuel is lower (table 3.10a and Fig. 5.20]. Changing from the complicated design of three zones with fuel in the intermediate shell, to the "classical” two zone design having fuel in the central region results in a dramatic 1in- crease in specific power to a prohibitive 26 kW.cm™? (table 3.1060, Fig. 3.27, tabie 3,177, Table 3.13 GSimplified calculation of the breeding ratio Median energy {from computer calculation) Pu-239 J-238 (arbitrary) Fast fissicon of ertile component (from computer output) Fast bonus Total positive Losses by absorption (from computer) Leakage (assumption) Total losses Total negative fOreeding ratio = total positive - total negative Breeding ratio (from computer output) kel o {barn) a C v o n n - 1 \) L} S (V-1) x § 1+ Y positive Lab leack L tot Ltotfig 1+ BR (macro) — ———— — —— —— Tahle 3.14a Three zones reactor: fluoride fuelled fast breeder reactor Influence of Fission Products. BR = Breeding ratio, total Wall .. o Fission Specific Core fuel/external Numb blanket products (Cs-133) B.R. power e ,? | atoms/cm® x 1n?" oW/m? 2 ¢ 163 Be 0.0001 1.437 4,33 164 He 0.0002 1.427 4,24 166 Be 0.0040 1.14 2.32 167 e 0.004n .17 2.0 Table 3.14bh Three zonex reactor: flucoride fuelled fast breeder reactor Influence of the beryllium-moderatnr F.P. = 0.0002 atom x 10%%/cm? Core Material ] AR, Sp@cific3power Number 7 cm 1.5 cm GW/m 164 e 1.427 4.74 162 Be 1.464 4.19 165 e 1,38 3.45 161 Ce 1.37 3.47 (see also Fig. 3.17) Fig. 3.16 Impact of fission products 1.5 _ with Be 1.4 - with B <« BR 1.3 - 1.2 - = wlth Fe 1.1 | 193 144 165 ; 166 1.0 [ | T | 104 1073 1072 Fission products Fig. 3.17 Impact of the beryllium and iron moderator on BR 162 1'45_1 [ Be BR tot 164 1.40 — 162 164 Be 1.35 — 5 4 spec. power kw/cm? 3 2 1 0 and spec. 5 spec. power kW/cm3 4 power Three zone reactor: fluoride uranium-plutonium fuel Table 3.1ba Influence of the radius of the internal fertile zone & — Radius of Breeding ratio of Core fertile fertile fuel fertile total zone, &m zone zone 165 50 J0.42 g, 056 0.884 1.364 1649 50 0.34 0.053 0,985 1.379 Ratio Ratio of of cores volume 165 237 1.21 1.05 0.B98 (0.4984 169 (see Fig. 3.18) Toble 2,150 Thpee —one Teactor: fluoride uranium-plutonium fuel ITnfluence of Pu-739 in the fertile material Pu-239 1r Spec. Number , fi_ H F.FP. B.R. Pec f fertile concentration tot pOWET [ ) . o cor material ou/m’ 166 0. 007 2 x 10" 1.49 3.79 1657 .01 72 x 10 1.37 3.477 prrme m— e b e e e e ] — e e e . e e e e . e e i [ mn e emem — — Ratio ot corec 15&- 8.1 10 1. 049 1.09 e . , . ¢ . 0¢ (see Fig. 3.19) Fig. 3.18 Three zones fluoride uranium-plutonium fuel Impact of radius of internal blanket 1.20— relat1v1.1o_ BR BR fuel 1.00 0.9 — R =160 cm R = 80 cm T T 1 0 1 2 2.37 3 internal fertile zone volume, relativ Fig. 3.19 Three zones:fluoride uranium-plutonium fuel Impact of Pu-239 in fertile material 1.50 BR sSpec BR power 1.40— — 3.70 Spec power 1.30— — 3.40 1072 1072 Relativ Pu concentration; U-238 = 1.0 Table 3.188 Tirrpe zonegreactor: fluoride uranium-plutonium fuel Influence of graphite as structural materisl st wall between Znd wall between Number internal fertile |[fuel and external B.R. Spec. of core zone and fuel fertlle zone Total power; 2 cm 2 cm GW/m? 164 Fe, Mo Be met 1.427 4.24 171 C graphite C graphite 1.45 3.65 -.———————-—-—r-—--———-—w--—-»—-——-—- —_——— e ——— ——e — — — —— — o —— —— — = Ratio of cores / hite hite ' i graphite graphite 1.00 0860 164 Fe Be (see Fig. 3.20) Table 3.16b Two zones reactor: fluorlide uranium-plutonium funol Influence of the geometry of the reactor Structure o Number B.R obEes of core Internal | Intermediate External total powEs zone zone zane GW/m 160 Fuel Blanket Cooling 1.424 26.7 two zone Zone zane reactor 161 Blanket Fuel Blanket 1.37 3.47 three zZone cooled reactor out of core Ratio of cores 160 i 1.04 /.69 101 (see Fig. 3.21) Fig. 3.20 Three zones-fluoride uranium-plutonium fuel Graphite 1nstead of beryllium as a structural material 1.50 — 5 BR tot spec power k'/cm3 1.40 A - 4 1.30 — 3 Fe-Mo, (1st wall) graphite (1st wall) beryllium (2nd wall) graphite (2nd wall) Fig. 3.21 Two zones: fluoride uranium-plutonium fuel Impact of geometry — 30 1.50 BR tot 5 spec power kw/cm3 1.40 1.30 — 10 0 Two zones Three zones t o . ] Table 3.17 Three zones reactor: fluoride uranium-plutonium fuel Influence of the plutonium concentration Pu total Number ;; ;1;1 Atoms B.R. B.0G. Spec. Power of Core Mol o cmox 107 tot. gain GW/m’ 165 14 0.0027 1.38 .38 3.45 1772 7 0.0011 1.05 05 4 .05 ————— —_ — ————— = — e — — — —— — A — e e —— —— ———- — o] Hatio of cores /< 5 . ToE 0 0.5 0.76 0.13 1.1/ i, “..2 Burner reactor with molten fluoride fuel Fuel: Pqu-B.Bu NaF-2,41 Z”Fu 3 Zpecific power ~20 kWem fotal power 11 Gw (thermal) ] 100 98,8 109 112,6 Radius, cm 3.6 High Flux Reactor with Fluoride Fuel Can a fluoride fuelled burner, as opposed to a chloride fuelled reactor be considered as a reactor for transmuting fission pro- ducts? (See chapter 4, for the high flux transmutation reactor). In such a reactor a fuel made up of PuF3/5.4 NaF/2.4 ZrF, has been assumed. The calculations have been carried out for a larger bruner of 11 GW(th) and the fission products are assumed to be generated in a total system of 55 GW(th). The results are not encouraging in spite of the fact that the neutron lfux in the target region was 1.05 times higher for the fluoride fuelled reacotr as opposed to the chloride fuelled reactor (Table 3.18, Fig. 3.22). The effective half life of both fission product nuclides was (in years) In fluoride In chloride fuelled fuelled reactor reactor (reference) Cs-137 8.57 5.93 Sr-90 1.73 1.83 These "benefits” must be balanced against a specific power which is twice as high as the reference case. In the fluoride cors this is 19.9 GW.m™* and for the chloride reactor 10.1 GW.m"-?. Such s high specific power is not realisable. In addition since graphite might be used in place of beryllium oxide as moderator (possible for a neutronic viewpoint) a signifi- cant improvement in corrosion problems is obtained. This has been proved by the excellent experience of Oak Ridge National Labora- tory with one proviso - at ORNL the fuel was LiF - BeF, - ThFg - UFg. - 8/1 - Table 3.18 High-flux burner reactor with fluoride fuel Total power 11 CW(th) Hurning fission products form a total system of 55 GW(th) s Neutron flux Specific power ol Dni ] Components 1018 cm 2571 (GW m=?) Callué [C?] (atom 107 *cm™?) total Transmutation oLume A thermal rate (5™ 1) I T~ 88.8 cm Cs-137 0.0116 Target zone Sr-90 0.0016 4.01 C5-137 : 1.8x107° Vol: 4.1 m? Oxygen 0.0145 2.21 Sr-90 : 1.2x10°°8 Deuter 0.0145 IT A8 - 109 cm Be .060 5,08 “oderator, wall Oxygen 0,060 1.83 with thin 5.25 graphite layer 0,235 111 7.4 - 112.5 cm Pu-2338 0.0017 susl ozone Pu-240 0.00042 Vol: 0.55 m’ PuU-241 0.00021 5.02 s —_— 9.9 GW Na 0.0075 0.0457 1 " r 0.0051 z 0.0340 IRV, "Mz.6 - 118.6 cm He 0.060 4.87 wall Oxygen 0,060 0.034 3.9 0,041 \ 3.6 - 218.0 cm e .06 0.0023 Feflector zone 2.8x10°° 4. A HIGH FLUX BURNER REACTOR FOR TRANSMUTATION 4.1 Need for fission product transmutation 4,11 Introduction The problems associated with the management of highly radio- active fission product waste has been intensively and exten- sively discussed (Fig. 4.1). Here only the transmutation of fission products (F.P.) is dealt with. The recycling of the actinides is not treated. Transmutation occurs by using neutron irradiation in a fis- sion reactor. A snort outline of this chapter can be presented in ths form of the following questions - why, contrary to many assertions, 1s neutran transmutation in a fusion reactor not feasable? - this in spite of the fact that the fusion machine has often been proposed far this purpose - why are recent opinions concerning transmutation in fis- s5ion reactors rather pessimictic? - could transmutation 1n a fission reactor be possible taking into account the neutron balance in a breeding system? - which fission products are candidates for irradiation in a fission reactor? - 1s the rate of tramnsmutation sufficiently hieh in a fissicn reactor? - in what t [BAS of reactor 1is the transmutation hysicall Y Y D(]SSith? - what are the limiling parameters for transmutatbion in a solid fuelled filssion reactor? - 15 a very high flux fission reactor possible if the fuel is in the liguid state instead of the solid state? Fie, 4.1 POSSIRBRILITIES FOR (see also table 4 1 1) '‘RANSMUTATION OF Long-tived fission products Other waste management methods Transmutation Changing neither <4f—””’/”’/”’7 30sr AND Anor Z Changing A but not Z Poton bombardment Changing 2 Neutron bombardment neutrons from accelerator N:'ustr_on Neutron capture emission () {(n,2n) 7 Periodic irradiation Contmupus {explosion) irradiation Secondary Primary neutrons from reactor Neutrons from fusion Neutrons from fission — — Proton bombardment, <‘\\\\\\\\~\“‘-; Thermal reactor Fast reactor with thermal central Z011€ Fast reactor 137CS Table 4.1 Transmutaticn possibilities for differsn. devices Machine Clux/ ~eactions, and remarks of authors Cnergy of original reports. Accelerator of Protans Reaction p,xn!] not promising. Ruled out medium and high 100 MeV on basis of energy balace criteria. energy protons Protons Spalation (p,xn) and (n,2n) (n,¥y) 1.70 GeY with secondary neutron flux Cs-137 as Not feasible within limits of current target and/or technology. The capital cost is thermalised prohibitive. flux of (see table 4.2) neutrons Fusion {thermonuclear) reactor in all cases with wall Fast flux of 14 MeV neutrons from (0D-T) & = 5x101 %0 em2g7 ! Neutran reactions (n,2n) and (n,y). Fast Flux of 5 x 10'°n cm™ 257! Thermalised flux 1in ceryllium trap Practically only (n,Yy) Thermal flux 6.7 x 10'°n cm 257! Attractive transmutation rate has naot been demonstrated but possible to transmutate all Cs-13/ and Sr-90 creatad by fission reactors Nuclear Fissile explosive or thermonuclear explosive Technically not feasible. No. of explosions per year very high. Appr. 3800 per year each of 100 kton. (For USA in year 2000 Cs-137 and Sr-40) Probably not ecceptable to public! /Y Fig. 4.2 FISSION PRODBUCTS @®veta stable Wate QObeta unstable \beta decay 10 b 1 daly 1\3 d 1ui) d l‘ year ]lJ y lC‘\ ¥ li‘;\l ¥ 1 T T T T T 1, i 10" 10° 10" 107 108 1? 10t il time, sec Fig. 4.2 FISSION PRODUCTS CANCIDATS FOR Half-1ife, years TRANSFUTAT 10N >>10'* | 107 stable | short \ nuclided living F.P. F.P. \ / \ b ey / / — — — —— Thansmutation, shortning the half-11fa very long living F.P. quasi stable F.P. Table 4.2 Passibility for transmutation of F.P. particulariy Cs-137 and S5r-80 in 2 fission reactor according to BNWL - 1500 Reactor Referance Flux Remarks Steinbtersg The authors use a wrong value: Thermal Gwer reactor Wotzak Kr-85 with large 0 = 15 barns = cact P Manowitz, 1964 instead of ¢ = 1.7 barns. Isotopic separation of Kr-isotopes. 3 x 10'? thermal Only I-129 can be transmuted. 10'® in the trap An equal or greater no. of F.P. L smaller 1n the would be formed in the fission Steinberg, 1364 , ) presence of the process per transmutation event. F.P. target. This reactor does not meet the high flux (trap) Claiborne, 1972 2 x 10'° thermal criteria of overall waste balance and of total transmutation rate. Meutron excess 0.15 - 0.3 at the expense of being no longer a . . viable breeder of fissile material. liquid metal i _ 15 . Fast Claibtorne, 1972 1 x 10 fast Alzo this flux does not allow the fast hreeder attainment of a sufficiently high transmutation rate and is, therefore, not a feasible concept. Fast with thermalil liquid fuel fast reactor with thermal column. this paper - how could such a high flux reactor with circulating liquid fuel and a thermal column cperate as a "burner” for some F.P. (Cs-137, 5r-90 etc.) transmutations? (Fig. 4.3) - 1s such a system feasable? Comment In BNWL-1800 it was noted that the calculation (in a moderating blanket of the CRT) represents a more realistic blanket config- uration with a neutron wall loading of 10 MW/m? (This is still a very optimistic value. M.T.]). In this case the following data have been obtained for a therma- lised neutran flux from a CTR with a 10 MW/m? wall loading.(Tab.l.3) Table 4,3 For 80% ¢ thermal b0 beg effective fraction (n.om 2s-1) (n,v) (n,2n) tVZ effective 291 ke CoT g 51 w1010 | o - 0.117 | o = 0.10° | A = 22.2x107 157! (n,y) (n+Zn) barn barn 7.91 x 10710 7.0 x 10°1° 3.9 years The conclusions of this study are that useful quantities of Cs-13/ could be transmuted under the projected CTR blanket loading conditions. The reduction in Cs5-137 "toxicity” 1s still expected to be at most a factor 3 down. In addition a study of the build-up of fission product nucledl in order to establish the roquirements of periodix chemical processing and associated costs has not been carried out. Fig. 4.3 Accelarator for transmutation BALANCE FOR 1 FISSIONED ATOM < '0@.«. XA SRS HEAT WASTE 0.008 PROTONSat GeV x 40 = 0,32 NEUTRONS for TRANSMUAT 10N MUTATION Fig. 4.4 Neutron balance Camment H.W. Lefevre (appendix to BNWL-1300) makes an interesting comment on the study of the transmutation of Cs-137 and Sr-30 in CTR: "Everyone knows that a CTR will be "clean”. Don't spoil that illusicn. I think that I would worry some about a CIR loaded with 50 kg of Cs-137". 4.1.2 Why some opinions concerning transmutations in a fis- sion reactor are rather pessimistic A recent and mecst intensive study of the use of a fission re- actor for the transmutation of fission products has been pub- lished by Claiborne (1972). He writes: "The problem fission products cannot be eliminated by any system of fission power reactors ogperating in either a stag- nant or expanding nuclear power economy since the production rate exceeds the elimination rate by burnout and decay. Unly abt eguilibrium will the production and removal rates be equal a condition that 1s never attained in power reactors. Equilib rium can be obtained, however, for a system that includes the stockplle of fission products as part of the system inventory since the stockpile will grow until its decay rate equals the net production rate of the system. For the projected nuclear power economy, however, this will reguire a very large stock- pile with its associated potential for release of large quanti- ties of hazardous radio-isotopes to the environment. It is this stockpile that must be greatly reduced or eliminated from the biosphere. A method suggested by Steinberg et al. i1s ftransmuta- tion in "burner reactors”, which are designed to maximize neu- tron absorption in separated fission products charged to a reactor. If sufficient numbers of these burners are used, the fission products inventory of nuclear power system can then reach equilibrium and be maintained at an irreducible minimum, which 1s the quantity contained in the reactors, the chemical processing plants, the transportation system, and in some in- dustrial plants. - - 43 - If the assumption 1s made that burner reactors are a desirable adjunct to a nuclear economy, what are the design requirements and limitations: it is obvious that they must maximize (with due regard to econcmics) the ratio of burnout of a particular fission products to its producticn rate in fission reactors, and the neutron flux must be high enough to cause a significant decrease in its effective half-1ife. 0Of the fission types, the breeder reactor has the most efficient neutron economy and in principle would make the most efficient burner if all or part ot the fertile material can be replaced by a Sr-Cs mixture without causing chemical processing problems or too large a perturbation in the flux spectrum because of the different characteristics of these fission products. The cost accounting in such a system would set the value of neutrons ahsorhed in the fission product feed at an accounting cost equal to the value of the fuel bred from those neutrons. The maximum possible burnout of fission products would occcur when the excess neutrons per fission that would be absorbed in a fertile material are absorbed instead in the fission pro- duct feed. The largest possible burnout ratic would then be the hreeding ratio (or conversion ratio for nonbreeders) divided by the fission product yield. The estimated breeding ratio for the Molten Sall Breeder Reactor [(MSBR), a thermal breeder, is 1.05 and for the Liquid Metal Fuelled Fast Breeder Reactor (LMFBR), 1.38. The yield of '*7Cs + %%Sr is 0.12 atom/fission, but a number of other isotopes of these elements are produced which would also absorb neutrons. However, if the fission pro- duct waste 1s aged two years before separation of the cesium and strontium, the mixture will essentially be composed of about 80% '*7Cs + ?%3Spr and 20% '*°Cs (which will capture neu- trons to form '?°®Cs that decays with a 13-day halflife conse- guently the maximum burnout ratio for '?7Cs + 2%y Wwill be de- creased by 20%. This leads to a maximum possible burnout ratio ot about /7 for the MSBR and about 9 for the LMFBR. Unfortunately, however, the neutron fluxes in these designs are well below 10'°%n em™?s” !, Any modificetions of these desligns to create high neutron fluxes will increase the nectron leakage and de- crease the burnout ratios significantly”. (Claiborne 1072 Comment it is not clear why Claiborne claimed that after 2 years ageing and separation of strontium and caesium the isotope composition will be From Crouch (1873) the fission products of U-235 have the fol- lowing composition (2 years ageing) (in at % per fissioned nu- cleocus). {see Table 4.4) Sr-88 (stahle) 3.63 Sr-90 (28 years) 4.39 Cs-1233 (stable) B.57 Cs-134 (2 years) 3.5 (7.08 0.5 from independent yield) Cs-135 6.26 Cs-137 5.99 Subtotal 30.34 The realistic dala are unfortunately more than twice those citod by Claiborne. The same negative opinions concerning the use of Flssion Reactors for F.P. - transmutation are given by the following authors: - A.S. Kubo (BNWL - 1300): "Fission products are not conductive to nuclear transforma- tion as a general solution to long term waste management”. - BNWL - 1800, itself: "In summary it is improbable that transmutation of fission products in flission reactors could meet any of the technical feasibility requirements for the production of stable daugh- ters”. - Claiborne (1972): "Developing sprcial burner reactors with the required neutron flux of the order of 10'7n cm ™ ?s™! is beyond the limits of current tecnhnology”. 4.1.3 Which fission products are suitable candidates for transmutation and in what quantities? A simplified breakdown of neutrons and fission products produced by fission of 1 fissile plutonium atom is given in Fig. 4.4 From this it must be clear that only a very limited amount of fission products can be irradiated by neutrons of the whole system to retain a good breeding gain and doubling time - in other words a s elf sustaining and expanding breeding system. For further consideration it is postulated that the maximum num- ber of transmutable nuclides equals T = 0.3. the proposed system for the transmutation includes two types of reactor: - power reactors in the form of fast breeder reactors with s total power of three to four times that of: - a high flux burner reactor. The crucial F.P. nuclidas are characterised in table 4.4 to- pether with others. The data available now makes it possible to estimate the number of candidates for transmutation in our breeder/burner system, using the tollowing criteria ~ thr total amount of all nuclides to be ftransmuted carnot be greater than the estimated value of T = 0.30, that is 30 atoms of F.FP. nuclides for each 100 fissioned nuclides. - the order of priority taken from this table is given as Cs > 5r> 1T > Te - in Lhe first instance no isotopic separation process is postulated. Table 4.4 shows the F.P. nuclides selected for transmutation. Table 4.4 The priorities for the tr-asmutation of fissioned products Soloctod Yield for fission of Atom/170 atom Pu-235 Assuming isotopic separation SSLEELE 109 atoms of Pu-23°9 Subtotal stoms/100 atoms Pu-230 Cs-133 (stable! .97 6. 31 0.14 Cs-135 .54 140 14,450 7.54 } 14.37 Cs-13/ .69 21.140 £.09 Sr-390 2.716 23.32 2.18 205 Z2.208 Sr-88 (stablel) {1.44 x O 23.349 0.029 J (2% isotopic separation efficiency) I-129 N7 24.5193 1.17 .55 1.18 1-127 (stable) .38 24.899 0.01 J Tc-389 5.81 .81 30.709 5.81 5.81 Kr-83 (stable) .36 - > .56 Kr-84 (stable) 5 474 Kr-85 672 0.67 } 9.7 Kr-86 (stable) 882 33.183 0.04 ) Total 33.183 24.28 9b 4.17.4 In what way could a burner reactor be coupled to a system of breeders? The aim of the calculations used here is to show - Lhat given a system containing some breeding power reactors with a breeding gain of G - the fission products from all of these reactors can be tans- muted in the high flux burner reactor, which includes of course the fission product transmutation for the burrer re- actor itself. The calculation of the ratio of breeder power to burner power 1s as shown here Transmutation rate (atoms s7!), T Yield of fission products (atoms/fissiaon), VY Effectiveness of transmutation device £ T = e 1 (Y(?%Sr) + Y('37Ce) + Y{other F.P.)) Y(3%3p) = 0.0471; Y('?*7Cs) = 0.064 Preeding gain for the total system without transmutation GY = 0.375 (arbitrary) Freeding gain for the total system with transmutation G Ratlo of fission to total capture: o = 0.24 (arbitrary) C = G2 - T/(1+a) Taking a numerical example with the same values arbitrarily chosen {including also another fission product with ayield of 0.17) T = 0.57Y (0.041 + 0.064 + 0.1) = 0,40 5T = 0.375 - =10 = 0,07 e 1+0.24 T which 1s sufficient for a power system having a slowly increas- ing capacity with a doubling time of 100 yr. This corresponds to Lhe near steady state case. To determine X, the number of power reactors X-Go-1 = (x+1) & T G +1 _ 0.05+1 ) . “EB_GT" 3.375-0.05 ~ °43 The correspondence ratio of the power of the breeder reactors and the burner reactor in this case equals This means for example for 8 power reactors each of 3 GW(th) (i.e. a system total of 30 GW(th) can deal with the transmuta- tion of the fission products chosen here. The electrical output assuming an aefficlency of 40% is (3x8 + 1x7)x0.40 = 12 GW(e) (Fig. 4.5). 4.17.5 Is the rate of transmutation sufficient? It 1s clear that the rate of radioactive nuclide removal 1in a field of particles 1s given by - inZ AQFF ) Adpca ' >\trarmmut(:fi:ion ot ey - Y2 (efF) where A, = constant of radicactive decay (s~ ') decay trans - trans o = cross section (em?) for a given reaction ® = flux of the reacting particles (particles cm “s” ') this value of Agoep will be used later for the calculations of the neutron flux required to permit the transmutation rate to match. Fig. 4.5 Scheme of the proposed breeding power system with "self-cleaning". (For the sake of simplicity, only the routes of the fission products *%°3Sr and '?7Cs are shown). 8 Fower Reactors, each GW (thermal) total ~24 GW (thermal) ) @@ Fission Products . Partition of Fission Products BURNER (inctuding isotope REACTOR ' separation} Vi GW (therm) T Other Fission Products Fast Core ¢ goSr Fuet 1370 Reprocessing of lrradiated Nuclides - — l Stable uclides g g‘Zr, 138, - 100 - Let us assume that the energy production is based on a set of n burners and nX breeders. At a time t, (see Fig. 4.5) when it is decided to stop the use of fission energy production in favour of other sources the total amount of a selected fission product present 1is (1) N(E ) = (X+1) ne—t eff with K = Y«P/E Y = yield of the selected F.P. P = power per burner (or breeder) (Watt) E = energy per fission (Joule) This amount of F.P. is located only in the burners, therefore sach bhurner can receive Kk ke?? (X+1) although then any production should only represent K AEFF in the steady state. - 1071 - At time t, the nX breeders are shut down and only n burners are in cperation. Later on {(time t,_4) the nuclide removal is such that a re-arrangement is possible and one burner can be stopped, ist F.P. contents will be loaded in the remaining burners etc. At the beginning of each time step tp the p burners which are still working contain the maximum possible amount of F.P. (X+1) = eff ) N(tn) ] N(tn_qJ _ N({t ) ) N(tp 1] _ N(tzl _ N[th ) “ n n-1 D p-1 Z 1 ) S Xe1) e 5 eff where N(t) represents the total amount of the selected F.P. One could imagine other schemes: for example the rearrangement could be made only when 2 burners are to be shutdown. From the reactivity point of view this solution is worse than the one preposed. Coming back tec the original solution, one has still to sclve at each time step (tp, tp-1) the burn-up eqguation. dh -, \ N where the right hand side i1s the dt eff F.P. production. L then the solution is Using (2) the time needed to go from p burners to (p-1) can be deduced 102 - with a summation one obtains the time t1 after which one burner only is in operation. p=n . - 1 (63 A ps (Ey-E ) Z :1”1_X+1 p=2 peX A more direct evaluation can be obtained iF¥ n is so large that the number of operating burners changes continously with time (p = n(t)) then by a single elimination of p between (2) and (3) one gets 4N . P v | o (4 M) = N(E Jer 2t sty ) n. eff X+1 ’ n NCL ) N X1 o xel (5 A _pplt t ) = = In s = 1n (n ) The two approaches give similar results except at the end when few burners are in operation., For tLimes longer than ty only one burner 1s operated and the amount of F.P. would decrease from We shall postualte that 1t has no sense to operate this last burner when the amount of F.P. is only 1.2 times longer than the asymptotic value which requires a new time interval (eq. 4 p = 1). - 103 - The total time t, - t,, will be the sum (6) + (7) which corresponds to the reduction factor n{xX+1) 1.7 Further reductions can only be obtained by natural decay (t>tOJ. Numerical application: with X = 4,n = 100 which means the esconomy was based before t, on 400 breeders, the initial F.P. amount is reduced 415 times when the last burner is shutdown. Then the re- quired time is defined by ggerltg-tp) = 8.93 (8.76 with the approx expression). If this time is to be less than say 60 years 2 re- actor generations) then Agpp = 4.7x107°s~' (ty, eff = 4.7 years). Now the problem of the intensity of the neutron flux desired for transmutation arises. Since the most hazardous F.P. nuclides are those which apart from their high metabolic activity and high re- tention in living organismus alsoc have a half life of the same order as a human life span of B0-70 years we arrive at the fol- lowing list of hazardous isotopes which are the mest important for transmutation. Kr-85 10 years, 20.9 x 1071071 Lo A dec _ - - -10_-1 Sr-90 tVZ 28.2 years, Adec /.78 x 10 s - = = -10 ~-1 Cs-137 tV2 30 years, Adec 7.32 x 10 5 desired 'half life’' = 4.7 years, A : = 4.7x10"%s 7! desired 35 we know xdesired i Adecay ' xtransmutation The most important problem arises from the fact that the two nuclides Sr-90 and Cs-137 have very small cross sections for neutron absorption in both the thermal and fast regions. 104 Cross section, olbarn = 1072%cm?) Ratio thermal fast therm./fast Sr-90 0.8 barns 0.0076 barns " 400 Cs-137 0.11 barns 0.0137 barns g 8 therefore to achieve Adesired fluxes should be: fast flux for transmutation of Cs-137: ¢ fas ®Fast desired—Adecay 4,7 x 107%s~! the neccessary 4x10°° t o] 4.0 x (Cs- 137 fast) v0.01x10” 2 107 (n em™ 257 1) thermal flux for transmutation of Cs-137: ¢ th Ax107° ’g 3.2 x 10'% (n cm %57 1) 0.11x10 24 thermal flux for transmutation of Sr-90: o “th ‘see also Fig. The - a ing des achieved) gquestion t fast flux 4.,0x10" 9 5.0 x 10'° (n cm 25~ 1) 0.8x10° 4.7). hen arises, of 4x10'7 of i in what device are such fluxes possible a thermal flux 6x10'%, It is interest- to point aout that during the period of 60 years which provi- the reduction factor of 415 the natural decay of Cs-137 would have reduced it only 1 (if the Agf¥f 4.7%x10°%s~! can be by a factor 4 which demonstrates the efficiency of the burner. Also the burning which occures during the first period reduces the am A Fop._8ff deca ount of times b. y (t tn) 7 times for Cs-137 Pig. 4.7 1016 4 Effective half life and neutron flux for transmutation wmcevemes THIS PAPER ——cwmee PARISH (1977) o= CLAIBORNE (197..> MSFR Mclten Salt Past Reac-or Neutron J flux (n/em's) \ % Intensive Neutron \ ontroll generator) Ther-nl AMATTITRFTETTTETTTEEECTEERRKRRTT (Light Water Reactor) Effective half life (years) - 106 - 4.17.6 In what reactors are the transmutations possible? From the point of view of this paper the most important process is the transmutation of some of these nuclides by neutrons in a fission reactor. The criteria given in chapter 2 limit the choice of system. That 1is - the number of F.P. nuclei cannot be too iarge in relation to the number of fissioned atocms in the burner reactor (reactor for transmutation) because the latter process also produces new fission products. - the fission reactor should be self-sustaining - that is a breeding system. - Lthe specific power of the reactor is preoportional to the neu- tron flux. Hiegh neutron flux means high specific power which is controlled by the effectiveness of the core cooling. - the specific power F and the neutraon flux ¢ are coupled by the fission cross section and the concentration of fissile nuclide (Ng) For thermal neutrons ogf 1s approx. /00 barns and for fast neutrons only 1.8 barns, that 1s 400 times smaller. For the given total power and the same specific power the product Ne+d for the thermal reactor must be approx. 400 times smaller than for a fast reactor. Since the critical concentration of fis- sile nuclides 1n a thermal reactcor can only be 10 times smaller than fer a fast reactor then for a given specific power the neu- tron flux 1n a fast reactor can be about 40 times higher than that of a thermal reactor. The cross section for thermal neutrons for the nuclides considered here 1s from 3 to 10 times larger than in a fast flux and this must be taken into account. All these factors bring us to the following solution of the prob- lems under discussion. - the highest specific power and hence the highest neutron flux 1s pogseible 1if the cooling process is carried out by the fuel itself and not by a separate cooling agent only. This directs our interest towards a reactor with molten fuel in spite of the exotic nature of this solution. - the high flux reactor must be a fast reactor (small for fast fission) - 107 - - because Otp > Ofggt the thermalisation of the high flux in a internal thermal zone is postulated, then it is possible that zone core therm. fast - the first approximation is made for an i1sotopically pure radio- nuclide e.g. Cs-137 without Cs-133 (stable) and Cs-135 and also Sr-90 without Sr-88 (stable). The discussion then results in: - transmutation of Cs~-137 (and some other nuclides) in a ther- malised central trap of high flux neutrons: ®therm = 5x10'® n cm™?%s7! - production of a high flux of fast neutrons 5x10'® n ecm™?s” ! and the high specific power of 15 kW em™® is achieved by means of liquid fuel circulating through an external cooler. - transmutation of other selectred fission products in an external thermalised zone with a thermal flux of 5x10'° or Tx10' n oecmm%st, - coupling of one burner - high flux fast burner reactor with a system of 'normal' power breeder reactors. A.1.7 What are the limitations of a solid fuelled reactor? Can the desired specific power of 15 kWem™® be achieved in a solid fuel reactor? These are the self-evident limits in this case. - the rate of burning of fissile nuclides is limited due to depletion of fissile or an increase of F.P. nuclides - the heat transfer limitation of fuel/clad to coolant - the temperature and temperature gradients 1in the fuel and cladding (melting, mechanical properties) - 103 - - the boiling of the coolant - the limitation of coolant velocity, pumping power, stability. Now we discuss these limitations in more detail - the dwell time in a solid fuelled reactor in core for the fissile nuclides must not be too short. concentration of fissile nuclides*maximum burn-up dwell fission rate We could write: where N = concentration of fissile nuclide N = — and f = 3.1 x 101°% fissions/joule = maximal burn-up = power (watts) b = R = fission rate [(fission s~ 1) R from this: : . Pofib b dwell o*QP.T UF'Q b = 0.03 (corresponds to 30.000 MwWd/t) LRED 5pg w 10-2% om? fiss d = 5 x 10'® n cm 25 ! tdwell = 850 s = 14.3 minutes hut also for b i .10 we achieve t = 4/7.6 minutes dwell - 103 - For a fast reeactor {(some arbitrary values) b = (1,10 $?St = 1.8 x 10°2?% cm? fiss o] = 5 x 10'® n gm™ % ! tdwell Conclusion: - the dwell time in a thermal reactor is prohibitively short, ~ in a fast reactor it is more reasonable but still very short, especially in the case of a solid fuel reactor - the limitation of specific power by heat transfer is as follows: Specific power, Pspec' in a "good” 3 GWip power reactor and wilth the appraopriate flux taken from literature is: 0.05 kW/cm?; o = 5 % 10' n cm%g ! spec th fast = = kW/ecm?; @ spec fast thermal P = 5 x 10'° n em™ 257! in a high flux reactor: (see also Fig. 6 1.8) thermal:P i ] o= kW/cm? 3 x 10 nocm %g7! a=y I spec ) th P = 1.5 kW/cm?®; @ = 3 x 10'® n ocm %g”! spec th fast: = = 1.0 kW/cm?®; & = 1.,5x10'® n cm™?s7! spec fast With the same geometry the very high flux reactor desired here would have the following flux for Cs-137 transmutation: I ~J ™3 Sl P = .y T 3 [ 5.0 x 10'%, the specific power P -b G 3 HH i th l nNJ - x = - 0 3 w = 4, 107, P f for ¢ < 0 x 10 the specific power PFast For a solid fuel we postulate the following "unit-cell” Dimension Volume Cross- Surface- section area area Cell: 0.9%x0.9x1.0 cm 0.81 om’ 0.81 cm? 3.60 om? Fuel: = 0,60 cm 0.783 cm? 0.283 cm? 1.6885 cm? Cladding: diam = 0.63 cm wall: s = 00,03 cm diam = 0.60 cm N.568x10 " %cm?® 0.568x107%cm? 1.904 cm? Coolant: G.521 cm? 0.521 cm? "desired” per unit fuel element surface area: In this specified cell of a would achieve a heat-flux, (for both bypes of reactors, high-flux-reactor, thermal and fast) L2 21 kw/em’-0.81 om’ -2 fs 1. 885 cm? 4 kioem we Using now a simplified model for the first guess of the tempera- ture gradient we can say: the amount of heat generated in the fuel must be the same as that leaving the surface of the cladding material. A 5 Tclad H?S By wWwhere s = wall thickness and A heat conductivity {(Weem™1ek™ 1) an optimistic value for stainless steel is A = 0.4 Weom 'ek ™! - 0.03 _ o ATclad = 9000 T T 6/57C It is evident that this result is not realistic. The solution of this problem may be the thermalisation of neutrons in a high flux fast core and the irradiation of Cs-137 in a thermal central zone. In such a thermal central zone we postulate (and this must be based later on corocalculations) ®th = e CI)Fast ta reach &+y = 6.0 x 10'® we require dfaet = 5.0 x 10'% n om 25710, For this fast flux the specific power can be assumed, if weo take into account the effective increase of the fission cross section because of the influence of the thermal trep. The simplified cal- culation results in a specific power of 10 kW cm™?. The corresponding heat-flux is therefore reduced to g PF.V ell H = PeEC CP 4.3 kW cm”? fs AFS and the temperature gradient to _ 4 L.o-2y _0.03 (em) _ 0 ATGlad = 4300 (Weem™7) 0.4 (Weem™ K- g M~J L This value is still rather high. A rough estimate of the thermal stress in the wall can be taken as coefficient of linear & . expansion (K™1) 5 = % . Bth C AT - F, - - TESU%Uim?E)QIBStiGity and the corresponding values for stainless steel (13-9 0OL) g = % < 151075 (KT e 323 (K) 2.5+10% (Kpecm™?) = 5500 Kpecom It is also evident that this result is not realistic! The resulting thermal stress in the cladding wall coupled with the high flux is prohibitive. Hlere, The most important problem 1s the local overheating by the thermal flux present at the interface of the thermal column and the solid fuel core. If, in the fast region near to the interface approx. 20% of the enrgy comes from the thermal neutrons, the fol- lowing ratio of fluxes must be achieved: T T %tn east T Peast?! U7 0 fast 1.8 “eh TER fast T U7 Ton Tt eagy i - G0 7 "% . ¢ TR Pfast That the thermal flux must be 2000 times lower than the fast, would seem very difficult to realise. 1.7.8 The liquid-fuelled fast reactor with central thermal zone A much better solution is using a liquid-fuel. The transfer of the heat generated is done, by pumping and cooling the liguid fuel ocut of core. For a unit cell of 7 cm3, which in this case consists of fuel only, we can write the following hesat-balance. ) = . [ ] —— . = *“‘ - 1 . -3 . -1 %SDQC (p+c) W i (pec) heat capacity (Jecm K™) W = velocity of fuel (em/s) T/1 = Lemperature increase per unit cell (Kecm™ 1) = = apecifi ) Wepemm™ 3 Spec pecific power (Weecm™ ) It we allow In the core a temperature increase of AT/l W 2 dececm <0 This velocity appears to be within the practical limits proposed for 2 fuel-velocity of 40 m.s™! for a reactor with 10 kW.cm™? specific power. This point however has to be seriously investi- gated, as erosion due to high velocities is a problem. For a /7000 MWith) core with a specific power of 21 kWecm™? the fuel volume is about 0.330 m?3. The target volume, that is the volume of irradiated (transmuted) fission products e.g. Cs-137, is postulated as 1.3 m®. The dia- meter of a sphericael core is therefore 145 cm. The temperature increase of the unit cell of fuel, in one pass through Lthe core, rguals approximately T = 3 (dacecm™ ') « 146 (cm) = 438°C fuerl e : and for a fusl inlet temperature of 550°C we reach an outlet tem- perature of approx. 988", The for socme years. Lane fas reactors: "As an alternative to the possibilitly test facility. The ability to achileve time eliminate the actor. just mentionsd 1is idea of a liquid fuel high flux reactor has A fast spectrum molten hiph fissile concentration crder to get mean neutron energiles switching from an NaF - UF, salt based) been discussed (18971) for example writes about high flux some consideratilion has been given at Oak Ridge of using a molten salt reactoras a fast flux primary virtue of this approach includes the very high power densities and at the same down time associated with refuelling the re- salt reactor however reguires a (i.e. 300 to 500 ¢ 27°U/1litre) in in the range of 10 to 50 keV. (on which the energy range to a chloride-salt reactor would permit a higher mean energy for the same fuel concentration but would require the development of a new technology assocliated with the use of chlorides. Since the fast flux level 1is largely debtermin- od by the power density a flux of the order of 10'°% or more cor- responds to a peak power density in the fuel salt In a range of 5 oto 10 MW/litre and to a power level of about 1000 MWith)., This means that there will be only 100 to 200 litres in Lhe core: however the external volume would be about 10’000 litres”. 4.7 The Neutron-physical Aspects of the Hipgh Flux Reactor 4,720 Introduction The idea of destroying the beta active long lived radionuclides is based on the following: B A Fi?oo= primary lsalon { > { ‘ T Soanbanooea denay ST re ot ! ! | e T ol o ‘:, l - '[ i J Tfl l\ ™ ! l NG 1s unft L L,y ) Ao b trradiavion A - abomic maso | /= atomic number v - LA+ L) 6™ [A+1}F Fo= stable nuolide _ - . - ST spontaneous decay (Z-1) = 1o very short ‘|/J 2 - Fig. 4.10 gives some of the given transformations which may occur under a high flux. (Remark: Fip. 4.3 and 4.9 omited) In this system the following simple assumptions are made - the amount of fission products come from both the fast power reactor breeders and the thermal burner. - the fuels and materials are continually reprocessed - the irradiated fission products are continously (or periodi- cally) separated in order to eliminate the daughter stable nuclides (e.g. Zr-90 and Zr-91) from the decay and burning of Sr-30 - the amounts of transmutated nuclides in the steady state [(85) irradiation are calculated by the obvious relationships for the i-th nuclide. > A,7.7 Neutronic calculations A reference burner reactor concept is shown in Flg. 4.771. The flux trap 1s surrounded by a BeO beryllium oxide spectrum con- vertor, a critical fuel thickness and an outer wall (see Table 4.5 and Table 4.8). Fig. 4.17 shows the calculated flux distri- hbution. The tontal flux in the fuel is similar to that in the flux trap [(Fig. 4.13) I ™J 8] Moderation requirements To form a thermal neutron flux trap one must naturally use neu- tron moderating materials. As is well known, light materials can scatbter neutrons past the neutren-absorbing intermediate-ennrgy resonance region. jH is the most efficient nuclide in this re- spect but als exhibits appreciable thermal absorption. Deuterium. 10, beryllium %Be and arbon 1§C are usual alternatives. Oxygen WfiD 1s rather heavy though frequently already present in a molecular combination. Other light nuclides have unacceptable nuclear or physical limitations. oL 4010 140 139 - 138 137 136 - 135 Cs-135/137 Atonilc Transmutat ion Number nf sele 92— 91 + 103 102 101 — 100 99 - Te-99 - 57 cted Nucolides-Tission products Sr~90 Zr-93% B~ B~ jfiosad 3 B = 28y 6Uh ST Y ir Nb Mo T 38 39 40 41 42 Loneg lived fissinn product belng the obiect of Lrancmutation @ ciable Nuclide () Heta unstabla, intermediabe nucl 1oe (:)ljjng Lived nuclide R- —> Snontaneous beta decav nalf-1life - t+ neutron capture (n,y) - 117 - Table 4.5 OBJECT: HIGH FLUX BURNER REACTOR WITH THERMAL ZONE REACTOR TYPE CEUOMETRY: INTERNAL Z0ONE : THERMAL, HIGH FLUX WALL : IRON, GRAPHIT INTERMEDIATE ZONE: CORE, FUCL WALL : IRON EXTERNAL ZONE - WALL, REFLECTOR : IRON POWER (C% thermal) .7 POWER DENSITY (GW thearm/m? core): MEUTRON FLUX, MEAN (n/cm?s) v FISSILE NUCLIDE: Pu-235 DILUTENT : NaCl COOLING SYSTEM ¢ out of core BREESING RATIOC 0 - PARAMETER STUDIED: TModerators in thermal zone Wall, thickness, beryllium, graphite Volume, specific power cooling parameters METHOD OF NEUTRONIC: ANISN S4 CALCULATION : 23 Groups Pq APP. GCC3 CODFE END - 118 - CEOMETRY OF THE HICGH FLUX REACTOR Neutron flux, ncm ¥ s’ Reflector Fe A Centra! Region /.-/ ya /"/".'Sr(IODl)z,ICSJOD / 0 radius, cm 78,5 875 94.6 97.6 200 Diam = 120 cm ______________________ o — —— —— Thermal Fast 1.6 MeV Table 4.2 Gacmotrey and CTet 41 Dowern = 7 GWlth): fission leutronics of the iligh-Flux Burner Resactor oraoducts from the total system of 30 GWlth)). Region Radius Componcnts Meutron Flux (cm] (atams 1077 "em™?) (n cm %5 1) total Inner Duter = thermal Cetral for trarmsmutation 0.0 78.5 13705 0.0116 3.83 16 90 0ol : L Volume 2.1 m° =T . 0U15 203 0 0.0145 3 0.0145 (Cesium and strontium deuteroxide) Wall, moderator /3.5 88.0 Be G.060 4.48 « 1016 0 (. 08d e Volume 0.82 m’ ). U8 1.7 {Beryllia: 8 cm, raphite: 0.5 cm) Core (fuel) 86.0 34.5 2390y, 0.0014 4.03 1018 Jelume .57 3 240p 0.0004 0.0158 ‘ oo “Hlpy 7.0002 Na 0.012 Cl J.018 (Plutonium, sodium chlorides: PulCl_.+f NaCl) 3 i i 4.00 x 10'°® Wall 34.0 )7 .5 Fe 0.08 v T Reflaector 37 .6 200 fri 2.08 Boundary flux 2.4 x 1013 10° SR Fig.y Heutron spectrum in the core Total mean flux “.O}'lolb n Specific power 12.1 kvlcm-) Total power 7 dwth arbitrary units W 1ot e 1 [ 11l - Jdedtrop erergy grpup Q1 19 17 5 3 1 o sl |t 1 T T T T T T T T l i 1. 11 10 1 3 7 t S 4 3 g 1 J Neu%rlo;\d Wik nem Letnargy i.13 High flux burner reactor Cross-section of the reactcr Reflector Target zone (Sr(0D Yzt Cs{0OD)) T2t v ud T et 7 od TTUIhy T 1o 200 Radius, cm —_—— - Thermal flux = (2% th groups) \ y 4 : \ - h' Fast 1'lux 5 th group (1,39 - 5 Wall Modegaton Reflector Target zone SESSUNAGNNNNANNY 200 N 3 " m 2 2 9 =, - Radius {(cm) Considering chemical and physical properties, the logical mate- rials to be used inside the lux trap are hydroxide and/or deuter- oxide compounds of the FP. fFig. 4.14A shows that just a small proportion of H molar fraction has a large deleterious effect on the Cs-13/7 transmutation rate. This is due to the H absorp- tion cross section. Therefore, Cs00 and Sr(0D), are preferred. As Sr-80 and Cs-137 alsoc have their fair share of resonances it 1s advantageous to thermalise the flux before reaching the flux trap region contalning these targets. Therefore a spectrum con- verter between flux trap and fast fuel is needed. Begaring in mind the high temperatures to be obtained in this reactor and possible chemical reactions with molten salt, H50 and 0,0 are unacceptable. This leaves Be, Bel and graphite or some variant therefore for consideration. Be {and B) compounds, of course, have also to thelr advantage a relatively low (n,2n) threshold (1.67 MeV). Location nect to a fast region can therefore pro- duce considerable very slow neutrons in the flux trap - which 1s a main objectlive of the burner reactor. Replacement of Be by € or Mo wall material should therefore lower the FP trans- nutation rate, and it does. (Table 4.7 Fig 4.14B) indicates an optimum thickness of about 5 cm He. For the sake of safety and higher melting temperature, Be0 is preferred over Be. Table 4,7 Effect on Replacing Be Converter upon the Relative FFP Transmutation Rates Case materials A (Cs-137) . (Sr-00) i transm. transm. L Be, Be 1.0 1.0 2 C, Be 0.77 0.64 3 Be, Fe/Mo 0.34 J.82 (Nnte. transmutation rate in arbitrary units) S.204 Influence of other parameters The influence of plutonium concentration (Fig. 4.14C) is an im- portant parameter of thls reactor. Increasing the plutanium con- centration sipgnificantly improves the transmutation rate but also increases the power density above a technically feasable level (Fig. 4.140). Sne also Fig. 4.15, Transmutation Rate, Arbitrary Units Transmutation Rate 122 L.14 Impact of some selected parameter v (b) thickness of moderator (Rel), ( (d) power density, P (kW cm A 1.2 — - “sr ' = - -~ - -~ s \\ v37CS - 0.8 — Reference 0.6 A nE T T 0 0.1 02 03 H/D Ratio (a) C 1.2 o — > ” / 1370 7/ 1.0 ~ // / K cm 0.8 - Reference 0.6 -1 A T T T T ) 0 0.1 0.2 03 0.4 Molar Ratio of Pu (c) 05 Transmutation Rate, Arbitrary Units 1.2 — { ariations: (a) hydrogen/deuterium, plutonium concentration, and 0.6 = Reference T 3 5 7 9 Thickness, cm (b) D 10 - 5 - 2 - 1 = A Reference L 1 | ] 1 R 0 0.1 0.2 0.3 0.4 0.5 Molar Ratio of Pu {d) - 123 - Increasing the reactor power (Fig. 4.16A) from 5 to 11 GW(th] improves the transmutation rate. However a power unit above 7 GW(th) seems to be beyond the technological limit even for the distant future. The reference case has been taken as 7 CWlth). The use of beryllium [(Fig. 4.168B and 4.16C) instead of iron 1in the reflector of the core improves the transmutation rate sig- nificantly. At the same time the power density increases pro- hibitively (Fipg. 4.160). The raference case contalins 1ron 1n the reflector. Molten fuel offers the only way of handling the very high power densities of 10 GWm-°®. In addition the very steep gradient of fission rate makes a molten fuel core essential since the local fission density can be one order of magnitude greater than the mean density. In a solid fuel core the high heat removal rates would not be achlevable. The use of boron to absorb the thermal neutrons results in a de- finite decrease in the rate of caesium transmutation (Fig. 4.1/, The large size of the thermal flux trap results In the fuel re- oion approaching slab geometry with attendant high neutron leak- agr. To better epconomise on neutrons several possibilities may e tried - use of an optimised reflector such as te, Ni, Cu or He to minimise the critical mass - use of the outer neutron leakage for breeding - use of the neutron leakage for additional FP ftransmutation To begin with a solid Fe reflector was assumed (Fig. 4.18]). 1.2 Thermohydraulic consideraltions We now examine the thermohydrauvlic implications [(for more detail see ch, 8). The crucial parameter here is the core power density. The given value is high but still near the present state of the art (Table 4.8). For comparison, power densitiss for some high flux reactors. (Table 8) Transmutation Rate, Arbitrary Units 1.2 0.8 - 0.6 - Effect of variation in some selected parameters: (a) power, (b) power 124 - (b) reflector, (c) power density, P(GW m-3), and (d) fission density in fuel, F (fission em=?) Transmutation Rate, Arbitrary Units 20 - 15 -~ ¥ 9 Power, GW(th) (a) (c) | 1 ( 1 1 ' f 1 | } I t t 1 L} 1 1 i i 1 1 : ] ) ¥ ] [} , ] H I Beryllium . . . . . 3 Fission density (fission/cm ) 1.2 0.8 — 0.6 - 1015 - 1015 e 10|4 - BeO (b) Fuel B Vo Y Beryllium Wall 88 Radius, cm (d) 94.6 Impact of plutonium concentration 125 - Fig.4.17 Impact of natural boron 0,9 41010 \ Actavity \ of Thermal \E‘.err{al Cs~137 |[flux in ~Jux 1020/Se rue}zzo e ‘\‘ -—ncm 3 =137 3 activity 0,7 410 0,60’ o u,s 10 M ™ 0 10 10 10 510 Boron thickness, mm Impact of reflector 800 in the fuel / Tuu - Temp. °c ¢0o0 ~ Liquid phase 500 = PuCl ¢ a1 3 4u0 Caesium activity 100 = Specific -4 power (relativ) ol - Activity b ol vs-1% {(relativ units) £ Wy = ;- ; T T T T T T T 1, to N LY 0.t g4 D04 U 2.2 MY 0 Filutonium molar ratio Fig. 4.18 Uranium reflector 120 71 120 Volume Activity of fuel of Cs~-131 - k) £ ( 100 —100 SRIE BN 8o+ 80 / il / 70 70 / L OO -1 60 4 Reference Reactor (iron / y{iv'ny Beryllium 5 em as additional reflector lTable 4.8 Thermohydraulics of the High-Flux Burner Reactor FParameter Unit Value Total power CWlth) 7.0 Core volume m’ 0.69 Fower density kW em™? 10.5 Fuel density g om”? 2.35 Heat capacity, mass J g'lK“l .83 HHeat capacity, volume J oom ikl 1.95 Diameter of tube inlet cm 120 Inlet velocity m s ! 15 Volumetric velocity mig ! 17 Hrat capacity E 7,233 Temperature increase 0. . (outlet-inlet) - A Temperature of fuel 8 . € 7010 inlet Temperature of fuel OC 590 out lrt o Mean velocity of fuel - - . m 5 ‘ in core Cooling --- Out of core in heat exchanger cooled by sodium fhis seems tn indicate that a total specific power rating of about 1T kgPu//™W(lth) may bhe achievable. - 127 - Table 9 Power density in high-+lux reactors Power density GW(th)/m® in core volume in coolant volume Feinberg, research reactor 3- 8-10 Merlekes CM-2 [(Soviet Union) 25 5 FETE (USA) 1.0 2 Lane (Molten chlorides) 5-10 5-10 HFEIR (USA) mean 2 a maximum 438 8¢5 Phenix 250 (France) .48 1.0 this paper 108 10«9 Thio crucial problem will be the efficiency of the external heai pxchanger. In the following example some typical heat cxchancor characteristics are taken to demonstrate the culated for 11 GWIlEh). ~pecific heat exchanger power (consarvative data) Total volume of heat exchanger for 11 GW(ith) Volumetric fuel ratio Fuel volume in heat exchanger Fuel In the core heat exchanger piping Total fuel out of core Fuol in core Total fuel 1n system Mean specific power of fuel in the whole system 11 GW(lEh) H.3m?3 Plutonium content of fuel Power rating of whole system The postulated power rating for the whole system For this case calculated the power rating 1in the breeder power reactors possibilities (oAl [ oy} O3 - 3 L Om 07 OW/m3 L3 gPus/em’ L3685 kePu/MWlth) kgPu/Mw(th) .15 kgPu/Mu(th) - 128 - 4.4 Some results Parametric studies were made as variations around a reference system which assumed P = 11 GW(th) (X = 2+9, K = 4-2) and Rg7 = 78¢5 em. The flux trap 1s surrounded by 5 cm BeD converter, a critical fuel thickness of 6+6 cm and an outer wall. Fig. 4.19 shows the calculated flux distributions for such a burner re- achtor. Note that the total flux in the fuel 1s similar to that in the flux trap. The calculated fluxes lead to the conclusion that (total spectrum, flux trap) = V2«(E=0.0253) Tho result indicates the relative effect of X on the ratio R of the FFP transmutation rate to FP production rate for the reactor system. It can be seen that X should be kept as low as possible. Absclute results will depend on the Cs and Sr densities in the flux trap. A value of R = 1 was achieved for both FP nuclides at X = 4«65, The F.P. atom ratio there was (Cs-137) / (Sp-90) = - i) //.,’;5- Another Important problem is the relatively high flux in the outer zaone or leakage from the core. this flux can be used for two purposes: 1) for transmutation of other fissiocn products which have rather a high abscrption cross section e.g¢. Te-88 o°' = 22 harns tJi = 2.7 x 10° years [-129 oth = 728 barns tJ; = 1.7 x 107 years In both cases a fiux of 10'" - 10'°n cm-'s™! permits a rather effective transmutation rate of 1 years T-120 X = 7 x 10 %s ! ¢ = 2 years To still further improve these reactions the use of a beryllium mocerator in the ferm of a 5 cm wall in the ocuter region of the core has been calculated. This gave an improvement in the trans- mutation rate of Cs-137 in the inner target region but also a very slgnificant increase of specific power due to the scattering of the neutron flux in the fuel region., IThe possibilities for transmutation of these two long lived fis- sion products will be descussed further elsewhere. 4.1% Fission density in the fuel zone 101 Fission density fission s emtsec H 1 14 107 124 Fission density in the fuel zone Beryllium s S Wall 190 104,6 106,6 - 130 - The neutron flux outside the core may alsc be used for breeding in a uvranium blanket. The breeding ratio may be higher than 1 but the decrease of the transmutation ratio is critical and seems to be too low for a burner reactor. Nevertheless 1t could be useful to check these possibilities in more detail using some of the available neutrons for breeding in an external blanket region. Table 4,9 summarises the data for the reference case. It can be shown that the transmutation rates obtained are egual: 90+, b = 2 = . - 0 —85_1 ol "?"F'F tf' + >\8 ] @"'XB q.flj X /]O /2 i te?? = 1.85 vyr ‘7 et 710 P37Cs: = U.2460 107871 Cs AS{¥ L ooox 10 e 1772 A - tQFF .05 yr I, ]/7 LY/ T = 3.3 B eff The results obtained are significant but rather pessimistic. To estimate the "profit” of the transmutation process, the hazard index (H) must also be taken into account. From Table 4.70 1t can be seen that the total reduction in hazard from both fission pro- ducts mguals 13.5, which is a better indication. L5 Comments on hazard coefficients It is perhaps valuable to estimate the usefulness of using the concapt of hazard coefficients 1n fission product management. Table 4.10 gives socmo values for Sr-90 and Cs-137. From this it can be seen thalt in a steady state transmutations requiring the amount of hazardous substances 1s reduced by a factor ~ 15 1In relation to the steady beta-decay. Table 4.9 The Transmutatbion Process faor Soleocted Fission Products rner reactor 7 o« A0%W) s fission rate in il ) ) > S | W I notal power P Lel Dat= faor Froperty Symbal Unit 902, 13704 Yield of fission prcducta Y --- 3.041 0.064 Production rate RD = F .y atom 57! 3.81 x 10172 5.05 x 10173 tom/10% " em?® Concentration C a{o?gw trm 0.0016 0.0116 @] & DEE' Volume of targe: ) cm’ 2.03 x 10° Number of atoms N =V s C atom 3.25 x 1027 2.35 x 10°%° Pecay constant AR 57! 7.86 x 10710 7.33 x 10710 Dzcay rate R = NAR atom s ! 2.55 x 10%°8 1.72 x 10t° Mean cross section 0. 074 om? .29 3.045 Total Flux, moan D n cm_z 5 ! 3.83 x 10t® Transmutation rate Rtr = Neg oD atom =t 3.61 x 10%° 4.05 x 10%° C Total destruction rate Rt = R+Rtr atom 5 ! 3.80 x 101° 5.77 x 10t? Vz i Destruction canstant tpf{ = IHZ/AdfF year 1.85 8.495 ici ( -rancsmubati 1 |7 Efficiency of T?ér mu 3.10H - . t/Z . /2 o 19 3.9 Inventory reduction ratio tr B Teff Steady-state equilibrium RO R 3.8 x 10b? 5.8 x 10t° B YTar the mean value of She yield, see bt S b, Tacle 4,17 =azard 1ndex for Che Improved Reforenoc Case” Ratio - - . - 37 90 Carameter Symbol nit ERere L7y 5 1 370 Gaximum permissicle concentratian water 3.7 o« 107! 14,8 49 air H 3.7 x 10770 2.2 x 107 ° 55 umeann SO Yield of Ffission products from “¥3(0 Y stomn/fiesion 1.0572 0.935 from 23°U Y atom/Fissian (51 1.1599 from 23°u Y stom/fiasion .24 J.10669 Mean value for fission rate aof ) 2 ) o L2307 %py 1011 Y 7.041 N.754 Hazard HeY - 2.5 0,064 Z2.114 Efficiency of hazard reduction by transmutation HeY /B4 . - - 137 .13 7.156 (Table ITI) Mean effective hazard 2.114 i3 s reduction 0.156 : *The hazara ooefficient (H) is defirn-d as the amnuat oFf air and/or water neei-d to dilute o2 amount of a ;pivaen nuc_ide present levels propoosaerd £or the maximum permissible concentration. B AN’ Fig. 4.21 Inventory of Strontium-90 in the power reactors and in the burner reactor Power of total system %% uWth Yield of Sr-90: 4,1 3% Number'] of Sr-40 atoms Inventory in the steady satate due to teta-decay only 1079 o (eg. in sal L1 — = ar— o /' 8.m,~10‘q atoms A ” 28 10 - inventory power reactors, mean 4 b il atoms Inventeory in burner reactor 0. 461077 atoms 107 o 1026 Charging, discharging Production rate of in power 5r-90 reactors 2 ~7v1019 atoms/sec 107 Y T T T 9,01 0,1 1 1 1990 1000 Time, years z aard ol T products: Ure-doand Cs=1137 1 Heacltor nacarid (B Total hacard Larie tranamat at fon 1 - time, years 4 1 - — -4 1 T T T T ¥ T i, years - 134 - The amcunt of strontium-30, the most hazardous nuclide, in the high flux burner i1s about the same as that found in the power reactors after 3 years of operation. However the most impressive result comes from considering the end of the fisslon power area. Where by compared to storage (natural decay with a combined half life of 39 years) without transmutation, the transmutation case shows that the amount of nuclides remalning will be reduced by a factor of 1000. (Fig. 4.22). In a high flux transmutation the reduction by a factor 1000 would be achleved with 26 years in the lifetime of one reactor generation, 4.7 Secondary processes It must be remembered that the relatively high neutron flux re- sults 1n the irradiation not only of the non-desirable radio- active nuclides but also the stable fission products, including the stable components of the fuel and structural material. A very short review of these partivular processes is given here. Natural chlorine contains two stable isotopes (underlined) 15 36 R~ 36 /8L Guy) OCL STTET s 64T volatile 37 38 B _ 0 e N VA I F LV -t Y tata However not only (n,y]) reactions are important here. Much more important is the followling raction 35, 8- 35, Cl (n,p) 18~ mg T — N ‘1 yat (see ch. / for experiment results) The presence of sulphur also influences the chemistry of the molten fuel (see ch.G6). Sodium having only one stable isotope is also transformed 23, 24 B 24 25 26 fhis chailn gives rise to increasing amounts of stable magnesium and over larger periods also stable aluminium. 2.9 Conclusions The system proposed for the transmutation of °°Sr and '?7Cs, fulfills the following criteria: 1. The energy halance 1s positive. J. The hazardous waste balance 1s stronegly negative. That is, the amount of hazardous material destroyed oreatly exceeds the amount of freshly produced, e.g. tritium, '°B, and the activation products of the structural material. 3. The rate of destruction (transmutation) is aporoximately at ieast one order of magnitude greater than that due to spon- taneous beta decay. 4. The periocd in which a thousand-fold reduction in the hazard can be achlieved is the same as the lifetime of one reactor, Chat 1s, 20U to 40 yr. The neutron halance of the system is positive. That is, i1t permits breeding to occur alone with the transmutation. B. The weakest feature is shown in the relationship between the nrobability of catastrophic release to the environment for the transmutation operation, Pyring, to the probability of a simllar event in the case of storage, where it is desired that Pstorg’ 7 - 1306 - where A, 1s the decay constant. B Further optimisations of the system are possible. The proposed system, a molten-salt fast reactor, while rather exotic from a technological point of view, is not as far re- moved from the present state of technology as some other trans- mutation proposals (e.g. high-flux high-energy accelerators, controlled thermonuclear reactors) may be. - 137 - 5. An internally cooled breeder with uranium-plutonium fuel 5.1 Design features and objectives In this chapter a further variant of the molten salt breeder is described. The core 1s cooled by a cooling medium circulating in tubes. The molten fuasl is 1Intensively miced. It remains in the core and no fuel other than that being drawn off for reprocessing is present outside the core. The unique feature of this concept 1s the use of & molten fertile material as the cooling medium for the core. Such a dual function for the fertile coolant results in some unusual properties for this reactor type. The study here is concernaed with a molten chloride breeder reactor. The most im- portant features are: (see table 5.1) - thermal power: 2.05 GW - 1.94 GW in core + 0,17 GW 1in the nlanket - electrical power: 0.85 GW in the optimum case | = 0.4) Derf - molten fuel consisting of {in mol %) 15% Pulls (of which Pu-239 + Pu-241 80% and Pu-240 = 20%) 85% NaCl {(no 2?80CI1,in fuel) 1 and fission products in the form of chlorides or in an elementary state. - molten fertile material (in mol %) nhe 7PfUCTS A0% NalCl and newly bred Pull,y and fission products - conlant flowlng in tubes: made up of the fertile material above. No other coolant in core. - blanket material: also made up of the fertile material above - the core is internally cooled. There is no circulating fuel outside the core,. - 138 - Table 5.1 OBJECT . INTERNAL COGCLED FAST BREFEDER POWER REACTOR REACTOR TYPE CECOMETRY : INTERNAL Z0NE CURE, Fuel (1) WALL : Iron, ‘lolybdenum MTERMED SIS i . ) PHTERMEDTATE Z00E cooling tube with fertile WALL : = EXTERNAL Z0ONE Fertile zone WALL, REFLECTRO Tron POWER (GW thermal) DR POWER DENSITY (GW therm/m? core) : 0.07 NEUTRON FLUX, MPAN (n/cm?s) 7 x 171° FIaSILE Pu-239/241 NUCLTDE - DILUTENT NaCl, UCL. COOLING SYSTEM @ Internal, in tubes by fertile material DRECDING RATIO : 1.38 PARAMETER STUDIED NDesign Tube materials Pu/lU ratio Temperature METHOD GF NEUTRONIC @ ANISH 54 CALCILATION ;23 Sroups P,oATD. G CODE material - 139 - - the fuel and coolant flow concurrently (Fig. 5.1) - the reprocessing plant is in close proximity tec the reactor ("under same roof") - the fuel in the core, and the coolant is pumped with a velo- city of 2 and 9 m.s™! respectively. - structural material: possibly molybdenum along with small amounts of other metals e2.g. Ni, Fe. The advantages of the proposed reactor are as follows - - no separate coolant, no "foreign” cooling agent (e.g. sodium, helium etc.) in the core which results in a more satisfactory system with improved neutron balance. - the fuel inventory is very small due to lack of a separate cooling system and the small out of core inventory due to the directly coupled reprocessing plant. - the fuel contains only plutonium and no uranium which simpli- fies the processing technology and removes the danger of uranium trichloride oxidation which alsc improves the cor- rosion properties of this medium - the high velocities of both fuel and coolant, significantly reduces the temperature gradients in the equilibrium state and reduces the mass transport mechanisms. These are very sensitive to ftemperature gradients and play a large role in corrosion. However the diadvantages are numerous: - the first and most improtant is of course corrosion. The mol- ten chloride medium especially in neutron and gamma fields, at high temperatures and velocities with chlorine being vir- tually freed in tLhe fission process from plutonium chloride presents a very serious problem which must, and probably could be solved. - the most likely structural material seems to be molybdenum alloy which among other thing gives rise to parasitic abscrp- tion of neutrons. - the fuel is circulated by a pump which must be located in or close to the core which increases the corrosion problems. - the high fuel and collant velocities result in high pumping costs and could cause severe erosion. 1o i hemat b} [: (= rcuit i HE = HEAT EXCHANGER cuT 1 CiR CRCUIT [T - 141 - Table 5.7 Molten chloride fast breeder reactor (MCFBR) "CHLOROPHIL" Electrical power, approx MW(elect) 800 Thermal power, total/in core MW (thermal) 285@/1940 Core volume m* 8.75 Specific power MW om” 3 220 Core geometry m height2.02/2.36 dia. Fuel: liquid PuClgq NaCl mol% 16/84 Liguidus/bociling point °c 685/1500 (appr. . Fuel mean temperature OC 584 Fuel volume fraction in the core 0.386 Paower form-factors radial/axial 0.60/0.78 -ast flux, mean across caore ncm %s”! 7 x 10b° “uel density at 984°C kg m 3 2344 Heat capacity, 984°C kJ Kg~ 1 k7! 0,95 Viscosity, 984°C ¢ em le! 0.0217 Thermal conductivity, at 750°C Woem b oK 0.007 Fuel sallt in core Kg 7900 Total plutonium in core/in system kg 2900/3150 Plutonium in salt weight = 36.4 Mean plutonium specific power MWth) kg_l 0.67 ::i?rzlzgzziim specific power in MA(ER) ke 1 067 Coolant liguid U-238 C13/NaC1 mol % 65/35 Liguidus/boiling pint °c 710/1700 Coolant temperature inlet/outlet °c 750/783 Coolant volume fraction in the core 0.555 Coolant density kg m” 3 4010 Coolant salt in core/in blanket kg 19,500/165,000 Total volume (85 cm) m? 47 .85 fuel (shell side, pumped), velocity m s ? 2 Coolant velocity ms } 3 Number of coolant tubes 23,000 Tubes inner/outer diameter cm 1.20/1.26 Tubes pitch cm 1.38 Breeding ratio, internal/total - 0.716/1.3886 Coubling time, load factor 1.0/0.8 yr. years 8.5/10.5 - 147 - Mean flux across core 7.0 x 10'°n cm™ %5 in centre 1.2 x 10'%n em 2?2571 Temperature coefficient of reactivity §k(%)/6T(°C) - fuel - 3.8 x 10°°2 - coolant + 1.29 x 10 2 Reactivity wall fo vessel 0.72% {10 mm thickness). SRR 5.3 Neutron Physics toceordine bo J0 Licou, 107 The 7 eroup transport calculation gives 125 cm {(8-18 m®) for tho critical radius of the core with a blanket thickness of 295 eom (36,42 m?). The detailed neutron halance is given in table 5.73. The relative fluxes for each eroup are to be found in tahle 1.3 for core centre and core boundary. The corresponding one group cross sections are given in table 5.4, In Fig. 5.2 the neutron spoetra (flux per lethargy unit) are compared to that of the fast critical facility ZPR - 3 - 48, This last spectrum is slightly harder but the spectrum of the molten chlorides fast breeder com- parcs facourably with that of a power LMFBR. From table 5.3 one can drduce the following parameters Ko = 1.384 Dreeding ratio Brcore - 0,716 Rblanket V. b70 HRtotal 1.386 For the eiven core (125 cm radius) the blankel Lhickness was variocod botween /75 cm oand 115 cm. Fig. 5.4 shows the variation of brording ratios obltained from the new transport calculations. The reactivity and the core brecding ratlo remain practically constant in this rankge making the adjustment of the core valume UNNECEeSsSsary. Above 100 cm improvement of the breeding ratio by Increasing the blanket thickness gives a poor return. For example to increase the breeding ratio from 1.40 to 1.45 requires a thickness increase of 70 em or a blanket volume increase of 32%. 143 Table 5.3 Neutronics of internal cooled fast breeder Core atomic densities (atoms x 102%) PU-239 6.5797 x 10 * atoms cm ? Pu-240 1.5699 x 107" U-238 3.5629 x 10 3 C1 1.8495 x 1072 Na 6.3017 x 1073 Mo 7.386 x 10" Fe 5.078 x 10738 Blanket (coolant) densities (atoms x 1024) U-238 5.4023 x 103 atoms cm ? C1l 2.2718 x 1077 Na 3.457 x 10 3 Neutron Balance Region Nuclide Atoms Abs tion Leak Producti o . i (em® x 1021 orp eakage roduction 2138 (n,y) 22.57 . 56”79 5. . . 3 25.30 (n,f} 2.99 6.23 239 [H,Y] 5.58 R FPu 0.66796 34,56 (h.f) 28.98 85.55 240 (n,y) 2.24 Pu 0.1663889 3.78 (h.f) 1.54 4,772 Na 6.3017 0.26 - (in fuel 1.10) C C1 19.485 3.16 - ore ! (in cool. 2.08) Fe 5.0/78 1.30 - Mo g.73886 Z2.04 - F.P. 0.0687 0.50 - Total core 71.10 27,40 38.50 238 {n,y) 23.15 U 6.427 23.70 (n.f) 0.55 1.50 Blanket Na 3.457 0.08 - Cl 2272 2.272 - Total blanket 26,00 2.9 1.50 Fig. 5.2 Chlorine absorpt. cross-asctions ENDF_B/III (Jan. 1972) Absorption cross-section of chlorine IT‘YT‘] Disggs : > 2 ] = - = E] \ /./ : iF o s # L o (BN - U e Lugdi 1o i luuLl L S gevs (mb) 2 = ° 8 2 va Fig. 5.4 ¢ welh Breeding Ratio C and blanket thickness t ( Core volume = 8.18 m’) o 148 i Bisnket Yeiume ¥ wo— LR} »- ’ Ref. Case 1.3 kB o t {cm) . i 1 A 1 L 130 n / " [ "e 1 L K2 b Fig. 5.5 J Sk Reactivity change with Core volume 1 b (Blanket Thicknesss95cm) 4 4 - * 1% P~ - b 1% L Ret.Case i . | b —=)c . I 1 (") b . (Y} ) 12 ~1% | EERLS 144 1187 e’ Fig. 5.3 Neutron Spectra and Chlorine Cross Sections (qu) %% 0 o =4 LA S SR BN LA 2R B S § T T T L N - q E 45 1. - e = = K] = i = el -1 = ] e I~ - g R e 82 + 1) 2 - = +« @& Cc L] | €858 3 a o O < [ 887 = o e S F &8~ 3 § T e &z ° PPRT & 2 I €068 © R + T IAARASLE A T T T T T 0.1 Mev AINN AQHWHIZT ¥3d XM14 Radiatl distributions ol the l.'. g™ fies (Sph. Geom.) transport calcuiations Asymptotic mede In The core i i P —— s e 190 r (cm) Fis. 5.7 Redist distributions of the 12'*11us P and specitic pawer (Cyl. geom.) On the other hand the reactivity depends on the core radius. Fig. 5.5 shows the variaticn of reactivity with increasing core size. Such acurve 1s very useful when 1t 1s required to translate the cost in reactivity of a supplementary parasitic absorbticn into an increase in core volume (or plutonium in- ventory). The transport calculations used in Fig. 5.5 could not be used to obtain the carresponding variations in breeding ratio, rather they have been calculated from the information in table 5.3. (assuming small core volume variation < 10% and no important changes in spectra). This gives 1) BRtotal - 1.386 - 2.45 8k (breeding in core unchanged) The effects of modifying the core composition can be evaluated by the same method. This is arranged to avold altering the preperties of the coolant and fuel by varying the coolant tube diameter. Ihe reactivity is very sensitive to this parameter. If ¢ and n are respectively the relative increases in fuel Table 5.4 0Une group cross sections based on the reference 22 group transport calculation (barn = 107 2%zm?) Type aof Core Soectrum Hlanket Spectrum Nuclide o Or v 9 Og v 238y 0.249 3.30 1072 | 2.746 || 0.326 7.71 107% | 2.715 239 0.329 1.709 2.951 || 0.536 1.841 2.917 240py, 0.527 0.364 3.052 || 0.790 0.174 2.998 Na 1.62 103 - - 2.044 1073 - - - .38 1077 i _ || 8.83 1073 i i (8.64 107 3)* (9.11 10 %)+ Fe 1.01 1077 ~ - 1.43 1072 - - Mo 0.109 ~ - 0.166 - - Fp 0.225 - - 0.362 - - * These valuos were computed on the bases of ENDE/A. IIT data. - 146 - and coolant volume for a constant pitch are as - 1.44 ¢ 1 + §.23 x 107 %¢ + 0.9027 n 17 + D.2/56¢ + 0.3944 n 3 1l 1e and k which gives Sk - - 0.924 ¢ n = 1.56 &k The corresponding relationships for breeding ratios (BR) are 12 0.716 - 1./72 6K core BRblanket - 0.670 - 1.05 dK 1 - 1.336 - 2. SK &Rtotal 1.336 VAV It can be seen that the penalty on the total breeding ratio for the same 8K is anly slightly greater while the penalty on the increase of plutonium inventory is five times lIess. Therefore a reduction in the diameter of the coolant tubes is preferable to an increase 1in coolant diameter, provided of course that an in- crease in coolant velocity is admissible. This last assumption is implicit in these calculations since the coolant density was kept constant. In the whole system the chlorine absorption represents 5.38%. We have seen that the GGC-3 values are different from the more up to date ones (ENDF/B-T1I1). On the basis of these new cross sections given in Fig. 5.3 (curve 6] and assuming that the re- ference spectrum is unchanged, a computation of the one group corss secltions pives 8.40 mh instead of 6.353 mb in the care, and 9.11 mb instead of 8.43 mh in the blanket. In this last re- cion the spectrum is softer and the increase of cross sections in the eneregy range [(E A 0.8 MeV) is almost compensated for by the decrease at bthe lower energies (10 keV - 0.0 MeV). The total absorbtion by chlorinn for the whole system is B.5/% ingstead of 5.38% giving a loss of reactivity of 1.2%. This loss could be replaced by a 1.8% increase in Pu inventory 1f a very small decrease (0.65%) of the conlant tube diasmeter is accepted. Otherwise by changing only the core radius a greater Increase of Pu inventory (10%) is required (fFig. 5.5) Fven with this latest data the problem of parasitic capture 1In the chlorine 1s not dramatic and there 1s no reason to believe that there is no need to enrich the chlorine *’Cl. This is con- sistent with the conclusions of Nelson. Fig. 5.3 eclearly shows the importance of the energy distribution (spectrum corresponds to the chlorine cross section minimum (65% of neutrons are 1in an energy range where opp <€ 5 mh). This fact was not perhaps re- copnised 15 years ago when fine spectrum calculations were not possible. This could explain the pessimistic conclusions of several eminent physiclsts. For the molbydenum alloy chesen (20% Mo) the reactivity penalty (2%) is guite acceptable. However the cost could rapidly become nrohibitive if the volume of the structural material and/or the molybdenum content should increase for design reasons. In the context of more detailed design studies this polnt may become more important than the definition of the proper chlorine cross- sections. It does however seem likely That the molybdenum cross sections used in GGC-3 were overestimated. In the simplified calculations, no core vessel was allowed for at the core/blanket boundary but all the required information is available - fluxes, one group cross sections etc. (Table 5470, Using the same alloy for the vessel (20% Mol the one group ma- croscopic absorbtion corss section is 2.56 x 10 Jem”™! giving a loss of reactivity: where e is vessel o, — P Sk (%) U727 e thickness in cm. For 19 mm thickness a value of 1.37% is obtained for loss of reactivity which would have to be compensated for by an increase of core volume of about 10% i.e. about 9 m?® instead of §.18 m’. A hetter solution would be an increase in the plutonium inventory of about 2.1% in the reference core (loss in breeding ratio 1.336 >~ 1.350). Transport calculations (GGC-3 + SHADUK]) have been made for dif- forent coolant fuel densitites, different temperatures [(in this rase with the density constant to determine only the fDoppler offect). The reactivity changes with respect ot the reference core are given in table H.bh. - 148 - Tnble 5.5 Reactivity changes Type of modification Parameter Reactivity changes ~1.Y (Keo) 51 - 5% -3.0% Fuel density 5. 3 { 1.1 (leakage) Coolant densit 59 .77 200 ke - 3% . e poient geEnsitby -1.23 (leakage) Fuel temperature +300°C 0.01% Coolant temperature +300°C -0.14% Ve high Full loss of coolant - #17 ne f 0 VEDY TLER ON Koo [ e GJ The partlial changes 1in ke or leakage are only approximate but the total reactivity changes are evaluated directly and are therefore more precise. The Doppler effect in the fuel is quite neglicgible due Lo com- pensation between the capture and fission praocesses. The effect of full loss of coolant is large and positive but considerably lower than might be expected from crude calculations (kee Changes in the reference spectrum). From table 5.5 one can deduce the feed back effect which is very important for kinetic studies o - ) _nlbkl | S IV N R k p fuel o tu=l pocoolant SR o The vold coefficient of the fuel (1st therm) is strongly negative. 1% voild CLENEEPEEE 0 gives a 0.6% loss in reactivity. If boiling occurs in the fuel it will be rapidly arrested by a decrease in reactor power. Lo oand I+ one considers that all density modifications come from thermal expansions (liquid phase only) one can define general temperature coefficlents. The thermal expansion coefficients are CS _ _ - -3 -1 (557 fuel - - 0.853 x 10°° K (90 ) - - 0.89 x 1073 K71} Replacing these values in above given equation leads to the fol- lowing expression °) T - "2 (3 " (%) 3.6 x 1077 (8T) . 4o+ 1.29 x 1072 (8T o in the second term the part played by the Doppler coefficient (4.8 x 107%) is quite negligible. For + 100°C in the fuel the loss of reactivity is - 3.8% which is very important from the safety point of view. Compared to any kind of power reactor (even the BWR) the advantage of this kind of reactor is qguite evident. For the Nelson (198687) value of the thermal expansion -3.107° instead of -6.3 x 107" one gets - 1.8% which is very close to the Nelson result -1.5%. If we postulate an accident condition and assume that the same increase of ccolant temperature immedistely follows the fuel temperature rise, the overall change in reactivity is defined by: This important isothermal and pessimistic coefficient 1s still nepacive. Nevertheless during a detailed study of this reactor concept 1t would be necessary to check the values of the ther- mal expansion coeffiliclent for fuel and coolant more carefully. The relative value of the coolant term which is positive might prove to be too high 1f the differences between fuel and coolant became too marked. This problem did not arise with the present data. - 150 - In a spherical assembly the fluxes in the core are given to a good approximation - sinBr AL fE) dlr,E) where the space function is called the "fundamental mode” (soclu- tion of V2y+B2y = 0 in spherical geometry). The critical buckling B? is obtained from homogeneous calculations based only on the cross sectlon data of the core,f(E) is the asymptotic spectrum which is space indepen- dent far from the core bhoundary. For the same 22 energy groups one gets: B? = 4.08 x 107" cm™?. Using this value a good fit of the "exact fluxes” have been obtalned from the complete trans- nort calculations (Fig. 5.6). The asympototic fluxes cancel for = 155.5 cm whero R, is the extrapolated radius. By definition the blanket saving 1s given by where Re 1s the core critical radius. The blanket saving depends mainly on the nuclear properties of core and blanket and on blanket thickness. However for thlicknesses greater than 60 cm this last effect is very weak. Finally the shape and size of the core have almast no influence on this saving. This parameter, for this reason so important in reactor physics, will be used In the next section for the one dimensional cylindrical calculatlions. The axial blanket thickness 1s taken to be equal to the radial thickness (95 cm), and the core height as H_ = 200 cm. The criti- cal radius of this cylindrical core has to be determined. The two dimensional transport calculations are too expansive (and un- safe) and only one dimensional calculations have been made, whilch is sufficiently accurate. Axial transport calculations are not re- quired since the balnket saving is known from the spherical geo- metry calculations. One can therefore assume the following flux shape. - 151 - ®(r,Z,E) - (c0s RZY(r,E) for any r value (including the radial blanket) and H H _c Zz IA ™~ A _c -2 -8? is the axial buckling computed from the extrapolated height: H8 = HC + 28 = 261 cm which gives g = fil = 1.204 x 10 2em™ ! [ (B2 = 1.450 x 107 %cm™2) The computation of core radius and spatial distribution have been made with the SHADOK code (cylindrical version) by introducing axial leakage defined by B?. Befare that a first approximation is obtained by introducing the radial buckling a?, that is to say assuming for the core only, the shype Y¥(r,E) = J (ar}f(E) where J_ is the usual Bessel function. One obtained a? = B? - g? where the critical total buckling is 4.08 x 107% cm™? which gives a’ = 2.63 x 107" ecm™? and a = 1.625 x 1072 cm~ 2. Then the extrapolated radius of the cylindrical reactor is R = = ~ = 148.0 cm. Finally with the previcus blanket saving we get a core radius proper of R = 117.5 cm. The direct transport calculations with SHADOK -code gives Ry = 118 cmi This cleariy indicates the value of the blanket saving concept. Nevertheless these transport cal- culations are still necessary because they give the radial dis- tribution of fluxes and specific power over the whole system and more detalled informations. Fig. 5.6 shows some of the radial distributions of flux and specific power. The energy production in the blanket is quite small (1.7% of the core power) because no fissile materials are present (and only fast fissions oceur in ??%U). In praectice it would be higher (say 5%) since the re- processing process would not be able to remove all the fissile nuclides produced even with continuous fuel (coolant) reprocessing. The radial form factor for specific power distribution is, for this core mean pPower o = ; 0.60 r maximum pPOower Note; it would be possible to improve this coefficient by choice of different lattices particularly the most reactive at the peripheral region. The axial distributicon of specific power is given with a good approximation. If the axial mean value is unity then this dis- tribution is: BH P(Z) = —_EJEEE—;OS B7 o 28in— ] juad l = T[ = P 1 2 Fl wilth R v T o8 1.7 x 10 Cm 3; o and H_, = 200 cm giving P(Z) = 1.284 cosn Z—é%”—;[_mo < 7 s 100 and a-, = 0.78 Ffor the axial form factor. This axiael distribution 1s very close Lo that used by Nelsan. The critical volume is higher for cylindrical geometry 8.75 m’ compared to 8.18 m’: this increase was expected. Usually the number of energy groups required for a good defini- tion of neutron spectrum is at least 17 for fast critical assembly studies and 22 can be considered desirable. Therefore a 27 group cross section set has been prepared, the code GGC-3 which allows 99 pgroup calculations for a rather simple geometry has been used for this condensation. The cross sections were produced separa- tely for core and blanket and the scattering anisotropy was limited to Py which 1s sufficient for this reactor type. Most of the GCC-3 library data were evaluated by G G A before 1967 but some are more recent. - Iron - evaluatoed from ENDE/BT data (Feb. 1363) - Molybdenum - evaluated from isotopes of ENDF/BI date (July 1363) - Pluteonium 239 - evaluated from KFK -750 Resonance Nuclide {(Feb. 1969) New data concerning chlorine absorbtion cross sections are available at EIR (Fig. 5.7) they are obtained from ENDF/B-III (Jan. 1972). Unfortunately this information came too late to be used for the transport calculatiocons. Fig. 5.3 (curves 5 and 6) shows that the GGC-3 values were underestimated above 0.6 MeV and overestimated between 10 keV and 0.6 MeV. The effect of this on the reactivity is not great. Taking also the molybdenum cross section from ENDF/B-III one can see that the GGC-3 values are too high. (experiments made with molybdenum control rods in fast critical assemblies could not be reproduced with ENDF/B-T1 which makes a new evaluation of data necessaryi. Fission product data are from Bodarenko cross sections for resonent nuclides are obviously shilelded. theory is included in the GGC-3 code, shown 1in table Nordheim (1961) some special data as (1664). The absorbtion The it requires Tanle 5.6 Resonant Atomic density - . R R T (K] C a om 2 ar Nuclide R. Np (x102%) R R i " " 238, \ 65.47 107 10472 0.58 925 215 7.10 20. (Coolant) ’ x ] - c b3 2395, 1,725 x ’iD—?’ 1257 0.44 83 25.3 23.0 1/8.3 (fuel) 4.32 x 107% 1257 0.40 83 105.0 [111.7 731.7 (fuel) - 154 - 5.4 Gafety Problems, Comments The molten chlorides reactor seems to be a relatively safe system due to the following rasaons - an extremely high negative temperatur coefficient of reactivity, since during a temperature rise part of the liquid fuel 1is pushed out of the core into a non-critical geometry buffer tank. The dumping cof fuel in case of an incident is also possible in an extremely short time. - 1n a more serious incident when the fuel temperature increases to 1500-170009C (depending on external pressure) the fuel hegins to boil. The vapour bubbles give rise to a new and unigue, very high negative "fuel vold effect” - the leakage of fuel to the coolant 1s probably not a serious nroblem because the coolant is continuously reprocesced. - the leak of coolant to the fuel for the same reason cannct cause large problems (provided the leak remains small). A rahter adverse property of such a molten fuel reactor is the need to Initially heat the sclidified fuel in a non critical perometry with external power. (e.g. from the electrical grid). This problem has been fully overcome in the case of the molten fluoride fthermal reactor (Oak Ridge National lLaboratory). 5. CHEMICAL AND RELATED PROBLEMS 6. Physical and chemical criteria for salt components The limiting criteria in the search for fuel, fertile material and coolants for internally cocled systems are as follows 1. 2. 10. 1. small elastic scattering for fast neutrons (Fig. 6.1) small inelastic scattering. low neutron capture cross-sections for fast neutrons thermodynamic and kinetic stability of plutonium and uranium compounds (Fig. G.2, t£.3, 6,47, melting point below 700°C in the pure state or in the dissolved state (Fig. B.5, 6.6, 6.7, 6.3, 6.9, 6.10). boiling point above 1500-1600°C for both pure and dissolved states {(low vapour pressure) [(Flg. G.117. stability against atmospheric constituents, oxygen, water carbon dioxide. (Fig. B5.727. good heat transfer properties and specific heat capacity (low viscosity, high conductivity ete.) (Fig. 6.13) good corrosion properties if possible (Fig. B.14) adequate technological or laboratory experience. relatively cheap. These wide ranging criteria are fulfilled best by the following compounds PuClq, UClR, NalCl (Table 6.1, 5.2) Flectronepativity Fnergy Decrement I Aot i Li— T T ' ST A AR AEET VNN 20T on 2F i 1A piues rar sl "o solutivn 5.1 Frey B.4 Lol Fael . trerial Tompaonent s v 4 4] . * ¢l N Lo tely i R B i -~ 1 1 Fell, B J N r 4 A'l ¥ &2 I 1 i a4 L uel - B - — onn o h . - ool Ul veJl ] PR i) - S tacl, Ladl [~ ] s - Anc \ L [N e~ sadi Ao ool I -+ 1 - all L - 4 [ S ) — - B — -+ - A | + RINEIE - —tan . . - Fie. 6.6 delting Point of - - Selected Salts }‘lg . 0.5 v s Chloride Fluoride 11004 T T T mole fraction mol fraction Fig. f.7 Furl,-Alkali Metal Chlorides Fig. 6.8 Phase Diagram for PuCl, /NaCl and ¥C1,/NaCl 342%5°C Coolant TEMPERATURE (°C) 700_ Fuel o w o 4 2 ~ 600 < w o B = w =4 — 500+ UCls solid b i | | ] | { | NacCt , solid ! | b 400 T N T T T T T T LOO Yy Y T Lo 'B]ST T T fi[ 0 1 20 30 40 50 60 70 60 90 100 5 101520 40 60 80 100 PuCly in SALT (MoL%) NaCl Mol% UCls - 158 - 4o Fig. 6.9 The System PuCl -WaCl-iCl, Fig. 6.10 Phase diagramme with thorium 00 1Y) 50Q, NaCl NaCl PuCl ucl Thel, PuCl} Fig. "3l Vapeur pressure - metai chioreies “ig. .t CHLORIDES-OXIDES EQUILIBRIUM DIAGRAM AT 1000K Cs ta - 400+ ° CHLORIDES MUCH MORE STABLE THAN OXIUES Crile NiCL MnGly ZrCls L1750 ¢ CHLORIDES Id 815 (987 190 J 1207 1490 C mg' a EQUIL:IBRIUM Y N P2 Na ) WITH 041, y . = ~300+ Pr o 2 ' 104 2 . N ir ®ite — ) ug® Pu - S -2004 T OLIDLS MUCEH AUEE 104 -~ 5 ® Al SVABLE THAJ Cho o RITEG w 4 x 2 o a cr ® 6 w o X ® Mn 10 4 x Maxmmum L2 Fog ! tuel temperature -100 . ® .y in this reactor - 104 ¢ [ ] Mo T T T T Y Y T 16" -100 -200 - 300 v v v y v T v v 200 400 600 800 1000 1200 1400 1600 aat%% ox1pes (k3.1/2 mo17l0) r Temperature [°C) o - 159 - Pig. 6.13 Salts properties at 650°C v. Chemical composition Data derived from Nelson for UCIB/PuCIB/MgCLZ/NaCI 0014 4 ) 0-0086{[wemaeg’] 0-00A 1-031 083 [%Jd'f’] Ce 0634 ' o l_ B Lo 454 VISCOSITY COOLANT ‘g o ' FUEL & € 3 4 a " T] 404 Eo;—an s’] w8 < Q 35 i 58 From Nelson (interpolated)MgCy °3 { o ~ 35 3 [g.cm'fl ! OLANT O w P 30 DENSITY I FUEL 254 ] T T T T 0 20 30 40 50 [ua %) Fig. 6.14 Free enthalpy of formation chlorides 01 octe 300 sbo 1000 500 2boo TEMPERATURE [K] Table 6.1 Properties of - 160 - fuel components PuCl UC1 u 3 3 NaCl Molecular weight 348.3 347 58.4 Postulated molar ratio-fuel 0.15 - 0.20 - blanket material - 0.65 0.20 Density solid state (kgem™3) 5.7 5.57 2.14 Melting point (°0) 767 835 800 Boiling poént at atmospheric 1730 1790 1065 pressure ( C) Melting enthalpy (kJemol~™!) 64.0 64 28 Enthalpy of vapourisation - 240 300 188 (kJemol™ 1) Temp coeff. of density (K™1) 0.0010 0.0010 0.0005 Specific heat (Jemol~! K71 140 140 77 hermal conductivity (Weem™ ! deg™ 1) Viscosity (geem™! 71 Fuel: 0.025 Coolant: 0.045 0.0143 :ET?]COQ$€. of viscosity 0.0005 0.0005 N _ £ ree enthalpy of formation 70 575 990 a4t 1000 K (kJemol™ 1) Table 6.2 - 161 - Other chlorides of plutonium and uranium Melting point ("c) Boiling point (70) Froe enthalpy of formation at 1000 (kJ/mol) Plutonium Uranium PuCl e LUC1 ucl 4 4 g 5 All efforts to produce pure solid PUCI4 have been unsuccessful; 550 (287) 178 only in gaseagus atate with free 747 (417 (372 chlorine, or in molten salt solution or in aqueaus solution as complexes T =180 = =760 Ax-18 x-165 Ex-130 = -/B8 = -B70 = ~/8( - 167 - 5.2 Corrosion of structural material H.7 Ceneral criteria It 1s clear that one of the most problematic areas in molten salt reactor technology i1s the area of corrosion. Some criteria can be formulated as follows - Lhe free energy of chlorides formation for structural materials must be relatively low, significantly lower than those of pluto- nium and dJranium chlorides but still lower than those of the main fission products - the partial pressure of the chlorides formed from the structural materials must be rather low which corresponds to & relatively high boiling point for these chleorides (Fig. 6.16) - the neutron capture cross secticn for (n,y), (n,p) and n,2n) must be low (see later] The structural materials are in principle different for the two typrs of core discussed. - internally cooled, using tubes plus the effect of the cooling agent - externally cooled by pumping the liquid fuel cut of the core These two variants call for different structural materials and different requirements hoat conductivity . : mechanical Cocling method of the structural A L, behaviour material internally by tubes veTry good very pood (thin wall tubes) externally not important not so Important f formation (KJ mol™h) Free energy Fig, f.16 chlorides = b ling p(vh_‘t v Free enthualpy of Pormation ch’ .PuCh J SALT- FUEL COMPONENTS 15004 M’Clz N.aCl ZrCla [ ) [\ LY 3 ? Wely :5; PtCle - 5004 AuCl 0 100 200 300 Free Enthaipy of Formation Afix [KTMOR:‘] Fig. 6.13 0— 50 -1m_ 150 H,0 HZCIZ ~-200 1/2 00, - 250 ———— 300 500 1000 1500 Temperature K 163 FREE ENTHALPY OF FORMATION Periodic Table [K.T- mol“Cl] o bl 80 se0 T~ T idos 1500 Motybdenum —Metal " Melting Point-2610°C (P=1022 g.cm Boiling Point-5560°C o4 —— - - = 04 (4lpropcnimnion 1 ~3 o r 1000°C TEMPERATURE °K Fig. 6.20 Pissjon Products in Molten Chlorides Media 36 Kr | 54 Xe fast o Gas 1 35 Br 53 1 FPE Extraction °4°% very slow 34 Se 52 Te 33 As 51_sb FPS Volatile chlorides 32 Ge 50 Sn 31 Oa 49 In FPS Low volatile 30 Zn 48 cd chlorides 47 Ag 46 pd Non 45 Rh FPE volatile 44 Ry metals 43 Tc 42 Mo Low 41 Nb FPS volatile chlorides 40 zZr 64 ad 63 Eu 62 Sm 61 Pm Non volatile chlorides 60 Na | FPA 37 Rb |55 cs FPA Low volatile chlorides - 164 - From these and other criteria the following choices may be made - for the internally cooled core tubes: molybdenum - for the wall of the gspherical core, in the case of the externally cooled reactor: graphite and berrylia. 6.2.2 Molybdenum as structural material The main corrosion processez result from the followling mechanisms (m = metallic phase, s = salt phase, Me = metallic component of irradiated fuel or coolant) 2 X > frd %4 Mfa) + (m) For the behaviour of fresh fuel PuCl., in NaCl tho most likely reaction is (T = 1250 K] AG T =+ 450 kd/mol C1. The equilibrium constant of this reaction is small and nguals 10717 50 that this reaction is completely unimportant. In the bionkelt zone the most dangerous reaction is connected with uranium trichlorides (chlorine from the fission aof PUC]H]. the nontrol of the UC1,/UC1 ., ratio in the fertile ccolant might he frasible due to the continuous reprocessing of this material togother with the cantrol of zirconium from the fission products oxidation state. An additional problem comes from the fact that molybdenum has different oxidation states +2, +3, +4, +5 and all of them have the corresponding chlorides. (see Fig. 6.17) Futher difficulties arise from the problem of the reactions between metal chlorides and oxygen and water. These reactions (for oxidation state +7) could be written in simplified form The metal oxides are mostly insoluble in molten chlorides which results 1n a serious disturbance of the fuel system. From this point of view the metallic elements could be divided into three classes (see Fig. 06.12). - those which are stable with H.,0 and 0,, that is the chlorides are more stable than the oxides (e.g.LNa, Cs, Bal) and partially Ca. - those which are not stable with H, 0 and O, and the resulting nroduct is a mixture of chloride, oxychlofide and oxide (e2.¢. Pu, U but also Zr, T1, Al, Fe, Cr, Mn, Mg - this is the most numerous group of metals). - those in which chlorides are converted to the most stable oxide in the presence of H,0 or O, (e.g. Mo, W) metals of this class seem to bhe less numerous than thaose in the other two classes. This property causes the rapid elimination of traces of water or oxygen in the molten chloride salts of Pu and U. It is also well xnown that traces of H,0 and 0, have a very big influence of the corrosion rate. Molybdenum belongs to the last class, the oxide is much more stable than the chloride {(Fig., B.14) This means that the traces of exygen or aven water will result in the production of molybdenum oxide. This effect requires consider- able further study. — ’] 6 [“) ~ 5.2.3 The irradiation of molybdenum and iron in a fast high flux reactor The high neutron flux irradiation causes physical and chemical changes 1n structural materials. Molybdenum is a mixture of stable 1sotopes. The most important by-product of neutron irradiation is the Tc-99 beta- emlitter with ty, = 2.1 X 10° years and belongs to the decay chain shown here. Mo -498 R B - , i - 94 — -94 (;%Q] (n Y] 10 'LU = (_]Dh TC _t_’!r? _ > ,I % ,lflsv (n,y) g . ~ - 400 ; -100 T P00 y I Ry T For approx 13090 kg molybdenum in the core in the form of coocline tulbies or about 13,300 moles. the Mo-898 gives 2300 mol. The irra- diation rate N (mols/s! equals r\'f}’,_Lfi‘A"‘ RIS e 40 23 W =D 7 L1 R N = (2LANTEY) s (Bx1073) o (I0x10727) it e = 1 2k 0 Tatam/s After an irradiation of /00 hrs a steady state concentration of Mo-98 is reached ByeYe i b7 rteagy T TEomeT 7 3x107'atons - 0.005 mol state The radioactivity of the Te-384 after three years irradiation of 1000 kg of molybdenum in the fast reactor core: - 167 - Activity Ig—izarJ = 1.2x10" "atoms/s x (3x3.7x107s/year) 107172 T T 3 Curle/tonne of Mo The diffusion rate of hydrogen from the molten fuel to the coolant and blanket {(here also UCl3 - NaCl) must also be mentioned. Une can assume that this melt containing hydrogen is saturated so that the porosity of the wall {(molybdenum) will play a minor role. The most important factor is the variation in the mechani- cal properties of the molybdenum caused by the uptake of hydro- gen. The problem of molybdenum corrosion in chlorine containing media is particularly complicated by the numerous molybdenum chlorides: MoCl.,, MOCIB, MDCld, MDCIE (Fig. B.17) A0 Fission product behaviour in the fuel The fission of Pull, causes the formation of two fission products ' oand BT and three atoms of chlorine PuCl, S O T For the fissloning of 100 atoms of Pu the following balance has been sugegested Li;i; 0,008 Se + 0.003 Br + 0.842 Kr 3 gas gas g8 + 1.05 RbC1 + 5.43 5rCl, + 3.03 YE13 + 21.5 ZrClq 100 PulCl + 0.29 NBC1_.(?) + 18.16 MoCl, + 0.28 MOClS + 4,01 Tcm Fa 5 et + 31.45 Ru + 1./73 Rh + 12.066 Pd + 1.88 AgCl m m m et et et + 0,66 CdC17 + 0,06 InCl + 0.325 SnCl, + 0.687 5bC1, + /.0b TeCl, + 6078 1 v 21023 Xe + 13.34 CsCl z £as oas + d050 Ball, + 5.74 LaClg + 13,98 Cefilq + 4,28 PrCl3 + 11.87 NdCl? + 1.44 Pm[filq + 3.74 SmCl% + LB EuCl + 0.03 CdClB the average balance of fission can be represented in the fol- lowing manner Wl ogg een 100 PUC13 15 - 1RA - for Table 6.3 The behaviour of fisslon products in the molten chlorides fuel. (Yields given represent products for 100 Pu atoms fissioned). Principal fuel Fission Remarks and structural products material 04 -Mo metal Pd(12.66), Tcl(4.01), Rh(1.73), (struct.) 2™ Rul(3.744), 1(5.17), Te(7.865) - Xel21.2), Kr(0.%4), Br Q — & — 0} - ol e no chlorin=e . c é the synthesis of = il possible metal £ . chlorides L] D—_ ]'_:)} - - 14 =facl -1 -MeCl (18.18) 7 X " e (corrosion -MoCl., (0.28) 0 3 c product) c N o T in -AeCl (1.;8) 3 | ; —Sb613 (0.57) o | -CdC1. (0.66) : | 501G (0.32) | B 5 | -InC1" (0.08) " 12 ! £ Couct. (1) ad (, 200 g Eh - i "+ -UC T, -ZrCl, (2.15) | 5° e - - no0 Q - L & - —Lr612 I CL C m — . - -Pull, -YCl, (3.02) | %o - - -PrCl, (4.28) o s -EaCl, (1,068 = N —C@Cl% (13,94 . 5 -LaCly (5.78) 0 & 3 L { E A0+ -NaCl -RLHCT (1.05) 5 -CsC1I (13,75 I -omCl, (3.73) -S5rC17 (5.48) - -Ball, (8.50) ) From the earlier published data (Chasanov, 1965; Harder et al., 19685 Taube, 1861) it appears that the problem of the chemical state (oOxidation state) in this chloride medium for the fission product element constituent requires further clarification. From a simple consideration it seems that the freeing of chlorine from the fissioned plutonium is controlled by the fission preduct elements with standard free enthalpy of formation up to ~ 20 KJ/mol of chlorine, that is up to molybdenum chloride. The more ’noble’ metals such as palladium, technetium, ruthenium, rhodium and prob- ably tellurium and of course noble gases: xonon, krypton plus probably iodine and bromine, remain in their elementary state because of lack of chlorine. Molybdenum as a fission product with a yleld of 18% from 200% all fission products may remain in part in metallic form. Since molybdenum alsc plays the role of struc- tural material the corrosion problems of the metallic molybdenum or its alloys are strongly linked with the fission product be- haviour in this medium. The possible reaction of UCly and PuCly with Mool resulting in further chlorination of the actinigdes-trichlorides to tetrachlorides seems, for PuCly very unlikely (AG'?00K - - 450 kKI/mol C1) but this is not so for UCL,. A rather serious problem arises out of the possible reaction of oxygen and oxygen containing compounds (e.e. water) with FulCl and UCls which results in a precipitation of oxides or oxychlo- rides. The con®incuus reprocessing may permit some control over the permissible level of oxygen in the entirs system as well as the contiinucus gas bubbling system with appropriate chemical re- ducing agent. Corrosion of the structural material, being molybdenum is also strongly influenced by the oxygen containing substances, A pro- tective layer of molybdenum however, may be used on same steel materials using electrodeposition or plasma spraying technigues. Note that all these considerations have been based on standard free enthalpy: but even a chanre in the thermodynamic activity trom = 1 to = 0.001 which means a change in free enthalpy of 174kJ mnl~ ! thus appears Insignificant as far as these rough calculations po. In the fertile material reactions also occur and the most impor- tant are fission process: Jd€l, ——= Fiss.products + 3C1 . 1250K oxldaticn process: UCl3 + V2C12 — UC14; AG “7 =25k emo; ! : : o - disproportionation UCl3 + 3UC13 3UCl4 + Umet - 1/n - t.4 Scme comments on reprocessing Breeder reactors as is known form part of a breeder system which includes not only the power reactor hut also the reprocessing plant. The advantages cf molten salt bresder reactors become particularly apparent when the reprocessing plant is under the same roof as the power reactor and when chemical separation processes take place in the high temperature molten salt media in a continuous cycle. From Fig. 6.2° it can be seen that all fission products might be classified into 3 classes. FPA = fissiaon products of alkali and alkali earth but also rare earth elements which have free enthalpy of chlo- ride formation greater than those of PUCIB. FPS = fission products of seminoble metals with free ent- halpy of formation smaller than those of PuCla. FPE = fission products existing in elementary form because of the low free energy of chloride formation oar negative balance of chlorine. The separatlion of plutonium or Uranium form the irradiated fuel by means of pyrochemical technigues cculd be carried out for oxample 1n the following way. Molten salt, primary phase Pu, 7 [(part of I remains) Metallic phase (part of FP remains) MJolten sall, secondary phase containing only Pu. This 1s the so called metal transport process. The proposed schematic of separation processes utilizing metal transport is given in Fic., 65.21 and b.22. Fig. 6.21 Fuel reprocessing flow Scheme FROM REACTOR PRIMARY SALT PHASE irradisted Fuel UCly PuCly FPA+FPS+FPE in NaCt PRIMARY SALT PHASE irradiated Fuel PA Chlori \ METALLIC PHAS SALT METAL in NaCy o g Ol \ Mg Metat in EXTRACTION Notten Metal RED!;COL‘-ON Tex 8007 ME TALLIC PHASE. Mg in molten metat Ymet Plimet 0FP§n + FPE ME TALLIC PHASE ME TAL SALT 49 m moltenmetal | EXTRACTION SECONDARY SALT / * OXIDATION PHASE 3 [Pha FPE et T = 800°C MgCly 1n NaCl = 3 = O ' z' PuCIgNoNcglCl: n Na z|2 FRESH FUEL o iBrELE MA g‘ SALT PHASE PRE PARATION MOLTEN FUEL ucly in T = 700°C NaCl FERTILE MATERIAL T0_REACTOR FPA — Alkali and Alkati earth fission products e.g- Cs Ba, Sr FPS — Semi and noble metals and metal chiorides FPE — Noble metals in metallic states and noble gases Fig. #,20 Tuel Heprocessing Mate T obala o IRRADIATED FUEL FROM CORE 379 Salt 7' K 210g Py s CONTINUOUS 00229 Pu-s 00459 FP- 5! FUEL REPROCESSING (50% EFFICIENCY FOR ALl FP) 370 St Pu INVENTORY ~ §00kg 70 Sait-s- 22gPu. st RENTENTION TIME 0022gFP- 3 DAYS -t 0-0225g FP-s FRESH FUEL TO dituted in v THE CORE Tesgsan B o 24 FROM REACTOR o £ 21-6gSalt-s” = 1359 U-s~! v 00860 Pu-s” CONTINUOUS : 0-005gFP- ' FERTILE MATERIAL REPROCESSING 50% EFFICIENCY Py recovery 0.033gPu s’ INYENTORY ™ 11600kg Udep! FERTILE MATERIA RENTENTION TIME 10 THE Fission Products c REACTOR ~10 DAYS to waste kS 2189 Snlt.s': 2 0033gPy- s Udepleted 0-0025g.¢~ 0:0025g FP- ' input In10'8g Satt. s~ 13-533g U+ s~ +10-8g fresh salt U-depleted 00355 g. s~ Pu gain gs 0-0115g-5~ - A7 - 6.5 In-core continuocus gas purecing g ging 6.5.1 The proposal In this type of reactor an in-core continuous gas purging of the molten fuel which can significantly improve the safety in an in-core accident, is possible. A mixture of hydrogen-helium gas is continuously bubbled through the liguid fuel in the core. The mean dwell time of the gas- bubbles needs toc be controlled and the mean transport time of the molten components to these bubbles must also be controlled (e.g. 1if speed-up is desired-intensive mixing, if delay-local addition of a further g¢as stream). The aim of the gas stri ing 15 as follows: g £ 1) to remove the volatile fission products which in the case of an asccldent control the environmental hazard. (1-131, xe-133, Kr-85 and precursors of Cs-137) and a2t the same time for the thermal reactor, removal of the I1-135, precursor of Xe-135, improves the neutron halance. 2] to control the production of delayed neutrons since most of the precursors and nuclides of this group are very volatile, p.g. + Hr-JI-isotopes. 3) removal of oxygen and sulphur, continuously (see Chapter 7) 4) «n s4tu control of corrosion problems on structural materials For the sake of a first approximation a gas flux of 30 cm® per sec. {(normal state) of Hy/He 1s arbitrarily assumed. At 20 bar pressure and with a dwelling time in core of 20 seconds, the gas bubbles will only ocoupy a fraction of the core equal to 10°° of its volume and have little influence of the criticality, (but the collapsing of bubbles results in a positive criticality coofficient ). The system proposed for continuous romoval of the volatbtile fission product from the core ilself has a retention time of some hundreds of seconds only. Each reprocessing mechanism which operates out of cere 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 prea- ter than that which gives a fuel in-core dwell time of about one week, Even with a 1 day dwell time, that is, if after one day the fuel goes through the reprocessing plant, no acceptable solution to the I1-131 problem is obtained since the activity of this nuclide is only diminished by one arder of magnitude. Nnly 8 direct in-core removal gives the dwell time in core as low as some hundreds of seconds. - 173 - 6.5.2 Delayed neutron emitters The principal question arise out of the fact that some of the short lived iodine and bromine {(perhaps also arsenic, tellurium) lsotopes are the precursors of the delayed neutrons. Table 6.4 Precursars of delayed neutrons for Pu-239 fast fission Probable Croup Half life Fraction Nuclide t V2 (seconds) % 1 52.75 3.8 Hr-8/ Z 22.79 28,0 I-137, Br-36 3 5.19 21.6 1-138, Br-29 al 2.09 32.8 T-135, Br-90 5 0.5493 10,3 5 0.216 3.5 As other possible nuclides the feollowing can be considered + As-8h; Kr-492, -93; Rb-92, -94; Sr-87, -98; Te-136, -15/; Cs-142, -143. The remcval of these delayed-neutron precursors from the core reduces the value of B, which is lower for Pu-2390 than U-235. Thus we have a problem of reaching a compromise between an as rapid as possible removal of the hazardous I1-131, and as long a dwell time In the core for the delayed neutron precursors: 1-140, I1-13%, 1-138, 1-137 and the appropriate bromine isotopes. In this case the mean dwell time of iodine in the steady state reactor is about 100 seconds. Tt can be seen that the activity of lodine for a 2.5 GW(t) reactor is of the order of only 10 kilo curies (for seconds) activity of approx. 10% (or 10% for 1000 second extraction rate]. The gas-extractiocn also influences 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 thelr associated precursors of iodine and bromine can be volatile at 1000°C. In elementary form: Xe, Kr, Te (7) In simple volatile hydrides: 82rH, IH In simple volatile chlorides: SnElz, SbClB, NbCle, CdC17. This amount of finally volatile components including I, Br, Xe, Kr amounts to approximately half the total fission products {i.e. 100 micromoles per second). In addition there is the cor- responding amount of tritium (fromternary fission). This amount of all fission products corresponds to a gas volume ratio of about 2 cm®/s or 10 times smaller than the postulated amount of hydrogen flow at 30 cm’/s. The extraction removes all short lived fission products which are valatlile under these conditions. Thus not only is the re- moval of the lodine isotopes and the consequent reduction in nroduction of xenon (e.g. for the atom number: A = 135, 130, 137, 138, 139) achieved but it slows down the in-core production of Cs-13/, Cs-138, Cs-1349, and then also barium-139 The higher components of the liquid fuel: PuCl, and NaCl. The Fuel consists of: i - 1o mols Pull,; boiling point 2040 K - 85 mol% NaCl ; boiling point 1728 K One can as a first approximation say that it would have the following compositicn in the vapour phase: 5 mol% PuCly - 95 mol% NaCl. ) The aorder of magnitude of vapour for pure components at a tem- nerature of about 1250 K is NaCl v 5 x 1077 bar; Putl, ~ 107" bar For the PuCl3-NaCl system one assumes here a lowering of the vapour pressure (thermodynamic activity coefficient approx. 0.1). At the postulated volumetric flow rate of 230 cm’ Ho normal per second, the vapourized amount of plutonium is given by: 30 cm? . -1 _ -8 TSE00 ot Tmn] X 10 bar x 10 10 mol Pu/s - 175/170 - This amount of plutonium is of the order of 107% relative to the amount of plutonium fissioned in the same time (approx 107" mol Pu/s). However, it still has to be recovered, which unfort- unately makes the reprocessing more complicated. Last but not least is the in-core gas extrection of two other alements - oxygen in the form of H2D : oxygen from impurities (i.e. PuBC1) - sulphur in the form of H.S: sulphur from the nuclear reaction: SCL (n,p) P8 : (s Chapter 7). I - ~J ~~J | /. EXPERIMENTAL WORK /.17 Chemical behaviour of radiosulphur ocbtained by *°Cl(n,p)?®°s during in-pile irradiation (accordineg to Janovici, 19/5) The rather large concentration of sulphur formed by *°Cl(n,p)®°S reacticn 1n the molten chlorides system proposed for the fast reactor makes it necessary to obtain the full information on the chemical behaviour of the radicsulphur. The most recent studies on the chemical states of radiocsulphur obtained by n-irradiation of alkall chlorides have shown the complexity of this problem. To obtain new data aon the behavicur of radiosulphur we have in- vestigated the influence of the time and temperature of irradia- tion and of post-irradiation hecating on the chemical distribution of the sulphur. EXPERIMENTAL — e e e = = — Sodium chloride ("Merck” reagent) was heated for B0hr at 2007 In an oven in vacuo. The dried samples of 100 mg sealed in cvacuated (107 "torr) quartz tubes were irradiated near the core of the "Gaphir” swimming poaol reactor for different periods at a neutron flux of 5 x 10'% and 4.3 x 10'% n cm™? s”!'. Reactor irradiations were carried out at an estimated temperature of 1509C and -190°C. After irradiation the samples were kept for 5 days to allow the decay of ““Na. The method of *°S-species separation. The crushing of the ir- radiated ampoule was made in a special device from which the alr was removed by purging with a nitrogen stream containing 10ppM of oxygen. After crushing a gentle stream of nitrogen was allowed to flow for about 10 min. The gases evolved were collected in coolerd traps. The irradiated slat was dissolved in 2 M KCN snlutioncontaining carriers of 877, CNS™, S05°%-, Squh. During dissolution oxygen was not completely excluded although nitrogen gas was passed continucusly throueh the system. The radiosulphur found 1n gaseous form was determined as barium sulphate. For the *°S-species separation the chemical method described recently by Kasral and Maddock was used. The radioactive samples were counted under a thin window Gelger counter. All measurements were made in duplicate with and without Al-absorber. Fost-irradiation heating. The sealed irradiated ampules were heated in an electric ocven at 770°C for 2 hr or at 830°C for gbout 5 min. and then cooled and crushed in a closed system under a stream of nitrogen. Results and discussions Effect of length of irradiation time. As can be seen from Fig. 7.1 S2- remains the preponderent fraction independent of the irradiation time. Formation of S°- is indicated by charge conservation during the *°Cl(n,p)3°S reaction. Alternatively it can be supposed that reduction of sulphur takes place by capture of electrons due to the discharge of F-centers. The presence of 5% in this oxidation state in the lattice is no longer contested. The precursors of higher forms may be S* as a result of an elec- tron loss from 5°. However, the interaction of chlorine entities formed by irradiation with radiosulphur tao form species as S5C1, sCl>7, SCl, may be an important mechanism 1in forming the per- cursors of sulphate and sulphite. During longer irradiations some of the sulphide 1s converted into higher oxidised forms. This may be a consequence of radiation- produced defects with oxidising character (e.g. V-centres or deri- vatives). It 1is possible that the concentration of defects re- sponsible for reduction of the sulphur decreases by annihllation when new traps are formed. The oxidation of radicsulphur with increase of radiation damage concentration may alsc be due to the reaction of recoil sulphur with chlorine atoms. The presence of OH™ in the crystal must not be rneglected. It has been suggested that radiolysis of OH™ can be responsible for accelerating the oxidising process. Effroct of post-irradiation heating. The effect of post-irradiation heating (including melting) can be seen in Table 7.1. Comparisons between heated and unheated samples are made for irradiations of 2, 12 and 724 hr. For 2 hr irradiation, results on samples heated at a temperature below the melting point of NaCl are also presented. As is seen, on heating, a part of the radiosulphur is found in a volatile form. The veclatile radio- sulphur appears at the expense of S° and higher oxidation forms. The results show that with temperatures above the boiling point of sulphur and above melting point of NaCl the S° and &% and/or 5¢Cly receive sufficient kinetic enerpgy to migrate to the surface or evoen to escape from the crystal and be collected as volatile radliosulphur. However, there are some differences in the 3og- chemical distribution on heating below and above the meliing point of NaCl (experiments Z2-3). 1t seems that for relatively short periods of irradiation (2 hr) only the sulphate and sulphite precursors account for the volatile radiosulphur fraction. For a longer time of irradiation, on melting the S° value decreases to about 2% and this corresponds f£o an increase in the volatile radiosulphur (experiments 5, 7). However, a small and practically constant yield of S° is found in the melt after longer ilrradiation Igble /] Irraad Fost-irrad. 2 - 0 2 JENU . o} [0 M * o Expt. e b ot e 55 S SUy ; 302 S volgtlle 1 Z hrs no /3.1 £ 0O, 9.8 * .3 16.9 £ 0.8 0.01 Of_‘ 2 " K 75.4 5.3 + 0. 3.6+ 2.3 15,4 + 1.1 Z hrs. nec " 59 ) /702 £ 2,1 11.0 = 1), 5.6 £ 2.5 5.0 £ 1.4 5 min. 12 hres. no 57.5 = (1.7 12.1 £ 1.1 Z0.4 + 0.6 .01 0 n " BSU,C 7.15 1.9 18.49 7.6 5 min. 24 nrs nag o4.4d4 + 2.5 1.3 £ 0.5 23.7 + 2.0 0.0 § 330°C ) 7 ) 8.2 + 3.4 2.3 £ 0.5 Z1.4 + 2. 7. + 0./ 5 min. Expt -5 = 4,3 q0t? -m~ 2 57! Expt. B-7 = & 104 noem™? 57! Culpnite fraction 1o less than 5% in nur fractlion GXOeTLIMmeNt s i Always lower than sulphate ~ s times. No significant changes are observed for the sulphide and higher oxidation forms. Comparison of these results and those presented in Fig. 1 shows that in the post-irradiation melted samples the radiation damage does not have the same effect as in the unmelted samples. Supplementary information can be cb- tained by studying the effect of high temperature irradiation on the distribution of the radicsulphur. It is possible that in the molten state the active oxidising agents have a different identity from those present below melting. The presence of oxygen and probably sodlum oxides during melting may have a determinant role in deciding the state of the radiosulphur. Effect of irradiation temperature. A compariscn of results obtained by irradiation at 423 K and /7 K (Table 7.2) shows that the higher oxidation fraction is lower (3%) at 77 K. As is seen the increased S° after low tem- perature irradlation occurs at the expense of the sulphate + sulphite and sulphlde fractions. The defects with oxidising and reducing character formed by low tLemperature irradiation such as F and V-centres (or derivatives] become important factors in determining the radiosulphur behaviour. /.2 Temperature dependence of sulphur species (according to Furrer, 19/77) Sipgnificant amounts of the order of magnitude of thousands of opm °°S would be present as steady-state concentration in a proposed fast breeder reactor fuelled with molten Pu/U-chlorides diluted in NaCl. To obtain information about the chemical be- haviour, mainly the distribution of oxidation-states, the in- fluence of irradistion temperatures (-130 and 159C) and the ffects of a post-irradiation heat treatment, solid NaCl was investigated and the results published. The subject of the present note are Investlpations at higher irradiation tempera- tures, especially with samples molten during irradiation. Fig. /0 Effect of length of irradiation time on the 35 -species distribution 100 1 - 2. . ] S0, +5042 ] 5° : 52‘ %, 50 4 = o T T T T T | 2 3 4 5678910 2 3 456789102 2 Irradiagtion time, hr Irradiation-temperature dependence of the oxidation- state distribution of *°S-species sulphur species (%) T T T T T 100 300 500 700 irradiation temperature (°C) Table /.7 Irrad. COHQitions Trrad. Sf_ G0 c0?” + sp?” flux, time temp. (K] % % 4 o 3 5.107 20 em 257! 4723 73.1 + 0.4 9.8 + 0.8 16.9 + 0.8 2 hrs. " 77 Shiod £ 0.8 31.5x 1.1 3.0 = .1 Caperimental Cquimolar mixtures of NaCl and KCI ("Suprapur”, Merck) were studied instead of pure NaCl {m.p.0658 vs B00YC) because the irradiation device permitted temperatures of up to 750°C only. The finely crushed slat-mixture was dried in vacuo at 250°C for 24 hr and subsequently treated with dried FHCl-pas at 300°C for 24 hr in order to remove traces of water and hydroxildes. 100 mg samples were weighed into quartz ampoules in a pglove-box with a purified nitropen atmosphere (07, H,0 T0pom). In order to study the influence of oxygen from the surface of the Si0--ampoules, parts of the samples were weighed into small cruci- hles made of pold-foil and cleosed by folding the foll. The ampoules wore ovacuated to a pressure of less than 137 mm Hg (24 hr) and during evacuation heated to 2Z509C for about 3 hr to remove ad- hering traces of HCl, sealed and irradiated for 2 hr at 500, ©00 or 7000C in the swimming-pool reactor SAFPHIR near the core at a2 total flux of about 4 x 10'% n cm™? s°'. The neutron-spectrum is not well characterised but known to be rather hard. The post-irradiation treatment followed closely the work of (8) as described In detail elsewhere (4.5). The ampoules were crushed at room remperature in a nitrogen atmosphere, volatile reaction products carried by a nitrogen g¢as-stream into cold-traps cooled with liquid nitrogen and the samples subsequently dissolved in an oxypen-froe carrier-solutlon containing cyanide and sulphlde, thiocyanate, sulphite and sulphate as carriers. The fractions were seporated. independently oxidized to sulphate, precipitated as HQS:fl and thn activities measured with a Lhin-window GM-countoer. Some experiments were made at 15000 irradiation-temperature in order to be able to compare the CaCl/KCl-system with pure NaCl. New irradiations of pure NaCl were carried out in order to study the influence of the HCL pre-irradiation treatment. - 183 - The data for at least 2 independent experiments at 150 and 500°C or 4-5 experiments at /00°C for each set of parameters (pre- irradiation treatment, gold-foil packing, irradiation-temperature) are shown in the following table. Sulphite, sulphate and volatile fractions are tabulated together as SX*, While the sulphite- fraction was always less than 2% of the sulphate-content, the volatile part showed large fluctuations, especially with samples packed into gold-foil, (0.20% of the sulphate content) caused by the variations in sample surface and tightness of the package, rven 1f mechanically destroyed before dissolving. The volatile species still remaining In the slat at dissolution, presumably as S5y¢Cly,, are immediately hydrolized to sulphate and only to a small extent to sulphite in the basic cyanide-carrier-solution. The following figure shows the temperature-dependence of the oxldatilon state distribution. The experiments at 1509 on Nall treated with HCl-gas confirmed the existing published results (4.5), which were obtained with a slat dried in vacuo only. The results of the investigations at higher temperatures and with molten samples show a monotone decrease of the S™?-species and a corresponding increase of the SX*-species with increasing irradiation-temperature. The content of 5% is influenced neither by the pre-irradiation treatment not by the irradiation-temperature and 1Is always about Z20%. Melting of the samples during irradiation does not i1nfluence the distribution of axidation states. The studies at 1500C show for NaeCl/KCl-mixtures a shift of about 20% in 5% towards higher oxidation-states, mainly $°%, compared to pure NaCl. No influence of the pre-irradiation sample treatment could be shown at 1500C, but it is of importance for work with molten samples. Untreated samples without gold-foil protection showed S5~°-levels of less than 2%. HCl-treatment increased this value Lo about 13%, an additional pold-foil protection to 26%. Uxygen of the guartz surface in contact with the melt is clearly significant at /00YC. The assumption that oxygen from the guartz surtface should be of importance for reactions cver the cas-phasne i1s not plausible, without mentioning that oxygen-levels in the evacuated ampoules (10°°, Hg) must be much greater but have no significant influence, as low temperature ecxperiments show. vol r X+ Irrad. 5-? g? S 5 - -treatm, 5 - i —~ 5t , salt-type HCl-treatm cld-packing femp. (9C) state (%) (%) (%) NaCl [ No 15C /3 10 17 (4) NaCl Yes Mo 150 /2t 7 13 £ 2 15 £ 2 NaC1l/KC1] Yo No 150 solid 55 + 7 25 £ 2 g x Z NaCl/KC1 Yi2g Yes 500 30 £ Z 19 2 2 51 £ 3 NaCl/KC1 Y253 Ho €00 17 = 3 272 3 61 £ 4 Nall/KC1 No Me 70 2 1 22 * 3 /6t 5 NaC1l/KC1 g Mo 70 Jolten 13 £ 3 27 3 G5 *+ 4 NaCl/KC1 Yes Yes Vasld Zb 4 10 = 3 55 + 4 Table 7.3 O«