INTRODUCTION .... ANALYSIS OF UNCERTAINTIES ... | 2.1 2.2 2.% MSBR 5.1 3.2 5.3 3.4 5.5 3.6 MANPOWER AND COST ESTIMATES .. CONCLUSIONS .... o ACKNOWLEDGMENTS .. CONTENTS Crogs Sections ... Computational Methods i, A . ssumptions Regarding Salt Chemistry REACTOR PHYSICS PROGRAM Investi i igation of Dynamic Characteristics z2.,1.1l OStability Analysis ... %2.1.2 Transient Analiysis ...::jo.'.".'. 3.1.% Flux Flattening ,........:...ae... Investigation of Alternate Core De;;.“..o Development of Methods .. o Cross Section Evaluation :n'..."“...-.. Development of Computer Cod;;‘.n......ai.. Experimental Physics Program .::::“.'D‘. 3.6.1 Dynamics Experiments .. LEGAL NOTICE nt sponscred work. behalf of the Commisgsion: d or implied, with Tes ntained in this rel josed in this Tepo count of Governme peraon acting on sentation, expresse { the information co , ar process disc 8 &N A< This report Was prepared a noT any States, novr the Commission, A. Makes auy warranty or repre racy, completeness, or usefuinese O of any information, apparatus, meth privately swned rights; oT B, Assumes any iaptiities with respect to the uge of any information, apparatus, method, or process Ag used in the above, ‘‘person acting on pehalf o ployee oF contractor of the Commisgion, such employee oT contractor of the Commigsion, digeemingtes, or provides access t, any information purau with the Commisston. t with such contractoT. use of, or for damages I'e disclosed in this report. i the Commission ¢ such contractor, to or employee of ant io his employ or hig employmen Neither the United pect to the accu- port, or that the use ri may not infringe suliing from the » includes any em- the extent that guch contractorl prepares, ment or contract 18 2l 27 2T 28 29 29 3C 30 z1 32 35 35 26 38 T T - PHYSICS PROGRAM FOR MOLTEN-SALT BREEDER REACTORS A. M. Perry 1. INTRODUCTION One of the attractive aspects of the Molten-Salt Breeder Reactor concept that emerges from the design studies conducted at ORNL is the prospect that very low fuel-cycle costs will coincide with very good fuel utilization, that in fact the curve of fuel-cycle cost versus doubling time will possess a minimum at a doubling time as short as 15 to 20 years®, and that this minimum fuel cost will be as low as 0.3-0.4 mills/kwhr(e). Our present estimates of the fuel-cycle cost as a function of annual yield are shown in Fig. 1 for two cases, i.e.,, with and without continuous remcval of 2°°Pa, That & reactor comprising essentially graphite, thorium, and 227U should be able to breed is not in itself surprising, for we have long had reason to believe that this is possible, provided the fuel is re- preocessed at a sufficiently rapid rate. That such rapid processing can be accomplished economically, however, and that a very high fuel specific power can be maintained while keeping neutron losses in #°7Pa to a very low level, appear to be unique properties of the fluid fuel reactor. It must be remembered that the excellent fuel-cycle characteristics projected for the Molten-Salt Breeder Reactor are based on a combination of a low net breeding gain and a high specific power. A net breeding gain of about 0.05-0.06 was found to be optimum (i.e., corresponds to nesr-minimum fuel cost) for the current reference MSBR design. This is of course a very small margin for breeding, and the calcu- lation of it is subject to some uncertainty. In considering the merit *Throughout this report, doubling time is defined in terms of compound interest, i.e., doubling time = 0.693/(annual yield). It thus applies to an expanding system of reactors, rather than to a singie reactor. {Annual yield is, of course, the annual fractional increase in fissile inventory.) ArE ) @ B, % ime { T Deoubling 20 NEY L v A0 i Trom Blanket. 50 &)H’ith@ut 233pg Remova b)fiith 253?& Bemoval, 100 { {o)ayugfaryyu] fiwu & fnnuael Fuel Yiel 3 (8 ield. S Yi £ Fuel~-Cycle Cost Fig, 1. e of the MSBR concept, we must attempt toc appraise realistically the possible magnitude and importance of uncertainties in the calculated characteristics of the reactor, and to consider what steps may be taken to reduce these uncertainties. A description of the Molten-5alt Breeder Reactor concept and of our current reference design for an MSBR is given in the report ORNL- 3996 {(Ref. 1), and will not be repeated here. Some of the important characteristics that are relevant to a discussion of reactor physics problems are given in Tables 1 and 2. {These characteristics are ap- propriate to a single 2225 Mw(t) reactor, operating at an average core power density of 80 kw/litere While they differ slightly from those of a 555 Mw(t) modular core operating at 40O kw/liter, the differences are not material to the present discussion.) Table 1. MSBR Performance Without Ps With Pa Removal Removal Nuclear breeding ratio 1.0538 1.074 Fissile consumption (Inventories per year at 0.8 plant factor) 1.03 1.17 Fissile losses in processing (Inventories per year at 0.8 plant factor) 0.006 0.007 Fuel yield, % per annum I, 06 7.95 Neutron production per fissile absorption; e 2.221 2.227 Specific power, Mw(t)/kg fissile 2.89 3,26 Fuel-cycle cost, mills/kvhr(e) 0.45 0.%% Doubling time, years 14 8.7 “Here defined as 0.693/(anmual yield). 1P, R. Kasten, E. S. Bettis, and R. C. Robertson, Design Studies of 1000-Mw(e) Molten-Salt Breeder Reactors, USAEC Report ORNL-3%996, Oak Ridge Nationasl Laboratory {August 1966). Table 2. MSBR Neutron Balance Absorptions Material Without Pa Removal With Pa Removal 2327 0.9710 0.9970 233Ppg 0.0079% 0.0003% 233y 0.9119 0.92L7 234y 0.0936 0.0819 235y 0.0881 0.075% 236y 0.0115 0.008L 2T Np 0.001k 0.0010 238 0.0009 0.0005 Carrier salt (except ®Li) 0.0623% 0.0648 51,1 0.00%0 0.0025 Graphite 0.0300 0.032% 135¥e 0.0050 0.0050 149 g 0.0069 0.0068 151gm 0.0018 0.00L7 Other fission products 0.0106 0.0185 Delayed neutron losses 0.0050 0.0049 Leakage 0.0012 0.0012 Total 2.2211 2,268 2. ANALYSIS OF UNCERTAINTIES Because of the operating flexibility of fluid fuel reactcrs, which allows criticality to be maintained by adjustment of fuvel concentration, we are not primarily interested in the problem of calculating the criti- cality factor per se. We are concerned instead with the fraction of source neutrons that is available for absorption in the fertile materials. Estimates of this quantity may be uncertain because of uncertainties in cross sections, in methods of computation, or in the assumptions mede regarding the behavior of fission products in the reactor system. These sources of uncertainty are discussed in the following sections. i 2.1 Cross Sections There are comparatively few nuclides in the MSBR for which cross section uncertainties lead to apprecliable uncertainty in estimates of the breeding performance of the reactor; only two or three nuclides have cross section uncertainties that could, alone, affect the breeding ratio by as much as 0.0l. The outstanding example, of course, is the 233U itself. Here the important quantity is the average value of 7, averaged over the entire reactor spectrum. This quantity may be uncertain for at least three reasons: (1) the value of n at 2200 m/sec is uncertain by perhaps +0.3%, (2) the variation of 1 with neutron energy in the range below 0.5 ev is not known well enough to establish 7 (in a thermal neutron spectrum with kT ~0.1 ev) to much better than 1%, and (3) 1 in a 1/E spectrum above 0.5 ev is also subject to an uncertainty of about 1%. The uncertainty in the thermal average value of 7 produces an uncertainty cf about +0.02 in breeding ratio, and appears to be by far the most impcrtant o source of uncertainty in breeding ratio. The ambiguity in the epithermal % is, fortunately, not sc signifi- cant now as it has been until recently. The ambiguity arose from a discrepancy that appeared to exist between average epithermal ¢« values as deduced from differential fission and total cross section measure- ments on the one hand, and from direct integral measurements of @ on the other hand. The differential measurements yield a value of «,% averaged over 3 l/E spectrum above 0.5 ev, of ahout 0.23. This value is subject to appreciable uncertainty, however, because Gé must be deduced by sub- traction of O and O from the measured GT' Furthermore, an adequsate statistical analysis of the probable error in o, as derived from the differential cross sections, has not been made. The integral & measure- ments are performed by measuring the 274U and fission product concen- trations in irradiated ©2°U samples. Results of the three most recent “Based primarily on the measurements of Moore et al. (M. S. Moore, L. G. Miller, and O. D. Simpson, Phys. Rev., 118, 71k (1960). 10 measurenents of this type are as follows: HEalperin a = 0.171L + 0,017 Ref. 3 Esch and Feiner G = 0.175 + 6.008 Ref. L Conway and Gunst T = 0.175 + 0.006 Ref. 5 Averasge T = 0,175 + 0.005 We believe that the close agreement among these independent measure- ments and the inherently greater accuracy of the direct integral « measurement support the lower value of ¢ in the epithermal energy range. The vslue used in the MSBR analyses was & = 0,173, leading to an average value of n, in a l/E spectrum above 0.5 ev of 2.1%. It may be noted that an uncertainty of 0.01 in o (>0.5 ev) generates an uncertainty of about 0.006 in the breeding ratic, for the MSBR reference configuraticn. A similar discrepancy between differentisl cross secticon measure- ments and direct ¢ measurements in the epithermal region has existed for 2357, In recent months the o values deduced by de Saussure, Gwin, and Weston® from their measurements of fission and capture cross sections for =2U are in much closer agreement with the integral « measurements for 22U than any values previously derived from differential cross section measurements, and there is good reason to hope that this trouble- some discrepasncy is very nearly resclved. Similar experiments for ¢ f and ¢ for 2337 are now underway by Weston, Gwin, de Szussure, and their 3J. Halperin et al., The Average Capture/Fission Ratic of 233U for Epithermal Neutrons, Nucl. Sci. Eng., 16(2): 2L5 (June 1963). L. J. Esch and F. Feiner, Survey of Capture and Fission Integrals of Fissile Materials, paper presented at the National Topical Meeting — Reactor Physics in the Resonance and Therral Regions, February 1966, can Diego, California. 5D. E. Conway and S. B. Gunst, FEpithermal Cross Secticns of 233U, Technical Progress Report Reactor Physics and Mathematics for the Period October 1, 1965 to January 1, 1966, USAEC Report WAPD-MRJ-32, p. 9, Bettis Atomic Power Laboratory. ©G. de Saussure et al., Measurement of &, the Ratio of the Neutron Capture Cross Section; for 2357 in the Energy Region from 3.25 ev 1o 1.8 kev, USAEC Report ORNL-3738, Osk Ridge National Laboratory, April 1965, and subsequent private comrunications. | 11 collaborators at RPI.? These measurements, (when combined with other data at energies above 1 kev),’ yield a value for o, averaged over a (1/E) spectrum above 0.5 ev, of 0.188 + 0.0, in much closer agreement with the integral measurements cited above. We believe, therefore, that the range of uncertainty in o has been significantly reduced by these measurements, and can hardly exceed * 0.0l, centered arcund & mean value close to that of the integral measurements. In addition to the related uncertainties in 7 and in @, there is also an uncertainty in the value of p = n(l + o). This is not of any consequence in the subcadmium energy range, since n is a directly measured quantity. In the epicadmium range, however, 7 is deduced from o and y, and must reflect uncertainties in both of these guantities. It is difficult to assess the uncertainty in p because of what appear o be systematic discrepancies between determinations by various methods. Nonetheless,; we presently believe it is unlikely that p lies outside the range 2.50 £ 0.01. The combined effect of the uncertainties in « and in p is an uncertainty of about 1% in fi, in the energy range E > 0.5 ev, Uncertainty in the value of n averaged over the thermal neutron spectrum is important because ~70% of the absorptions in 227U occur in the subcadmium neutron renge. Direct measurements of 7(E)/n(0.025 ev) have been made by several investigators since the early 1950's. The existing measurements are not in good agreement with each cother or with values deduced from differential cross section measurements, nor 4o they have the very high precision required to determine évg to an error as small as that in n_ itself [n = 2(0.025 ev)]. The problem is illustrated by the data shown in Fig. 2, where the symbols represent direct relative n measurements, normalized to Ty = 2.29&*, and the sclid line represents the values used in the MSBR design studies. Averaging over a Maxwellian flux distribution pesked at 0.1 ev, 7L. W. Weston et al., Measurement of the Neutron Fission and Capture Cross Sections for 233U in the Fnergy Region O.4 to 1000 ev, USAEC Report ORNL-TM~-1751, Oak Ridge Naticnal Laboratory, April 1967. * Except for the Harwell (1966) measurements, which are normalized to a value of 2.29 at 0.073 ev. K 0 SRR U ERERE: T + v F e i - ity b ,m._.r‘w.:.“ e - it : L. . L v 1 . ' E Lt LI f N L 3 v ! b . + ._. o [ . P i i ! i- . i . s, . . . Lo P v ! i - b G o4 s vl A . i m.|+: .w L‘ er w”| g e p— e n ‘omk 4 @ . -+ m H n L_ " . * . . . - . . . . . [ » . ¥ ! v ‘" i . M Vo “ o A, i PN “Q C oy . i : [ .,m.;;?:,.w,,,.u.__m S P : : o4 + a e P : i . : T P [ Pl b ol o @ ot b 1 : i ; P e e Mo T EEREREEE i s : : [ Sy Vol Pty b bt s ! AN : .b~fl.‘_“.. O [had bdy i et A + TR poe i B Bib gt e+ i - =t b B O : tmp e b oped s b _ F ~4 ] Sty sy “@ : T fim. i t b vy H ¢ Pt b e - s i . P i -+ P FEE T A e e i L ..:..w...T.H. “eped Lyt s .m_’m 4 ; ,® Vo o : C i I T " T YT YT [ T - 1 i i ] N [P v o i s oo 4 b ¢ 4 m*?., . I Vo .. [ et . - - - i+ Sop o1 s . “ ] s w H s dop b pobe g P - [ P4 | PR O e B T dty s — - 2 bl L . i . T DT o Tend T bbb R sl Y e A SR b bos- : i P b . ; A ¢ _id SRR I 4 .nvw M. Hw* m.m 7. L b 4 gt : B : P b a ey . } 14 Tt 1 P i q4-% 4 S boeod o - I - - 4-4 4 S S B O t Lyt . . [ P b 44 L-1 s b et e ' [ Foy : Pt B +-4- R 4 ook | . + vq P o - 4 e n}.%i.l ’ sl . { : bt Ak - -+ T T T I S i ESREE i G dddty 5 L B Tl : H bl 1 _ : 2 i A m - Fohoby b { i : idld i LE m. m wm , M " 4 « 4 Lopolodg g b Jddda ik @ - : W ;L Pt ! .m t +2 N T b S m 4 L n : e T . ,. TTHTE @ 3 b5 1. LI SE . ; i Cli i Af e w8 S i 4 : + » Er Tt e i T iR m |~ £ o bbbt p b e T £ & = ~ b e b e e -4 g & o Aoy : [ s boroah IR [ ¢ o M o o] P vy m._ srrT b ol T = 4 4t < gt Ao G [ 4o .mrun . - m m m - o1 k- ek ek by s Bl . @ @ -t o4 - 4 + + kb e o ok b - C b 3 Wi M a4 b . g g4 e h ‘ P - & B e f T3 Ea Sl opeeet eep e A 2 -+ + <) T P bRt B oS R e rryrd vt Prrbbbibt i 11t @ o 4 DL e R IRAERRRTRARARERETILLE SESIEEEEI Vet b 4o L3 A b4t b 4 - -+ 2 e O =l —hee d H..w't i g .H. § oAb 4 } b R m 4 off a - bbb bt b b e b fobobod e o > et e Pt T2 d 88 s b e : ‘ i . + b v -4 Hfli.t»..fli b 5 m - Py m 4+l _,;-ww+ i 4.w § . . iohd - m i Pe bbby -k i ! . w;.:m +9 .o i I Poe b C: o e ¢ . ~ T T Pl by ._ bl & o ' : i i ] . & ¢ by _ ; wt.w } +—4 ! ; g - . m W*_ »_« oA © Or 1 ov or o W o N - v s bt b Eorob ot » 44 d-i 0 P : bt i G Peppess AR ERE R RRRE b : Pl Proilia it s de [} O [ i _ um« i Doy 4 : i v Yo b [ i : i . I by 1 ! [ Pt b [ i i . [ o i oo Py l fi —r deopeegod - : i i— : 4 i . i fod -} a by e} [ . Coror i [ w . Lo w o n.fi; H_ & +oped b 4t . I Sy kb teb g e o oio s i . . + AT Tt m b b hadg e 4 Mm LorELRi il - chmbee bog g i - bebea i bobd o ode 4 * Vbbb g fop -+ e g Ebercb e o b e b e gy oot bbb T = —r - - + wfi, - . B m.i@ o b b4 s dob b gniet poib i M mw” I : ,IM‘ N Poe4op %y" ped e fi.e i f-p 4 e ! : +m.". .v..+.4, ...w|4 1 b . »‘v 1 = ‘M 5 Lo .M . - b e %, .- qL H M., b Ap A,T.T - _ -} Rt ST T . - m_ oo okog i . b e L. bt baRE dod e b e e b4k ,.MLI ; L b bbb e e g T 4 j + 15 B 2.4 —_ 2. 2 o 2 uotidrosge /sucainau) b r=f ° & 2.0 Neutron Fnergy (ev) Relative Eta of 2337, Fig. 2. 13 one can easily obtain wvalues for H'ranging from 2.26 to 2.30 and the true value could possibly; though not probably, lie outside this range. This uncertainty in the average thermal n of 222U remains the most impeortant single contributor to uncertainty in the breeding ratic of an MSBR. The Ué and Gf measurements of Weston et al. are now being ex- tended downward in energy to about 0.02 ev, and it is expected that this will significantly reduce the uncertainty in the average value of 1. One of the most abundant materials in the MSBR, and one of the most important parasitic neutron-absorbers, is fluorine. As is true of other light elements, the resonances of fluorine are predominantly scattering resonances, and the radiative capture widths are difficult to determine accurately. The capture widths are not know to better than +3%0%, and the high-energy {n,a) cross sections are equally uncertain. These un- certainties affect the estimated breeding gain to the extent of about C.005; while not large in an absolute sense, this is a non-trivial fraction of the breeding gain, and it would facilitate further design and coptimization of molten-salt reactors to have improved accuracy in these cross sections of fluorine. A more accurate determination of the resonance capture integral would itself be an appreciable help in re- ducing the limits of uncertainty in the fluorine absorption rate. Uncertainties in remaining cross sections, including Li, Be, C, Pa, and fission products, prcbably do not contribute an uncertainty in breeding ratio greater than about 0.01. The effective cross sections of thorium may indeed bhe subject to considerable uncertainty, arising from uncertainties in resonance pa- rameters, from methods of computaticon of resonance self-shielding, and from variations in geometry of the fertile salt passages. Variations in passage geometry may well contribute the greatest uncertainty in thorium abscrption rate. Further analysis of this possibility is re- quired, but is likely to lead to requirements orn the mechanical design cf MSBR cores, rather than to the need for further measurements of cross sections or resonance integrals. Uncertainties in cross sections of 23411 and 236y are of minor consequence, since these meterials reach equilibrium concentrations rather quickly. The 224U is a fertile 1y material, while =°°U is a poison. The eguilibrium absorption rate in each depends primarily on the capture-to-fission ratio of the fissile precursors, = -U and #7°U; however, there is some small dependence on the 224U and 27U cross sections because some of the material is ex- tracted from the fuel stream, along with the fissile isotopes, as excess production. The 275U cross sections are known with about the same precision as those of 237U, but are of far less importance, since less than 10% of the fissile-material absorptions are in 2251, The various cross secticn uncertainties that contribute signifi- cantly to uncertainty in the estimated breeding performance are sum- marized in Teble 3. In this table, we show ncminal ranges of uncertainty as fractional deviations from what we believe to be the most probable values. We refrain from calling these deviations prcbable errors,; be- cause in many cases they do not represent standard deviations of a normal error distribution, and hence do not really represent confidence limits in a conventiocnal statistical sense; they do represent our present - Judgment of the ranges within which the true values hsave perhaps a 50% or greater probability of falling. Also shown are the corresponding un- certainties in breeding gain. In the case of 227U, 274U, and 22°U, the consequent changes in £°°U/223U absorption ratio and in Z°%U absorption rate are taken intc account in the indicated uncertainties in breeding gain. Since the uncertainties listed in Table 3 are sll independent; and, with resgpect to the most probable values of the various cross sections, positive and negative deviastions are equally likely, we have combined then by teking the sguare root of the sum of the squares as the overall uncertainty in breeding ratic attributable to cross section uncertainties. The resulting value, (Z&?)l/g = 0.026, reflects primarily the uncertainty in the average thermal 7 of 237U, The effect of cross section uncertainties can also be appreciated by reference to Fig. 3. The various curves of fuel-cycle cost versus annual fuel yield that are shown in Fig. 3 represent the result of re- optimizing the reactor to compensate for specified alterations in cross 15 Table 3. Effect of Cross Section Uncertainties on Breeding Ratio Nuclide Cross Sectiona 6&b AC SBRd 233y 1, 0.00% | £0.007 (7 (t)/n,) 0.01 +0.022 1 (f) 0.01 +0.,009 2357 n (%) 0.01 +0.00% 7 (f) 0.013 +0.001 23477 o, {(t) 0.1 0.033% - g (f) 0.25 0.049 +<0,001 236y; a (%) 0.1 - -- o (f) 0.% 0.008 +<0,001 233pg o, (%) 0.1 - o, (£} 0.1 0,0003 -- 1op T, (t) 0.07 0.01% +0.001 o, () 0.3 0.008 +0.003 o(n,a) () 0.% 0.006 +0.002 Tri o (t) 0.1 0.02 +0.002 o (f) 0.1 0.001 -- %Be o (t) 0.1 0.002 - GP,En(§§) g'i5 20229 }- %0.002 0,0 . .003% F.P. o, (t) 0. 0.0L +0.001 o (f) 0. 0.01 +0.00% ®The notation (t) signifies the energy range below 1.86 ev, and the notation (f) signifies energies above 1.86 ev, except for “~°U and #°°U where the break point is 0.5 ev. bfia is the fractional uncertainty in the cross section. cApproximate typical sbsorptions, relative to ne source neutrons; may vary, of course, from case to case. dUncertainty in breeding ratio resulting from indicated cross section uncertainty. S FUEL CYCLE COST [mills/kwhie] Q7 os / N\ fffifi S — ] 04 ] — e / 03 ) 0.2 1) REFERENCE CROSS SECTIONS - WiTH Pa REMOVAL 2) REFERENCE CROSS SECTIONS - WITHOUT Pa REMOVAL o 3) INCREASED FLUORINE ABSORPTION 4) ALL, ABSORPTION CROSS SECTIONS INCREASED EXCEPT us ? o 5} ALL CAPTURE CROSS SECTIONS INCREASED 10% 5 -4 -3 2 -4 0 { 2 3 4 5 6 T 8 9 ANNUAL FUEL YIELD (% /yr) Fig. 3. The Effect cf Cross Section Uncertainties on the MSBR Performance. : 10 L7 s section values used in the calculations. Curves 1 and 2, which also appear in Fig. 1, are the reference curves with and without 235pa re- removal, respectively. Curve 3 results from increasing just the fluorine absorption cross sections, for the case without Pa removal, while curve 4 results from increasing the absorption cross sections of all con- stituents of the core (except the heavy elements Pa, U, and Th) by the percentage amounts shown in Teble L. Table 4, Assumed Increases in Capture Cross Sections (Percent of reference values) Isotope Thermal aé Epithermal Ué SLi 1 10 o Ld 11 10 Be 11 i5 - c 9 S ¥ T 32 149q, 10 20 151gy, 10 20 Other fission products 10 10 To obtain curve 5, capture cross sections of all nuclides, in- cluding the heavy elements, were increased by 10% at all neutron energies. By far the largest effect of this perturbation is a de- crease of about 0.03 in the average value of 7. All of the perturbations represented by curves 3, 4, and 5 are re- lative to curve 2, i.e., without Pa removal. Comparison of these with curve 1 shows the very substantial incentive for continucus removal of the Pa. (All of the perturbations shown are in the unfavorable di- rection, representing an adverse resolution of all cross section un- certainties. Deviations in the other direction are of course equally likely, so far as cross section uncertainties are concerned., ) 18 In summary, the cross sections which particularly require further e investigation are: 1) the variation of 7 of 27U with neutron energy in the range of 0.0l to 1 evy 2) the absolute values of 7 and p at 0.025 ev; 3} the radiative capture width, the (n,Q) cross section, and the rescnance capture integral of 1°F, Further analysis of data already available may either reduce the uncertainties assigned to some important quantities, such as the average epithermal «Q, or may pinpcint specific measurements which would be especially helpful in reducing such uncertainties. 2.2 Computational Methods Verification of computational methods, without ambiguity from cross section uncertainties;, is usually difficult to obtain. However, cur experience with the MSRE leads us to believe that on the whole our methods are quite adequate to deal with this type of reactor. Briefly, : the methods employed in the statics calculations were one~ and two- dimengional multigroup diffusicon theory. The neutron spectrum and group- averaged cross sections were obtained from GAM-THERMOS cell calculations. A compariscon of predicted and subsequently observed values for some of the important charscteristics of the MSRE is given in Table 5. The good agreement betweén predicted and observed values lends con- siderable confidence in the wvalidity of the methods employed. OSimilar methods were used by General Atomic in the prediction of critical loadings for the Peach Bottom Reactor, which is complicated by nonuniform cistri- butions of fertile msterial end poisons,; by singularities such as control rods and poisoned dumny fuel elements, and by appreciable self shielding of the heterogenecusly distributed thorium. Observed reactivities were nonetheless within 0.005 of predicted values, and since this agreement prevailed over & range of core loadings, the possibility of chance cancel- laticon of systematic errors is considerably reduced. It mast be acknowledged, however, that the MSBER configuration is somewhat more complicated than that of the MSRE, and has complexities of s 19 : Table 5. A Comparison of Predicted and ‘ Observed Characteristics of the MSRE Charscteristic Predicted Observed Critical concentration of 275U, g/liter 33,06 33,1 Fuel concentration coefficient of reactivity, gi fi 0,234 0.223 Isothermal temperature coef- ficient of reactivity, &k/k/°F ~8.1 x 10" 7.3 x 10-% Reactivity worth of three control rods, % 8k/k 5 o 46 5.59 Reactivity effect of fuel circulation (loss of delayed neutrens), % Sk/k 0,222 0.21 a somewhat different character from those of the Peach Bottom Reactor. A sketch of the present concept for an MSBR lattice cell is shown in Fig. 4, from which one may appreciate the importance of a careful calcu- lation of the space- and energy-dependence of the neutron flux, both for thermal neutrons and for resonance neutrons. While estimates of the potential performance of the MSBR concept are not seriously affected by errors of a few percent in calculating these details of the flux distri- butions, the design calculations for a particular reactor require higher precision, primarily fio provide assurance against fuel cost penalties that might arise if the critical fuel concentration were appreciably different than expected. Although we have no a priori reason to doubt the adequacy of presently available methods, it will be necessary to verify their adequacy both by investigating the effect of further re- finements in technique (cf. Sec. 3.3) and by comparisons between calcu- lations and the results of carefully selected experiments which reproduce the details of the MSBR cell geometry (cf. Sec. 3.6). 20 MODERATOR GRAPHITE) ——— . FUEL PASSAGE (UP}——— —— ST s FUEL PASSAGE (DOWN) 34 0D FUEL TUBE | MOGERATOR HOLD DOWN NUT GRAPHITE) AR W et ' . —SPACER e} ——METAL TO GRAPHITE e SLIP~JOINT i T ————METAL TO GRAPHITE E' BRAZED JOINT i |~ —BRAZED JOINT 4 T = "'M \-r | Ei«——i‘-'ua_ INLET - k) e b PLENUM - S AW . . =—-—-FUEL OUTLET e PLENUM ] _ — . ! + Fig. L. Molten-Salt Breeder Reactor Core Cell. 21 S 2.3 Assumptions Regarding Salt Chemistry As is well known, the conversion ratio in a thermal-neutron reactor depends very much on the rate of processing of the fuel, largely because it is by this means that neutron losses to fission products may be con- trolled. In the processing scheme proposed for the fuel salt of the MSBR, the thirty cr more chemical elements of which significant amounts are present in the fission products may be expected to behave in quite different ways, depending on their chemical and physical properties in a very complex environment. The assumptions that were made regarding fission product behavicr in the MSBR studies are cited in Teble 6. (For a description of the processing system, see Ref. 1.) Table 6. Disposition of Fission Products in MSBR Reactor and Processing System l. Elements present as gases; assumed to be partly absorbed by graphite and partly removed by gas stripping (1/2% poisoning assumed): Kr, Xe 2. FElements that plate out on metal surfaces; assumed to be removed instantaneocusly: Rh, Pd, Ag, In 3. Halogens and elements that form volatile fluorides; assumed to be removed in the Se, Br; I, Nb, filuoride volatility process: Mo, Ru, Te, Te I, FElements that form stable fluorides less Sr, Y, Ba, La, Ce, volatile than LiF; assumed to be separated Pr, N4, Pm, Sm, Eu, by vacuum distillation: Gd, Tb 5. Elements that are not separated from the carrier salt; assumed to be removed only by salt discard: Rb, Cd, Sn, Cs, Zr Tn most instances, (except perhaps for groups 2 and 3) we still be- lieve these to be the most probable medes of behavior. It must be acknowledged, however, that these assignments are not in all cases certain, and one must ascertain the effect on MSBR performance of possible, if imprcbable, deviations from these assumptions. ............... X N o Because of their combination of high fission yield and high neutron- absorption cross section, and because their fluorides are probably not more stable than their carbides; one is particularly led to examine alter- native modes of behavicr for the elements of group 3, especially molybde- num and technetium. It is entirely possible, even probable, that these elements will form neither fluorides nor carbides, but will rather form inter-metallic compounds with cther metallic fission products, e.g., those of group 2, or simply remain in the salt as colloidal suspensions of the metal., In this event, these elements would still be removed in the vacuum distillation process, and there would be no change in the neutron balance. There remains the possibility that some fraction of these group 3 fission products might react with the graphite moderator, forming metal carbides, and hence remain indefinitely in the core. Deposition of several fission procducts, including Mo, Nb, Ru, and Te, has in fact been observed on graphite specimens in contact with the fuel salt in the MSRE. If one assumes that these samples are typical of all the MSRE graphite, one can calculate the fraction of each fission product species that remains in the core. These fractions, calculated from activities observed on the graphite specimens removed from the MSRE in July 19666, are shown in Table 7. It is immediately obvious, of course, that any mechanism that can leave fission preducts in the core indefinitely is potentially very serious, especilally so in a reactor with very high specific power. It can easily be shown that the additional neutron absorption that would result would be nearly proportional to the fraction, f, of the atoms in this group that remain in core, instead of being removed in processing. The time required for each species to saturate depends, of course, on its cross section. The poisoning effect of each of several fission product auclides that would result from 100% retention on the graphite of an MSBR is shown in Table 8 as a function of time, in full-power years, after startup of the regctor, As an application of the information given in Table &, Table 9 shows the average poisconing that would result in an MSBR if the various nuclides were deposited to the extent cobserved in the MSRE {(as shown in Table 7). (Two different assumptions were made regarding the behavior of 95Mo, that is, that it behaves either like its precursor, “°Nb, or like 23 Table 7. Fission-Product Deposition in the Surface Layersa of MSRE Graphite (Percent of Totalb) Graphite Location Isotope Top of Middle of Bottom of Core Core Core 9SMo 13,4 17.2 1l.5 132me 13.8 13.6 12,0 103wy, 11,k 10.3 6.3 9SNp 12 59.2 62.4 1317 0.16 C.33 0.25 9574 0.33 0.27 0.15 1440, 0,052 0.27 0.1k 895y 2,2k 3.30 2.7k 140p, 1.38 1.85 1.1k 1410, 0,19 0.6% 0.36 1370 0.07 0.25 0.212 aAverage of values in 7 to 10 mil cuts from each of three exposed graphite faces. bExpressed as percentage of the guantity of each species produced in the reactor that would be deposited on graphite if each cm® of the 2 X 10° cm® of moderator had the same concentration as the specimen. ol Table 8. Ioss of Breeding Retic Corresponding to Complete Retention of Certain Fission Products in the MSER Core Time After Startup (full-power years) Nuclide (%ii;l 2 5 10 15 20 2 9Mo 5.4 0.0167 0.0323 0.0453 0.0507 0.0528 ?7Mo 36.2 0.0026 0.0062 0.0115 0.0163 0.0201 8o 116 0.0008 0.0020 0.0038 0.0055 0.0073 100yq 118 0.0007 0.0017 0.0033 0.0049 0.0065 ?297¢ 3.9 0.0174 0.0312 0.0399 0.0425 0.0434 +01Ry 9.1 0.0055 0.0118 0.0184 0.0222 0.0244 102gy 53 0.0008 0.0020 0.0036 0.0051 0.0066 104py 82 0.0002 0.0005 0.0010 0.0015 0.0019 103Rn 0.51 0.0166 0.0169 0.0169 0.0169 0.0169 105 7.5 0.0012 0.0024 0.0035 0.0041 0.0045 107pg 11.4 0.0002 0.0004 0.0006 0.0007 0.0008 i 126mg 58 0.0001 0.0002 0.0004 0.0006 £.0008 128 290 —- 0.0001 0.0003 0.0004 0.0006 13 0mg, 193 0.0002 0.0006 0.0012 0.0018 0.0023 Total 0.0631 (.1083 0.1494 0.1732 0.1889 Mo, 98Mo, 0.0041 0.0099 0.0186 0.0267 0.0339 T0%40 101Ry, 10%Ry, 0.0065 0.0143 0.0230 0.0288 0.0329 104Ru 126me 128pe 0.0003 0.000¢ 0.0019 0.0028 0.0037 130Te P 0.035 C.067 0.099 C.120 0.136 25 Table 9. Average Polsoning as a Function of Exposure with Deposition Fractions from First MSRE Samples - 1 t P =z j; P4 )at! Time (years) Assumption | 2 5 10 15 20 95Mo acts like ®°Nb 0.0072 0.0151 0.0229 0.0278 0.0%20 FSMo acts like Mo 0.004% 0,008L 0.0121 0.0147 0.0166 99Mo.) Table 8 also liste the combined contributions of several groupings of isotopes and the totals for all the isotopes listed. The pcisoning, P(t), shown in Table O represents the current loss of breeding ratio at time t after startup with clean graphite; also given in Table 8 is the average loss in breeding ratic, defined by P = (1/t) j‘t P{t’ )at’. The noble metals (group 2 in Teble 6) constitute Snother group of fission products whose behavior may well be different from that assumed in the MSBR studies. Since about two tons of these materials (mostly ruthenium) will be produced by one 1000-Mw(e) reactor over a 30-year period, one would prefer that they not deposit on metal surfaces, as was assumed to occcur almost instantaneously. The alternative, and more likely, possibility seems to be that they will react with other fission products (evg.; molybdenum), forming intermetallic compounds, or remain in ele- mental form, and in either event be removed in the residue of the vacuum distillation process. A calculation of the additional poisoning that would result from having these nuclides remain in the fuel stream for the nermal processing cycle indicates a maximum loss of breeding ratio of 0.001, which is certainly nothing to worry about. If, for any reason, all of these nuclides were to remain in the core indefinitely, the asymptotic poisoning effect would be about 0.08. This would of course be serious, but the probability of its occurrence seems vanishingly small, 26 The behevior of xenon {and krypton) in an MSBR system is, of course, & very importent, with some 0.0h in breeding ratio dependent on nearly complete removal of these gases by sparging with helium in the fuel pump. Experience with operation of the MSRE gives every assurance that this can in fact be done. The residual xenon poiscning in the MSRE appears to be appreciably less than anticipated on the hasis of the known permesbility of the graphite, an observation which may be accounted for by some slight entrainment of small helium bubbles in the circulating fuel salt. The assumption with respect to group 5 fission procucts is that they remain in the fuel salt essentially indefinitely. It is perhaps at least as likely that cadmium and tin will behave like group 2, that is, as Jjust discussed, be removed in the regular fuel processing cycle. Such a con- tingency could only improve the breeding ratio. However, the combined yvield of all the fission product chains from mass number 111 to mass number 124 is only about 0.3%%, so that at most 2 gain in breeding ratio of 0.00% might be realized. The reasons for the fission product behavior observed in the MSRE o are not yet fully understood. The role of varicus factors which may influence this behavior, and the mest promising means of limiting the deposition of fission products will be thoroughly investigated in a research program described in another report.8 The subject is introduced here because the behavior of fission products constitutes the principal source of uncertainty in the expected nuclear performance of an MSBR. An additicnesl assumption of some consequence, not listed in Table 6, is that the 237Np formed by neutron capture in 2°°U will be removed from the fuel stream by the fluoride volatility process., If this were not the case, and the 2°"Np were to remain in the fuel stream, along with the uranium, there would be a loss of ~0.01 in breeding ratio. We be- lieve that The neptunium can in fact be removed, by proper operation of the sorbers in the flucride volatility process, and the potential loss in breeding ratio Jjust cited indicates that there is good reason tc do s0. &4. R. Grimes, Chemical Research and Development for Molten-Salt Breeder Reactors, USAEC Report ORNL-TM-1853, Osk Ridge National Laboratory, June 1957, id o 3. MSBR REACTOR PHYSICS PROGRAM As a result of the analyses summarized in the preceding sectlions; we are quite confident that an MSBR will breed under conditions that produce minipum or near-minimum fuel costs. There are nonetheless a number of aspects of the physics of MSBR reactors which require further investigation, both to establish an adeguate basis for the detailed design of an MSBR and to gain a much hetter understanding of the dynamic characteristics of these reactors. 3.1 Investigation of Dynamic Characteristics The design studies of molten-salt breeder reactors that have been carried out up to the present have emphasized the normal,; steady-state behavior of such reactors, in order to determine their potential per- formance with respect to the goals of rescurce utilization and econocmic power. Less attention has been directed to such questions as the dy- namic response characteristics of an MSBR, as influenced in detail by the design parameters, and to possible sbnormal modes of behavior that might result from fallures anywhere in the systen. In order tc take full advantage of its breeding potentisl, the MSBR design must minimize neutron losses to control rods and associated hard- ware (such as thimbles)., 'This implies that it is highly desirable for the MSBR to be strongly self-regulating. While there are no reasons to suspect unsatisfactory dynamic be=- havior in the MSBR, the system has new features whose effect on dynamics cannot be predicted quantitatively on the basis of past experience. For instance, the system will use cireculating 237U fuel, and the small de- layed neutron fraction of 237U will be reduced to an even smaller ef- fective value by fuel circulation. Also, the system is a heterogeneous, two-flunid, circulating fuel reactor and conseguently has slmost every time delay conceivable in & reactor system (heat transfer from graphite toc fuel, fuel transport in the core, blanket transport in the core, etc.). The negative temperature coefficients of reactivity which are to be de- S signed into the system are no guarantee of stability unless the time lags are suitable, 28 The experience acquired with the MSRE provides understending ebout this type of system which will aid in analyzing the MSBR. The pre- dictions of MSRE dynemic behavior® ‘were experimentally coni‘irmed,lo in- dicating that satisfectory mathematical models and analysis procedures ~ were used. Experience with the proposed 233U loading in the MSRE will further extend our understanding. | g 3.1,1 Stability Analysis Anelysis of the dynamic behavior of the MSRE“was based on calcu- lations of the eigenvalues of the time-dependent equations for the neutron density, on analysis of the system frequency response (transfer functions) and on;computation_of,the transient response ‘to various perturbations in system'operating'parameters. - These methods must be epplied to clerify the complex relationship existing between the dy- namic behavior of the MSBR system -and the design parameters. The anal- ysis must of course include calculation,o: all temperature- and'power- ‘dependent reactivity effeets.~:An“investigation'of_the-effects-Of'long- term dimensional changes in;thefgraphite‘structures;(resulting from fast neutron bombardment), and of tolerances or indeterminacy in the geometry of the salt passages will'befreduired The possibility of oscillations or other instabilities -assoclated. with" mDVement of graphite structures, end concomlitent changes in saltfpassage geometry, although thought to be remote, must be investigated. ' Lome - . ‘ Drawing upon these studies, and the transient analyses described below,a conceptual control.and safety system.must be developed which involves the smallest possible steady-state loss of neutrons to elements of the control system, while providing ample operational flexibility andf protection.;._- 95 J. Ball end T W Kerlin, Stability Analysis of the Molten-Salt; Reactor Experiment, USAEC Report OFNL TM-lO'(O, _Oa.k Ridge National Laboratory, December 1965 - 19R. B. Briggs, Molten—Salt Reactor Program Semiannual Progress R Report for Period Ending February 28, 1966, USAEC Report 0RNL-'5956, Oak Ridge National Laboratory, June 1966 b 29 @ - A program of experimental investigations must be developed for the breeder reactor experiment in order to provide additicnal verification of the models and physical properties employed in the analysis for the MSBR configuration. Extensive pre-analysis of the experiments, to facilitate selection of the best experimental conditions, will greatly enhance the value of the experiments themselves. 3.1.2 Transient Analysis Because of the mathematical methods used; the dynamic analyses discussed above deal primarily with the effect of comparatively small disturbances in the reactor system, and are therefore principally applicable to normal operating conditione of the reactor. lLarger dis- turbances can of course arise from abrupt changes in load, from pump stoppages, or from any of a number of other rapid changes in operating conditions. The effects of such changes must be analyzed to determine whether system temperatures will inherently remain within acceptable limits or whether, on the contrary, specific ccntrol actions must be taken. Additional studies will be required in connection with the | safety analysis of the MSBR. All possible sources of positive re- activity addition must be identified and evaluated, including those wvhich might result from failures outside the nuclear sysitem proper, and could therefore be regarded as secondary criticality accidents. The methods presently available for studying nuclear excursions in an MSBR must be carefully examined; some extensions and improvements in these methods may well be re@uiredg particularly with regard to the transient temperature distribution within the core and the transient distribution of delayed neutron precursors. 3.1.3 Flux Flattening The length of time during which the graphite structures in an MSBR can continue to perform their function depends partly on the fast neutron flux level {i.e., on power density) and partly on the gradient of the power density, as well as on the nature of the graphite itself. The use- S ful life of the graphite may be extended somewhat by flattening the power distribution, a&s for example by varying the size of salt passages from place to place within the core. Such variations could also influence the reactivity coefficlients associated with these salt passages. Both the desirability of flux flattening and the effect of doing so on reactivity coefficients should be investigated. 3.2 Investigation of Alternate Core Designs While it is unlikely that there 1s any configuration for an MSBR that would have significantly better breeding performance at low cost than does the present reference design, there may be aliernate core configurations that could yield essentially the same performence while possessing different, and perhaps desirable, mechanical features. A search for such alternatives should be carried forward to provide addi- tional assurance that the prototype reactor design will represent the best basic core concept. - 2.5 Development of Methods : Further improvement and refinement of computational methods is needed in order to establish a satisfactory level of confidence in the procedures — whether most elaborate or relatively simple — that will be used in design of a specific MBBR, such as the 150-Mw reactor experiment, and in order to provide for precise interpretation of related lattice physics experiments (cf. Sec. 3.6). As is usual in geometrically complex reactor lattices, the key problems relate to the calculation of ¢(£,E), the neutron flux as & non-separable function of position and energy, in the source-energy region, in the resonancé region and in the thermalization range. Problems of this sort are present in many types of reactor lattices, and cannot be said to have been fully resolved. The special features of the MSBR lattice relate to the physical separation of the fissile and fer- tile materials in separate salt streams, to the gecmetrical irregularities of salt passages, and to the significant scattering contribution of the fuel salt itself. Both two-dimensional multigroup neutron-transport methods and Monte Carlo methods should be tested, and one or both approaches used 31 e in the analysis of lattice experiments to determine the amount of detail in the description of ¢(§,E} that it is necessary to obtain in order to account for all important characteristics of the MSBR lattice. In the same vein, and in view of the dominant importence of calcu- lating correctly the spectrum-averaged capture-to-fission ratio for 233y, it is highly desirable to develop suitable procedures for calcu- lating Doppler-broadened; self-shielded cross sections for the fissile materials without assuming asymptotic flux shapes above each resonance, and, of course, to do this in a complex heterogenecus lattice. It is not likely, in fact, that any really large effects,; in an MSBR; are associated with the details of the flux distributvions implied by such refinements of analysis. However, the objective of achieving an un- usually high degree of reliability in the design calculations in order to guarantee the performance of the reactor within very narrow limits requires both meticulous attention to detail in the calculations, and supporting experimental work (Sec. 3.6). Because of the relatively small size of an MSBR core, which results from its high power density, and because of the continuous removal of xenon from the fuel salt, as well as the thorough mixing that would occur even if xenon were present in the salt, there will be no tendency towsrds flux instabilities of the kind normally expected in large power reactors. The question of non-separable time- and space-dependent effects will nevertheless arise in connection with the analysis of poctential asccidents. Further investigations will be regquired to determine what extensicns in computational technique may be needed to describe the reactor adegquately for such transient analyses, and, depending on the outcome of these in- vestigations, additional work mey be necessary to accomplish the indi- cated development of methods. 3. Cross Section Evaluation There is a constantly accelerating rate of acquisition of new experi- mental information on neutron cross sections of interest in reactor calculations. OSuch information must be collected, evaluated, and assimi- G lated into our computational structure. Many of the cross sections 32 ‘discussed in Sec. 2.1, while not individually contributing major un- - certainties in the nucleer-celculetione for en MSER, need'furtherrenely- sis end evaluation to ensure that best values are employed in our enaly- ses and that uncertainties and sources of error are more quantitatively assessed than has yet been done, - ‘ The aesimiletion of new information on 233U CYOSS eections, espe- cielly, requires eignificent_effort,'in determining-thezreeonence pa~ rameters. fihet best fit the experimentel date, in deriving“etefieticel' distributions of these parameters for use in the unresolved resonance region (1nc1uding proper allowance for resonances not identified in the differential cross section measurements), and in expreeeing the resulting information in terms best suited for reactor computations, _Sohe'of'this wvork 1is oustomerily and epproprieteiy performedby the experimentere .themselves, notably,the fitting'of'perameters“to-the resolved resonencee; but the reactor physicist etill;hee much.toido, eepeciellylif the desired representation of the cross eections for'fhe'purp0se-of'reactor ceicu- latione is not in terms of the conventional. peremetere. - In addition to enelyeie, evaluation, and, in some 1nstances, the - theoretical celculation of needed cross eection date, the maintenance of en up-togeete,croee section library is e‘reguler housekeeping chore that each mejor reector project must ecknowledge and support. - 3, 5 Development of Computer Codes - “In. eupport of the MSBR deeign studiee, which culmineted in the ref— erence design described 1in- Ref. 1, & procedure wae devieed for finding eutometieelly the optimum combinetion of es meny as. twenty varieble pe—‘ remeters of the reactor eyetem, such ee core eize and height-to-diemeter retio, fuel- and fertile-stream volume frections, thorium end urenium | | concentretione in the eelt, blenket thicknese, proceeeing retee, fertile - salt hold-up volume, end others. Celled OPTIMERc;llthe progrem usee | 11}1 F. Bauman end J. L. Lucius, oprmeac ‘A Reactor Design _thimizetion Code, Osk Ridge Netional Leboratory (to be issued). # 33 one-dimensional multigroup diffusion theory, alternating between radial and axial directions in the core to synthesize a two-dimensional model, and generates space- and_energyfintegrated reaction rates_for each type "~ of nuclide. The isotope‘chain'equations areosolved to‘find the equilib- rium fuel concentrations corresponding with a specified processing rate. .'Solutions of the diffusion equations ‘and of the 1sotope equations are interleaved in & convergent iterative‘procedure which is better described in Ref*g . The program systematically'Searches (by a method of~Steepest gradient) for that combination of variables that gives the optimum value for a selected figure of merit, such as. 1owest power cost. This code ' has proven to be extremely useful in arriving at an optimum.core design. It still hes some restrictions whose removal will meke the tool still more. useful and convenient in evaluating proposed alternative core con- o cepts and possibly in exploring the changes in design and ‘operating conditions that might result from changing conditions in the nuclear o power industry, such as increases in the cost of fissile material. These improvements will require a fairly modest effort, and should be underteken. -In connection with the maintenance of & master-cross-section library, 'from‘vhich datea can'be retrieved and processed for various specificncom- putational needs, data-handling procedures need to be improved and some additional codes developed to facilitate full and relisble use of the library. Many of the computer codes that will be used in further analyses of the MSER reactor need to be transcribed for the latest generation of digital computers, and in some instances altered and improved'to‘take' full adventege of computer capability. R 3.6 Experimental Physics Program As was discussed in Sec. 2.2, the general approaches employed in the MSBR studies have proven quite effective in analysis of the MSRE, ~ the Peach Bottom Reactor, and others. However, the validity of these approaches, or of the improvements discussed in Sec. 3. 5, as applied to ~ the complex lattice geometry of an MSBR, should be confirmed by a few well-selected and carefully executed experiments on the characteristics 3 - of an MSBR lattice. The most appropriate type of experiment. to 111 'this need appears to be the kind of lattice substitution measurement, and associated flux and activation measurements, that- can be made in the - Physical Constants Test Reactor (PCTR) and the High- Temperature ,Lattice Test Reactor (HTLTR) at the Pacific Northwest Laboratories.» - Extremely accurate determinations of lattice reactivity can be made with & small number of typical lattice cells, requiring far less materiel and fabrication cost than would be needed for exponential~or critical , experiments. For lattices with k, close to unity, and with & precision - of perhaps 5% in determining (k -l), one may expect to determine Keo for the lettice to. within about +0.001, or possibly better.=w ' - A measurement of k,, does not by itself, of course, provide an un- ‘ambiguous determination of breeding ratio. A nearly direct: measurement of this important quantity can be obtained by measuring the ratio of absorptions in thorium to fissions in 233y, i.e., (Aoa/Fé3) -~ In natural or slightly enriched uranium systems, the analogous ratio, (Aee/Fbs): cen be measured to within about 1%, or possibly a little better, if extreme care 1s taken, Far less experience has been accumulated with the thorium-2>3y system.(which, of course, involves: different. character-' istic decay gamma rays), and it is not quite. cleer hOW'high a precision cen be achieved in this measurement. Further investigation of this question will be required, and some development work may be needed, before. we can determine Just how mich information cen be obtained, and _‘with vhat precision. \It appears nevertheless that & program ‘of such " lattice. measurements on the PC'i‘R or the HTLTR, including determinations of reactivity, flux distributions, and - activaticn ratios ‘can gc far to provide the detailed understanding of the lattice characteristics that 7will e required for the design of an MBBR ' ' _ In connection with PCTR and HTLTR experiments, it is both possible - and desirable to obtain additional information related to various re- . A_ activity coefficients for the 1attice under study.r Temperature coef- ,;v_ | ficients, density coefficients, effects of displacement of various com-:g ponents of the 1attice cell,can all.be measured with high accuracy if the experiment is appropriately designed with these measurements in Ll 52 mind. In additicn, there will be a velocity selector available at the HTLTR, with which one can undertake measurements of the low-energy neutron spectrum as a function of position in the lattice cell. By performing some of these measurements (e.g., reactivity and activation ratios) on various lattice configurations, some of which may not be typical of an MSBR per se, but which are chosen to emphasize one or snother particular aspect of the neutron balance, one may gain further understanding of the detailed behavicr of the neutrons in an MSBR lattice. Questions of exact experimental design, such as use of frozen salt as opposed to molten salt, the method of containing the salt, and so forth, have not been explored. Some of the lattice cells -~ perhaps as few as seven — should contain primarily 77U as fissile material. For this purpose, not more than a kilogram or two of 27U should be required. Further work is needed to develop a detailed experimental progfam along these lines, and to determine how many separate lattices should be investigated. In order tc estimate the scope of the effort required in these experiments, we assume that not more than five lattices would be studied, and that three of these would be studied in the PCTR, and two in the HTLTR. 3.6.1 Dynamics Experiments While the lattice studies in the PCTR and HTLTR can provide some information on reactivity coefficients, they cannot,of course, tell us anything about the overall dynamic behavior of an MSBR. ©Such studies will have to be carried cut on the reactor experiment. A detailed program for these experiments must be planned in advance, in order to ensure that adequate provision is made for them in the design cf the reactor. The experiments will include measurements of frequency re- sponse and transient response to various perturbations in system op- erating parameters, as a function of reactor power level, fuel circu- lation rate, and control mode. The experiments themselves and the asscociated analysis will of course follow completion o¢f the prototype, and are not included in the time period covered by this report. %6 Y, MANPOWER AND COST ESTIMATES @ Results of most of the investigations discussed in Sec. 3 should be available as a basis for the detailed design of the experimental MSBR. According to the proposed schedule for this reactor, the design should begin in FY 1968 and be completed by the end of FY 1971. The reactor physics program outlined in this report should therefore largely be com- | pleted by the end of FY 1970, and the manpower allocations and cost i estimates shown in Taeble 10 have been prepared with this schedule in mind. The total cost of the program, over the three~yesr period FY 1968 to FY 1970, is estimated to be about $3,100,000. The program outlined above is designed to provide, by the end of FY 1970, a secure basis for the design of the 150-Mw reactor experiment. In the ensuing fiscal years, 1971-1975, it will be necessary to carry on a continuing program of reactor physics investigations in support of the 73 MSBR concept. This program will comprise further analysis and evaluation of new cross section information as it becomes available, continuing = improvement and refinement of methods of analysis, further studies of i operational problems and characteristics of molten-salt breeder reactors e et e r as influenced by details of design, the search for better or more ec- cnomical approaches tc reactor control, and a continuing study of potential safety problems — in short, a continulng effort to gain & more complete understanding of the characteristics of this reactor concept, so that the twin objectives of safe, reliable operation and economical power pro- duction can be most satisfactorily accomplished. A need for additional { supporting experimental work may be recognized as the program progresses. | We believe that a support level of $200,000 per year for the five-year period FY 1971~FY 1075 will be required for this program. 5. CONCLUSIONG The reactor physics efforts that have been discussed in this report ! should provide a sound hasig for thoroughly reliasble assessments of the performance of = thermal molten-salt breeder reactor as proposed in Table 10. Manpower and Cost Estimates for MSBR Fhysics Development Program FY 1968 FY 1969 FY 1970 3-Year Total Section Activity MY Cost® MY Cost® MY Cost® My Cost® 3.1 Investigation of Dynamic Characteristics 0.7 24 1.6 60 2.2 84 4.5 168 3.2 TInvestigation of Alternate Core Designs 0.5 18 1.0 38 0.5 20 2.0 76 3.3 Development of Methods for Analysis 0 1.5 56 1.2 46 2.7 102 3.4 Cross Section Evaluation 0.5 18 0.5 18 0.5 18 1.5 54 3.5 Development and Improvement of Computer 0.5 18 1.0 38 1.0 38 2.5 9 Codes 3.6 Experimental Physics Program Lattice experiments — planning, 0.5 18 2.0 75 2.0 75 4.5 168 design, analysis Procurement, measurements 200° 200P 400 Dynamic experiments — planning 0.3 10 0.4 15 0.6 23 1.3 48 Totals 3.0 106 g.0 500 8.0 504 19.0 1110 ®Costs ere distributed roughly 80% for direct salaries and overhead, and 20% for computer charges. (Cost in thousands.) bIncludes estimated costs for all necessary hardware, including fuel, but not including value of figsile material used; includes also estimeted expenses of Pacific Northwest Laboratories for per- forming experiments. Le 38 i b Ref. 1, and, together with cperation of the reactor experiment, should permit selection and detailed design of a full-scale MSBR. The proposed program will result in improved nuclear data, in & much better under- standing of the dynamic characteristics of such reactors, and in con- firmed method of computation. 6. ACKNOWLEDGMENTS The author acknowledges with thanks the invaluaeble assistance of C. W. Craven, Jr., T. W. Kerlin, B. E. Prince, and others in the preparaticn of this report. g CRNL-TM-1 85 7 Internal Distribution 1-50. MSRP Directorts Office 9. H. A. Friedman Room 325, 920L-1 97. J. H. Frye, Jr. 51. R. K. Adams 98. C. H. Gabbard 2. G. M. Adamson 99. R. B. Gallsher 53%. R. G. Affel 100. J. H. Gibbons 54. L. G. Alexander 101. E. E. Goeller 55. R. F. Apple 102, W. R. Grimes 56. C. F. Baes 103, A. G. Grindell 57. J. M. Baker 104k. R. H. Guymon 58, 8. J. Ball 105. J. Halperin 59. H. F. Bauman 106. B. A. Hannaford 60. S. E. Beall 107. P. H. Harley 61. M. Bender 108. D. G. Harman 62, E. S. Bettis 109. C. S. Harriil 6%, F. F. Blankenship 110. P. N. Haubenreich 6L4. R. E. Blanco 111. F. A. Heddleson 65. J. 0. Blomeke 112, P. G. Herndon €6. R. Blumberg 113. J. R. Hightower 67. E. G. Bohlmann 114. H. W. Hoffman 68. ¢. J. Borkowski 115. R. W. Horton 69. G. E. Boyd 116. T. L. Hudson 70. J. Braunstein 117. H. Inouye 7l. M. A. Bredig 118, W. H. Jordan 72. R. B. Briggs 118. P. R. Kasten 73. H. R. Bronstein 120. R. J. Kedl T4, G. D. Brunton 121, M. 7. Kelley 75. D. A, Canonico 122, M. J. Kelly 76. 8. Cantor 123, C. R. Kennedy 77. W. L. Carter 124, T. W. Kerlin 78. G. I. Cathers 125, H. T. Kerr 79. J. M. Chandler 126. 8. 8. Kirslis 80. E. L. Compere 1l27. A. I. Krakoviak 8l. W. H. Cook 128. J. W. Krewson 82, L. T. Corbin 129, C. E. Lamb 83. J. L. Crowley 130. J. A. Lane 8. F. L. Culler 131. R. B. Lindauer 8. J. M. Dale 132. A, P. Litman &6. D. G. Davis 133, M. I. Lundin 87. G. de Saussure 134, R. N. Lyon 88. 8. J. Ditto 135, R. L. Macklin 8. A. S. Dworkin 1%. H. G. MacPherson 90. J. R. Engel ' 137. R. E. MacPherson 9l. E. P. BEpler 138, F. C. Maienschein 92. D. E. Ferguson 139, C. D. Martin 93, L. M. Ferris 140. C. E. Mathews o, J. L. Fowler 141. R. W. McClung ________ 05. A, P. Fraas 1h2, H. E. McCoy R 143, 1hk, 145, 146, 17, 148, 149, 150. 151, 152. 153, 154-168, 169. 170. 171. 172. 173. 17k, 175, 176, 177. 1i78. 179. 1.80. 181. 182. 183, 184, 185, 227322l , 205, D06, 227, 008, 229-243%, 2k, oh5 =246, 247, 248, 2LG=-063, 26k, 265-266., H. C. McCurdy 18. 0. L. Smith H. F. McDuffie 187. P. G. Smith C. K. McGlothlan 188, W. F. Spencer C. J. McHargue 189, I. Spiewak L. E. McNeese 120. R. C. Steffy A. B, Meyer 161. H. H. Stone R. L. Moore 192. J. R. Tallackson J. P. Nichols 193. E. H. Taylor B. L. Nicholson 194, R. E. Thoma .. C. Qakes 195, J. S. Watson P. Patriarca 196. C. F. Weaver A, M. Perry 197. B. H. Webster E. B. Piper 198, A. M. Weinberg B. E. Prince 199. J. R. Weir J. L. Redford 200, W. J. Werner M. Richardson 20L. X. W. West R. C. Robertscn 202. L. W. Weston H. C. Roller 203, M. E. Whatley H. C. Savage 204, J. C. Waite C. E. Schilling 205. L. V. Wilson Dunlap Scott 206. G. Young H. E. Seagren 207, H. C. Young W. F. Schaffer 206-209. Central Research Library J. E. Shaffer 210-211. 7Y-12 Document Reference M. J. Skinner Section G. M. Slaughter 212-221., Laboratory Records Department A, X, Smith 222. Laboratory Records Department, F. J. Smith LRD~RC G. P. Smith External Distribution D. F. Cope, RDT-CSR A, Gismbusso, AEC-Washington R. E. Heineman, Pacific Northwest Laboratory, Richland, Washington W. J. Larkin, AEC, ORO C. L. Matthews, AEC, QRO T. W. McIntosh, AEC-Washington H. M. Roth, AEC, CRC Milton Shaw, AEC-Washington W. L. Smalley, AEC, ORO R. F. bweek, AEC-Washington Division of Technical Informaticn Extension (DTIE) Research and Development Division, ORO Reacteor Division, ORO