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NUCLEAR WASTE DISPOSAL UTILIZING A GASEOUS CORE REACTOR

Richard R. Paternoster

N76 -17640 (NASA-CP-14641 9 1 NTICTVAF WASTE T)ISPOSP" L UTILIZIFG A MkSFOriS CC"P.^AC T S? (Florida nniv.) 42 p HC $4.00 CG^L 40B Unclas s- 0,3/44 14130

ABSTRACT

A feasibility study was undertaken of a gaseous core nuclear reactor

designed to produce power to also reduce the national inventories of long-

lived reactor waste products through nuclear transmutation. Neutron-induced

transmutation of radioactive wastes can bean effective means of shortening

the apparent half life.

S,9 (^ t{^, ;, -

8^9S^r I'll)

NUCLEAR WASTE DISPOSAL UTILIZING A GASEOUS CORE REACTOR

Introduction

A major objection to the increased use of nuclear electric-power generation

stems from envisioned difficulties related to the disposal and management of

hazardous radioactive wastes. The vast majority of radioactive wastes are

short-lived and easily managed with present waste disposal management techniques.

However, several isotopes present in nuclear fission reactor wastes have half-

lives of several thousands to several millions of years. These problem wastes

are the main point of focus in the debate over disposal of nuclear wastes. .

Ultimate disposal of the extremely long-lived wastes could reduce the

waste disposal problem to humanly controllable time scales. Removal of long-

lived wastes from the total accumulated waste would be the first step in this

direction. Separation of iodine-129, the trans-uranium actinides, and other

long-lived isotopes could be accomplished at several different stages in the

waste disposal cycle. This could mean extraction at the fuel reprocessing site,

after a five to ten year cool-down period, or after a 20 to 30 year decay

interval. These separated long-lived wastes would be disposed of by methods

which would completely remove them from the biosphere and hence any further

consideration. Extraterrestial transport of nuclear wastes (ETT) may become

feasible in light of recent advances in the space program. Ultimate disposal

into deep space or the sun is a remote possibility. The only other presently

known method of ultimate disposal is through the process of neutron-induced

nuclear transmutation. Induced nuclear transmutation of radioactive wastes

converts long-lived problem wastes into relatively harmless short-lived or

stable isotopes.

r . TRANSMu'rATION OF RADIOACTIVE WASTES

The concept of neutrun-induced transmutation of radioactive wastes was

first suggested by Steinberg et al. 11] as a means of effectively Hhortening Cs137, the half-lives of waste Kr 85 , Sr 90 , and This would be accomplished in

a neutron flux through the reactions

Kr8S(n,Y)Kr86(stable)

91 Sr 90 (n.Y)Sr S Y 91 S -; Zr 9.48 hr 58.6 d

137 (n,Y)Cs 138 6 Ba138 Ca (stable) 32.2m

In examining the neutron capture cross sections for those reactions they con-

cluded that neutron fluxes approaching 10 16 n/cm2 sec would be necessary for

shortening the half-lives appreciably. They further suggested that the most

logical approach would be through the use of dual-purpose facilities for power

production and waste 'ransmutation. In such a scheme the major cost of waste

burning, namely the cost of producing neutrons, is written off against the

cost of electric power generation.

Wolkenhauer [2] has studied transmutation of high-yield fission products 137 Sr 90 and Ca in a controlled thermonuclear reactor (CTR) blanket. Core and

Leonard [3] have also studied transmutation of massive loadings of Cs 137 in 137 CTR blankets with special flux trap designs. Cs represents the most dif-

ficult case for this technique because it has the lowest thrrmal neutron cap- :i ture cross section of the moderately long-lived fission products. 'fheHe

schemes appear attractive since the CTR power plant operating on a deuterium-

tritium fuel cycle would produce an excess of neutrons. Within a few decades

such systems may be found technically and economically feasible. C,

r 141 w

Claiborne 14j has considered disposal by transmutation of transuranium

actinide wastes. Most of those long-lived alpha emitters have appreciable

neutron capture and/or fission cross sections. Table 1 gives the capture and

fission cross sections at thermal and resonance energies used by Claiborne.

Actinide transmutation occurs by direct fission such as

Am242m(n.f)

and

Am244(n,f) ,

or by neutron capture followed by fission such as

Am241(n.Y)Am242(n,f)

and

Am243(n.Y)Am244(n.f) .

In this scheme, those long-lived actinides are continuously recycled back

into power reactor fuel elements and in this manner burned out. At fluxes

presently available in power reactors (1013 - 5 x 1013 n/cm sec) this would be

a somewhat slow process spanning over several decades. Furthermore, serious

problems arise in handling and safeguarding the recycled actinides during

fabrication of fresh fuel elements.

Requirements for'Tranemutation of Radioactive Waste

Given a high enough neutron flux or a long enough period of irradiation

any isotope with a non-zero neutron capture cross section can be transformed

into other isotopes. However, the transmutation of radioactive waste imposes

a number of constraints upon the process. 14:'

' TABLE 1. THERMAL CROSS SECTIONS AND RESONANCE INTEGRALS FOR CAPTURE AND FISSION IN THE TRANSPLUTONIUM ACTINIDES it

Thermal Cross Sections Resonance Integrals Nuclide (barns) ( barns) Capture Fission Capture Fission r X241 95 925 3.1 2150 0 242m 95 2000 6000 0 0 Am2 4,3 95 0 2900 0 0 X243 95 105 .5 1500 1.5 X244 95 0 2300 0 0 245 95 0 0 0 0 242 96^ 30 5 0 0 243 96 200 600 500 1850 244 CM 96Cm 10 1.2 650 12.50 245 96Cm 343 1727 120 1140 Cm, 96 1.3 0 121 0 96 CM 247 60 120 500 1060 248 96Cm 3.6 0 170 0 Cm 249 96 2.8 50 0 0 Cm250 96 2 0 0 0 Bk249 97 1450 0 1240 0 Bk250 97 350 3000 0 0 Cf249 98 450 1690 1.5 2920 Cf250 98 1900 0 11600 0 Cf 251 98 2850 3750 1600 5400 Cf252 98 19.8 32 44 110 98Cf253 12.6 1300 0 0 98Cf 254 50 0 1650 0 143

The selection of prospective candidates for waste transmutation must begin with the radioactive wastes themselves. Reaction rates for neutron- induced transmutation processes can be specified in terms of neutron capture and/or fission cross sections. Unfortunately, high-yield fission product wastes have small neutron capture cross sections. An example is the cap- tune cross section for the (n,y) reaction in 10.7 year half-life Kr85.

At the time of Steinberg's study (1], the thermal neutron capture cross section of Kr85 was believed to be 15 barns. Using this figure it was con- 85 cluded that disposal by transmutation at Kr could be economically and technically feasible. A recent measurement (5] of this cross section found it to be 1.66 i 0.20 b. Thus, an order of magnitude increase in the fluence is required to reach feasible transmutation levels.

Another criterion for neutron-induced transmutation of radioactive waste is isotropic purity. If maximum utilization of available neutrons is to be achieved isotopic impurities of a given waste isotope must be eliminated or reduced to 85 lowest possible content. in the case of Kr transmutation, three stable krypton 86 isotopes, Kr83 , Kr84 , and Kr are present with Kr 85 in fission wastes. The 85 low concentration of Kr in krypton wastes (7.2 percent) and its low capture 83 cross section compared with Kr (ac = 2.5 barns) makes enrichment of Kr 85 necessary.

A final requirement for successful transmutation of radioactive waste is that the end product nuclide of the transmutation process be no more hazardous than the initial waste product. The ideal end product of a transmutation process would be a stable nuclide with a zero neutron capture cross section.

A nearly ideal nuclide in this respect is waste 1129(t1/2 15.9 x 10 6 years).

The transmutation reactions in this case are

,- t 3f I

NUCLEAR WASTE DISPOSAL UTILIZING A GASEOUS CORE REACTOR

Introduction

A major objection to the increased use of nuclear electric-power generation

stems from envisioned difficulties related to the disposal and management of

hazardous radioactive wastes. The vast majority of radioactive wastes are

short-lived and easily managed with present waste disposal management techniques..

However, several isotopes present in nuclear fission reactor wastes have half-

lives of several thousands to several millions of years. These problem wastes

are the main point of focus in the debate over disposal of nuclear wastes.

Ultimate disposal of the extremely long-lived wastes could reduce the

waste disposal problem to humanly controllable time scales. Removal of long-

lived wastes from the total accumulated waste would be the first step in this

direction. Separation of iodine-129, the trans-uranium actinides, and other

long-lived isotopes could be accomplished at several different stages in the

waste disposal cycle. This could mean extraction at the fuel reprocessing site.

after a five to ten year cool-down period, or after a 20 to 30 year decay

Interval. These separated long-lived wastes would be disposed of by methods

which would completely remove them from the biosphere and hence any further

consideration. Extraterrestial transport of nuclear wastes (ETT) may become

feasible in light of recent advances in the space program. Ultimate disposal

Into deep space or the sun is a remote possibility. The only other presently

4 known method of ultimate disposal is through the process of neutron-induced

nuclear transmutation. Induced nuclear transmutation of radioactive wastes

converts long-lived problem wastes into relatively harmless short-lived or

stable isotopes. 144

I130m I129(n.y) f T.! 8.9m

-.34i 1130 1131 < (n.Y)

0-112.4 hr 5-181 130 Xe131 (n,y) Xe132 Xe (n,y) (stable) (stable) ( stable)

-.._._ r Thus even with several reaction paths the vast majority of 1 129 will be con-

verted to stable isotopes of xenon. The situation is more complex in the case

of actinide waste transmutation. In a sustained irradiation of waste actinides,

such as curium, quantities of higher actinides, such as berkelium and califor-

nium will inevitably be formed. These higher actinides are generally alpha

exii.ters with varying half-lives.

P Transmutation Systems and Waste Disposal

The transmutation of 1129 is first examined in the following. After a I 128^ 1130 . 1131 . 1132 160-day decay period, the short -lived iodine isotopes I121 129 and I133 decay away leaving stable and the 15.9 x 106 year half -life, 1. 127 and The spent fuel from present day LWRs contains about 40 grams /MTU of I

231 grams /MTU of 1129 at 160 days after discharge and 33,000 MWD/MTU burnup. A These two isotopes constitute roughly 0 . 78 percent of the total fission product

wastes with the 1129 isotope being about 85 percent of this amount.

The proposed means for disposal of 1129 by induced nuclear transmutation

is by neutron capture to 1130 which beta decays with a 12.4 hour half-life to

stable Xe130. The thermal neutron capture cross section for the (n,y) reac- 1129 tion in is 28 barns and the resonance integral is 50.4 barns. The re-

maining 15 percent of LWR iodine wastes is stable I 127 with a thermal neutron

absorption cross section of 6.2 barns and a resonance integral of 177 barns.

4 TI

A schematic of the proposed nynten In shown In Nly •,ure I. The ii', percent

1 129 -15 percent 1 127 mixture available directly from the fuel repr or. esr:ing pinnt as li (III hd 1 2 would I,v enriched to 95 percent 1 ) In nn Inr,topi- enrlrh- ment facility. A laser isotope separation facility would be ideally suited

for such a small scale operatlon. The enriched 1 179 would he Inr:ert,d In

the core of the power-burner reactor in a suitable concenrration. T•Ite 1129

upon capturing a neutron and decaying to stable xenon lsote,pes, primarily

Xe 130 could be harmlessly vented to the atmosphere or sold commercially.

The specification of 1 129 enrichment is optional, but desirable from the

standpoint of maximum neutron utilization fur 1 1`9 hurnlnt;. If 100 kg of 129 unenriched iodine waste (85 percent 1 -15 percent 1 127 ) is initially loaded

for transmutation and exposed to a thermal neutron fluence such that 50 kg

1129 of are transmuted, then about 2.6 kg of stalle 1127 are also transmuted

to stable Xe127. Irradiating 95 percent 1 129 enriched iodine in the same

fluence results in the transmutation of 55 kg of I12JI9 and only 0.9 kg of 11

The actinide transmutation system would involve loading of several waste

isotopes simultaneously for transmutation. 71ie actinide wastes considered

here are the isotopes of americium and curium. 11te amounts of the sil;nlfi

Cant isotopes present in 1.WR fuel wastes following a Len year decay period

241 are shown in Table 2. The two prevalent isotopes are Am and Am 243 . 11te

most important isotope present in curium wastes is th,• 18.1 year hall -life

Cm2 A decay interval of ten years aft er (II r:c;hnrgo f rorn t he react or is

reasonable to allow the wastes to cool clown. Following this cool-do •.:n period

the wastes would be loaded into the power-burner reactor for transmutation.

After irradiation the transmuted waste would be rt • moved from tIII , reactor and

Htored to allow decay of short-lived act Inl(les. Fol lowing this: cool-duwit

period the fission product wastes wuuld he chemically removed and the r,_-u-ttning

A I iI I f.

FUEL

POWER FUEL SOLID +LIQUID REACTORS REPROCESSING WASTES PLANT

u + PU I RECYCLE

IODINE AND NOBLE GAS GAS NOBLE WASTE STORAGE RFCOVFR H 1127 85% 1 129 - 15l

1 129 1127 ENRICHMENT r0 ATMOSPHERE 1129 99%

XE130 GASEOUS-CORE BURNER REACTOR TO ATMOSPHERE OR COMMERCE

GASEOUS-CORE VOLATILE WASTES

Figure 1. Pathway for 1 121) Waste 'Transmutation in Lhe GascOu:.-C'ore Reactor. m, 1 ,

I^1

TABLE 2. RELATIVE AND ABSOLU'T'E AMOVN'I'S OF AMERICIUM AND CURIUM ISO'T'OPES PRESPN'T' IN AMERICIUM AND CURIUM WASTLS FROM A'T'LAN'T'IC 1 ItLVLR ENCC LWR F II C:. AT TEN YEARS AFTER DISCHARGE AT 33,000 MWd /MTU DURNUP

Concentration Fercent of in Grams in Isotope Isotope Spent Fuel Spent Fuel Load Present in Grams/MTU Element

241 Am 48.3 4166.84 32.37 242= Am .92 79.37 .62 242 Am 1.10 x 10 -5 .001 Insi(jnifi=anti•• l1 243 AM 100 8627.0 67.01 244 Am Insignificant` Insignificant" Insignificant••• 245 Am Insignificant` Insignificant`• Insignificant•**

Total Am 149.22 12873.21 100

0242 2.21 x 10-3 .191 Insignificant`•• 243 Cm .0679 5.66 .28 244 CM 22.1 1906.57 89.98 245 CM 2.14 184.62 8.72 246 cm .251 21.65 1.02 247 Cm 3.31 x 10 -3 .286 Insignificant` 248 cm 2 .29 x 10_ 4 .0198 I ns i (In i f icarit"` 249 Cm Insignificant' Insignificant" Insignificant"` 250 Cm Insignificant* Insignificant** Insignificant"*

Total Cm 24.56 2119.20 100

`Denotes 5 10 -5 gramslKm . "Denotes < 10 -3 grams. ***Denotes S 0.1%. Y} Fc

actinides would he recycled into the next loading of wat+tc. A schenuclic of

the actinide transmutation system is shor.wn in Figure 2.

THE GASEOUS-CORE NUCLEAR REACTOR CONCEPT

Gaseous-fueled, externally moderated cavity reactors have been under

theoretical investigation in the United States since about 1954. the primary

motivation for developing the gaseous-core reactor has been for high specific-

impulse nuclear rocket propulsion. The two configurations that have received

the most attention are: (1) the open cycle concept, in which the fissioning

fuel is partially confined in the core by vortex-generated hydrodynamic forces,

and (2) the closed cycle or nuclear light bulb concept, where the fuel is

physically contained within a thin, radiation-transparent wall and recirculated

for removal of fission product poisons. Uranium-235 fueled core regions

ranging from 0.5 to four meters in diameter surrounded by reflector-moderator

regions of practical thickness require minimum critical masses ruiping from

one to thirty kilograms. The critical particle densities of fissionable nuclei

correspond to molecular densities of gases at less than atmospheric pressure.

Of the possible gaseous fuels enriched uranium hexafluoride is the most

Obvious first choice. It sublimes at 54.6°C at atmospheric pressure and re-

mains completely in the gaseous state at temperatures below 2000°K 171. Three

gaseous-core experimental facilities have been constructed to date. A 1.5

fit kW(t) U 235 F 6 fueled research reactor was constructed USSR lit late

1950s (8). '!tiro gaseous-cure research reactors have been constructed in the

United States. The most recent was a spherical gas-core critical experiment

(9,101 performed in early 1910 at the National Reactor Testinf; Station.

The gaseous-core reactor concept has found little tetrestial application

to date. A recent patent disclosure II11 proposes a UF 6 fueled reciprocating, t /, 1 1

FUEL

FUEL POWER WASTE REPROCESSIN REACTOR STORAGE PLANT 99.5 TO 99.9% FISSION PRODUCTS U + PU U + PU 0.1 TO 0.5" RECYCLE ACTINIDES STORAGE 99.5 TO 99.9% (10 Y R DECAY) ACTINIDES OTHER THAN U + ^U

CHEMICAL SHORT-t I VE D SEPARATION ACTINIDES

r WASTE LOADING

GAS-CORE BURNER REACTOR

Figure 2. Pathway for Ac • t inide W.1.1 ;t C 'I'r.ZI I f ; Inllt.atloll in the Gascou_;-Cure Reactor.

r VIC r

engine. .his p,rlsed nucIvar reactor can e • pl 1:, analul ••„uri Iu Ili, • I[I It , ,i,•rl

combustion engine; on 1 y L he work ing f lu ld is a UF 6 -Ile s w I xl ut e and the !!park t, of th y i plug is an artificial neutron source. A detailed analysis 1121 concept

concludes that efficiencies greater than 40 percent are attainable. An 8400

MW(e) MID power plant design has been proposed [131 which takes advantage of

the high temperatures generated in the fissioning uranium plasma. This analysis

concluded that large commercial power plants using a gaseous-cure nuclear heat

source coupled to the MHD generator could have thermal efficiencies in excess

of 70 percent. Another design study by Gritton and Pinkel 1141 indicates that

attractive power levels in reactors of practical size can be obtained with gas

pressures and wall temperatures within the potential capability of known

structural materials. For a spherical gas core reactor with a radius of 152.4

cm to generated about 4000 MW(t) would require a uranium partial pressure of

about 11 atmospheres. Practical heat removal considerations wit11 water and

lithium as coolants limit power output to 200 to 1500 MIW(t), respectively, for

a static fuel configuration.

The Dual Purpose Gaseous-C ore Power-Production s Waste_Ttansmutation Facility

The design of such a dual purpose facili!y would Ideally combine both the

economic production of electric power with attractive rates of transmutation

for radioactive waste disposal operations. This means hiy;h efficiencies and

high neutron fluxes are necessary.

Figure 3 shows a schematic cross section of the• proposed gaseous -core {5 reactor. The fuel is UF 6 enriched to 6 percent U The interior cure region

is four meters in diameter stir rounded by a reflc..tur -modrratov of Ii l ll with r►

t', 'ckness of 500 cm. 11u• D 1 (1 Is vncalce • d In in outer rci lector of `,till cm

j X i

4TROL D_ 20D LIVE

'ARGET PORT

COOLANT COOLANT INLET OUTLET

r GASEOUS•FUE REGION )RESSURE SHELL

SURFACE PROTECTIOP ;RAPHITE LAYER EFLECTOR

CYLINDERICA D20 CAVITY IODERATOR VESSEL PLENUM

ZGASCOU-S •FUEL OUTLET

Fiquve 3. Gaseous-Core Power-Burner Reactor. C

t r i

of reactor grade graphite. the additional nulei y,^.iphll^ i^ I le to y hi 1)111- t f vided to return as many of the neutrons to the core as possible. t Figure 4 shows a pussihle flow diagram fur the UF h g;im4" n cote i tar tor.

Fuel flows through the cure, and is heated internally by fission energy. The

gaseous UF 6 leaves the reactor at 18UU°K and MOO psi and enters the regeneratur

where it gives it some of its sensible heat to an intermediate helium cooI;nU

loop. The UF 6 then goes through the primary heat exchanger giving up the

remainder of its sensthle heat and its latent heat to the helium working fluid.

From there the UF 6 liquid at 600% enters a sump tank. Here all the volatile

fission products are removed. These include Kr. Xv, , II= ide annpounds of iodine,

and several others. Nonvolatile fission products could be removed by an on-

line reprocessing stream utilizing the fluoride volatility process 1151.

From the sump tank the liquid OF is pumped through an evaporator at 8005

and returned to the core.

Electric power could be produced if the helium working; fluid was expanded

in a turbine and recirculated back to the heat exchanger. Assuming the waste

heat from the UF 6 -lie cycle was rejected to the surroundings at 100°F (4 )'C)

and ideal Carnot efficiency of 82.4% is obtained. This simple calculat full

neglects all thermodynamic irreversibllities and heat losses. 'life point is,

however, that this high efficiency is the result of the hfg;h temperatures

produced in the gaseous-cure nuclear reactor. If 15% of the ideal Carnot

efficiency could be achieved In practice (Iypic,il In power plant design), a

system efficiency of about h5% would result. Ki t h ram +re , If 1,)Z of the

reactor power is used to circulate the gaseous fuel anti working; fuel this

leaves an overall plant efficiency of 5U%. A power plant with this efficiency

would reduce the amount of heat wasted to Lite environment by about a factor

of two over present-day power plants.

I '^ 4 i`

.. w y 1^ U1 ^ ][ .' Z Q (n O Y d Q ^ z 7 ^ r-1 N x r Lai ^O W JUG 1 to Q WW — I [y W 1- Cc SNU J O 1 ^, FQ Oo H C I 4 o R O I 1j > a r- _ ro Cc

o ¢ I a ^ u ,^ c z •.^ = Z I 1^^ 4^ U X /^ J 7 W v z I ►-+ o O

a s 1 I Lr }}} U1 > a W / I o ^ a O I v z ^ a x 1 ro^ ^? `- ^ ► o 0 1 C7 a a ¢ „ ul 6-+ ej. U I O N ¢ ¢ o E o ^^ a ro 4 ^^ W :T

♦ Q J ^ 1 ¢ W r ¢ ¢ 1 rl W 11 N ^~ I 1 ro tC) ^ -ZJ I Q O V) JQ► Q¢ O u i co N WW¢- 1-

N ¢ J U r+^ ' p 0. b ]

2 ^/ I U J Z W L a I 7Wa •O IO L,a m Q1 i ( 1 o 1 >4 WD u N 4 W .,4 > T ^^ ¢ Cis V ri LJn ¢ 1. lY ~'' m W O N U W

r , III!; The 1 wo uu l or rad role rel le( • t ed by U.', 11)4 • 1 rf t; ul 1) X 11 ► e(lii 11 e?; a

minimum of 14 kg of U to nLrintaln criticality in the cold, clean condition.

lh ►e to the. bill laup of f Ir;r;Ion product polrun ► v at hol , oporat inl,, rond I I ion.,;,

additional fuel must be added to compensate for this negat ive react ivity effect.

U23SI.,S Thus, a mass of 140 kg of (L► 67 enriched UF i^) ir; m. ► intalned in the core

at any particular instant in time. Control of excess reactivity is I,y means of

adjustable control rods located In the graphite reflector and by the addition of r burnable poisons in the U 2 o. Fine reactivity control would ideally be through

the use of these burnable poisons; specifically radioactive waste products 129, such as I continually dissolved and burned in the 1),) 0. '11it, control rods

would then only be used for startup and shutdown control. Further reactivity

control would be aserted through the addition of burnable radioactive wastes

in the 16 target ports. These ports, located a distance of 12 cm into the 1)2()

reflector-moderator, would be used to load and transmute actinide wastes which

are not suitable for dissolution in the moderator. The large negative tempera-

Cure coefficient associated with thc gaseous fuel density, and the Doppler

coefficient of the 6 percent enriched fuel should make the reactor hoth verY

easy to controi and inherent l y sa l7 e. I•:n►c• rgency shutdown runt rol i>; provided 10 , by injection of hi„h-pressure, gaseous BF 3 (thermal neutron absorption cross

section, 3813 barns) direct l y into the core.

Let us now consider the w. ► ste transmutation aspects of the proposed

facility. Mentlon has been made of lo.+ding radioactive waste iodine-129 in ,I the U 2 0 reflector-moderator and actinide wastes in the target ports of the

129 reactor. . Waste 1 could be loaded continuously at the same rate . ► 1 which

it burned up and removed from the reactor. Actinide transmutation would Involve

batch loading and subsequent recycle alter an appropriate hurnuut period.

C YI

The yt► estloil remalr► s ul how murk radioactive wa!;le can he 14 ' a414-1 1 Into the reactor. Here one would like to load the largest amour► t pouttlhle without poisoning the reactor out of operation. Ud ioactive Iodine-129 added to the

D2 0 will primarily affect the thermal utilization factor, f. Using the known fuel composition and reactor configuration, values of t were computed for the feasible concentrations of 1 129 added to the D 2 0. A maximum concentration of

2600 ppm of I 129 , or about 425 kg, can be added to the D 2 0 belore the reactor becomes sttbcritical.

The insertion of actinides into the core of the reactor will have two complicating effects; the addition of fuel (and hence neutron.,;) in the form

245, of fissile Am242m, Cm 2 d , Cm etc., and the addition of neutron-absor-hlrK

fission products during the course of neuttor► exposure and act lit hutnul1.

Claiborne (41 has concluded that the effect of recycling 99.5 % of the waste

actinides back into light water reactor fuels is minimal. A maximum average

reactivity decrease of about 0.8 percent is attained in about five cycles

through the reactor. This reactivity decrease can be counteracted by only

about a 2% increase in the fissile material, which would amount to increasinf;

the fuel enrichment about 0.1%.

Many of the actinides have substantial thermal Iisslon cross section!;, shown in Table 1 and hence, can achieve criticality if assembled in sill ficicnt amounts. The only significant transpl tit onium actinide nuclides for which A 144 criticality in a thermal sy!;tem could occur are Cni that contain , ; Iii ;ile (:M" 45 252 '", I . ► nd (I f t hat cunt a Ins ;tit I is lent Lar) ,,et. precursor , Cf Ili r:tu ;e of the vet v

24'► low potential critical tn. ► ss of Cm (11.42 g) .111d ► :t 251 ('40 );) the po!;slbi I It v of thermal criticality must be anticipated and included in the design. The published data of Clark 1181 indicates that concentrations of 15 g/liter for 245 251 Cm and 6 g/liter for Cf may achieve criticality In a lit mof;eneouS mixture. 9 I'll)

Advantage_. and Disadvanta^e• r► of the l;a:,ruu:I Cure I'll keact or l:uurt l t

Compared wit it convent tonal sol Ill-cot e unc Ivor react t i t s 1 111 • ga!;eou!• colYe

power react or Wollld It. IVe ,I • vl • I.I I ad y aIII 1 111'ti/' .1.1 v.111 I •It;tand• .'i 411 !..Id

vantages derive mainly Irom the fact that the fuel is in the gaseous tit at V.

A basically simpler i"I vrna l rea ctor core wi t hoot- t ht• )It'l • d l of I iw I pills,

cladding and spacers is posnible. Elimination of furl elemont fabrication

costs, UO 2 to UF 6 conversion charges would result in conrslderable savings

in the monetary and energy cost of power production. Continuous replenishing

of fissile fuel and cunt inuouH removal of fission product puisurls wuultl };Ive

rise to t b y puss ib 1 1 I t y of cunt Inuous operat. ion with Increased I uo • I and 238 neutron economy. Due to its inherent low inventory of li and its 111};11

neutron flux, less actinide waste is }venerated by tit,- OF gas core reactor.

Operating nuclear charariviL Llcs of the gas-cure reactor were sinu11,11od with

the ORICEN computer coda 1191. Figure 5 shows the hit iIthip of t ission 1)1oduct

and actinide waste in the UF 6 reactor and in a typical LWI% of ey1.131 power

output as a Iunction of time. After three years of cont-inuous operation the

UF 6 reactor contains about one order of magnitude less of tun};-lived actinitlt-

waste Lhan the LWh. This behavior Is due to th( • steady-state u.lture of

refueling, operat Ions, an upposed to the tt bate:ll " Int-1 loading, It-ChIIi-Int . of

present Clay LWR:s. In addition, with continuous lul • I ( U )lr ) addit ion .old remov.11

Of xenon and other volatile fission product poisons, fuel burnup:: of 250,O00

to 300,O00 NW(IMT11 w t- re olitaill-I In t ht• eompnict result::. I"ht• prat t iral 411 r.

advantages of the 11F 6 gas-curt , re.ICtot ,Ire primatiIv tIII,- to tht• inevitable

dissoviatIon of the fuel and its associated fluoride compounds. 5peritically,

they are: fuel dissoclat ion, corroslon-limited I1IvI In , of rote .Intl prim•1ty

:;ytitenl conyloaent rs, .Ind accoont:thi l ll v of heavy-elemcnt bv - product 1 luorideli

M •

I r 1 11

c^ I^

^ O

u1 U) .r4 W

N O ^1'

^ Q w crW O U 1: O DC ^+ O U F— p u W U a ►+ a cr w 0 3 u c o a 00 a ^a J _^ v

L +J

C 11 ^ O U c > 71 Q it ^ (I; 111 14 44 O O +^ J U (: fd o^a v 164 ^ v C) A. c C) O "' O lL V

C) v U O r U U• I+. -1 (st-lol y -wvbJ) G31VinwnDDv iism 3AUDVOIGV8 1 III Llle cIrculat lisp, l tit- I sy:111 1 n1 produced .11 ItIg1 11uwl-1 11111-1 .11 Ion. 111-1 . 1111.14. h Is convlderabIt: experience In hand 11111; I l(luld I IIIUrfill , n.111 : 11^)111 1 1`140114-11 1

Salt Reactor (MSR) program. Also experience In handlInl; ill' it the gaseous t^ _ ` p diffusion plants should prove extremely useful.

RESULTS AND CUNCLUSLUNS OF THE TRANSMUTATION STUDIES

^r Transmutation Rates and Reactions

The UF6 gaseous-core reactor was chosen for this study as a nruteon source

because of its obvious potential for producing high neutron fluxes and generating

useful power simultaneously. A reference reactor design was produced whlch

could satisfy both of these requirements. The reactor power specified was

3425 Mw(t) which resulted in an average thermal neutron flux across the core 14 2_ of 6.44 x 10 n/cm sec. The ORICEN computer code (19] was then used to studv I

the transmutation rates of several waste isotopes present in the discharged

fuel from reactors of the current LWR-type.

129 241 2421n 243 The waste isotopes chosen for the study were: I Am Am Am Cm244^ 246 Cm Cm Cm andand Cm The amounts of the initial waste loading were

determined from the composition of 20 discharged LWR tuel loads stored for a

decay interval of ten years. Tile compo.:ition of the Am and (:in loading; is 1129, the same as that given in Table 2. The quantities of Am and Cm used In

the study are approximately equivalent to the waste product in 60 reactor

years of operation.

III rate s of transmutation it is n,-cesmAry to ch.Iracterize the

neutron energy spectrum of the reactor and the cross section dependence on

spectrum for the particular target material. T11us the reaction rate depends

not '3nly on the magnitude of the neutron flux but also the shape of the neutron

h energy spectrum. 'I11e contribution of resonance energy neutron; to the reaction

.^ I ` I r • I 'iV

Cruts Is expressed by the rat to ul re!cuM:utce-tu-t ht' tmal nt • ui tun I Iux("i, 'Tres/+t11'

The rates of waste transmutation were found to strongly depend on tit(- value

of 4'res/'ntit* 'l'hc transmutation rate curves for I are shown In Figure h

for three values of /Q) in the range of interest. In this calculation 4, res tJt the thermal neutron flux was 6.44 x 10 14 n/cm2 sec and the initial loading is 129 29 400 kG of I (tilee waste I accumulated from 60 reactor-years of LWR opera-

tion). The effective neutron capture cross sections of 1 129 for the values

of 0.058, 0.29 and 0.58 were calculated to he 20. 0. 32.3 and 40.9 cores/nth barns, respectively. 14 For the specified neutron flux of 6.44 x 10 n/(-m 2 sec, tralismutattun

rates of 68 kG/yr, and 77.6 kG/yr were calculated using the values of 4,resNtit

- 0.058, 0. 29, and 0. 58, respectively.

The calculations for actinide transmutation were performed lit a similar

manner. The calculated transmutation rate curves of americium wastes for

0.058, 0. 29, and 0.58 are shown in Figures 7, 9, and 11, respectively. O res 4tit The same curves for curium wastes are shown in Figures 8, 10, and 12 respectively.

It should be remembered in examining these curves that the Ain and Cm wastes

were irradiated at the same time in the neutron flux. Furthermore, these

curves are the total waste composition of the initial waste loading plus the

actinide by-product waste produced by fuel burnup in the UF 0 gaseous core

reactor. It was felt that this would more realistically take into account

the waste produced In affecting the transmutation process Itself.

The curves presented In Figures 7 through 12 can ht• better understood

with the aid of the cross section data presented in Table 3. The capture-to-

fission ratio, a = o Y /o f , is important in actinide transmutation. T'he relative

probability that a compound nucleus decays by fission is 1/(1 + (l), and the

relative probability that it decays by emission of capture Y-rays is ,/(1 + ( A). C,

10 D58 N N

N z 29

r W Hw Q 58 3 6. O l7 X 101

0 10 U 360 720 10140 144() 1H00 DAYS OF IPkM)IATION le Fiyure 6. 1129 waste tranSmlltation versus irradiation time for 0.058, 0.29, 0.58. 4, res /W th - Ih+

10 W HU U.1 D

Q H UM

Q 0 1 10 H I

a1y u NE^ Q

I:. O

0 10

-1 lU U 360 720 IUHU 1441 1600 DAYS OF IPPAUTATION

Figure 7 Americium waste transmutation versus irradiation time for 4, res / ^th - 0'058'

. ! .*

! 1^1

f j I lU2

14 a i H r^ V Ev a 101

N u H CL C)

Q J N H'l. J t.l 100 L2. [J AN

4. 0 U

-1 10

1 1 1 1\ I I I 1 I_ 10-2 1 U 36U 72U l uhO 14.10 1800 i DAY: OF I:01ADIATILI

t^ Figure d. Curium waste t-ansinutation versus irradiation time for ''res /2tti _ 0.058.

I ^p

14 7. I+ 10 n /cm -sec r Wad i nqs : - 8,.34 kg { ^r - 1.58 kg I+ ^ - 172.5; kg 10

aW .a (1 i? a u U Q 101

t,7 N'L Nt tl 3 W H 4

L. U x 10^

10 -1 1 1 1 1 f 1 1 1 1 1 1 1 1800 U 3E0 120 1U1:1u 1,;4U c riAY S OF 111RAU1 AT I ON

Figure 9. Americium waste transmutation versus irradiar_I•n i time for Q res /m th 0.29. I(II1

V ,

101

W Q H

.7Z ;J

H 1.) aH 0 N 100 r a c^ Nz H 3 aw w E~

I" C) u 10-1

-? 1 0 0 360 72U IUAU 1,14J 1800 DAYS OF IRIW)IATION r Figure 10. Curium waste transmutation versus irradiation time for %Q res th =0.29. i

Ii ^

W HQ U z;.J

H H LZ n < 101 U Hz Hz Q

3 U. O

1UI

1 11 1 1 1 I 1 1 1 I 1 10-1 1 U 360 /lU luou 14.10 1600 DAYS OF IKPADIATTON

Fia_ure 11• Americium waste transmutation versus irradiation time for res / ^th = 0.58.

- - _ -- - A.

. . I 1 10 w Na a ua

.a d i-i H

0 l^ 10 H H Z'

w H U) Q

I ^. U 0 :L

10-1

0-- t 10-

DAYS OF IRRADIn•r1oN

Figure 12. Curium waste transmutation versus irradiation time for 0 res / ^tti ^ 0.5H. I

1 a/ ,;n r

rD r+ ON En CD ^n

U OD r1 r• 1 cn O o st O v OD to r~ N b w r o Ln o ., o 44 co V) o Ln o A v r v r- I^ 44 v rn .-1 .1 CD .1 r. b

N W z5 W O O ^D ^O r1 w M M .i qD OD r 41 OD o^H N ET v (n N F > 0 U W W cn [ w c^ N r'1 CO O in CD cn x ' 1 1 o Ho cs o u; v 1 a, O m O v O xxo (N %D UOZ

w^4 w+ r O N v ED O ^Wm w CD r . i rl N O 04 N w+ v O v .1 —0 b r O 7 ^ Hm a w O O LI) O^ Ln 44 4 %D O r ON u1 W O 41 N N t!1 N .-1 N ^fl T F. ,Eno b C Q 3 u ul ri 7 E to Ln N r p^ f"1 ^ UW r1 ,-1 I H x tD O .+ O N O 1 .••1 v WR0 E14 w W w r o r ui W OD o OD M m v ul O x v r v .-^ w v M o O b . [+. rl

W w r O M to -^ rn W O kD In 1n v N .-7 Gl r N .-1 CO W .--1 N T .-1 r Q b F

n^ E CJ 1 N rn M e In ^D '•'^ V r}' •? rf 'cT •7 V N N N N N N EE N[[ 2 r

i L 11":t

'1'1111+1 a 111y' ll vu 1111 . of ,t it I4-w 111Id..11,lllly of 1 In1114111 I11 11„wl11y. f► 4•u111oil

ah11orr" loll, d vonvet'rtrly, a low vtllue of 1t meaurl .1 1111,11 plohabl I II y of

fission.

From the data In 'fable 4 the major reaction paths ran be ldentitled. The

major reactions for the americium isotopes are

Am Z4Ln► 241

Am242(n.y)Am243 r and

Ain 243(n,l)Am244(n,f)

243 Am i Tl►e saturation behavior of Ain is due to by-product 243 produced by succes- j 238 live (n,y) reactions in the U present in the fuel of the reactor. The

major reaction paths at the curium nuclides is not so readily identifiable.

It can be said, however, that successive (n,y) reactions on the curium iso-

topes are responsible for the production of significant amounts of nuclides

with mass number greater than 244. 11 ►e production curves for significant

transmutation by-products with A 2 247 is shown 111 / 1 1' = 0.29 and 0.58 in res th Figures 13 and 14, respectively. The by-product production curve for 4,rt,s0th

= 0.058 has been omitted because the quantities of these isotopes produced

in this case is at least two orders of magnitude lower than in the others. A

I1ie Effect of Transmutation on the Hazard Potential of Radioactive Wastes

The relative inhalation hazard (R11111) and the relative ingestion hazard

(RIGH), are quantative measures of the potential danger which could result

from release of radioactive wastes into the environment. The success of a

particular transmutation scheme can he judged in terms of the overall reduc-

tion in the values of the RIG11 and Milli following transnn ► tation.

.3 ()

r —

i(,Y 14 ^ T

J ,^

-^ .^

10 1

O 0 r 6k: lO ^

^ ^

n= - lO l

- lD 3 DAYS OF IxoxolxTzow

^ Figure 13. Transmutation by-product production versus ^ irradiation time for ^ ' ras/^ th = 0.58.^

. .

d _^. T A.. 1

f/ 1

i 1 /0 r

III

i I

I r r .•

101

0 41 U

L 100 q ^ a H U I . a I D I 0a 1 a I

ao I 4. O U x 10

2 I 1 1 I f l 1 ) 1 I 1 l 1 lU_ 1800 0 360 720 IOcJ 1410 R DAYS OF IIII'.ADIATION

Figure 14. Transmutation by--product p roduction versus ersus irradiation time for ^res /O tti - 0.29.

V i

M Him

III

I I'1 I'he effect of Irauucmulatloil on thi , nin • Ildi • Itlfill alld ItWil Oi I I!+ C, shown in Figure 15. vie first sectiun of the abscissa of this figure reprc-

g ents they reduction In the haznrd measure due to the five year irradiation

at = 6.44 x 10 14 n/cm2 sec for the three values of No sig- 0tit ^res4tit* nificant reduction in the hazard measure occurs after irradiation due to the

II29. 15.9 x 106 year half-life of The top curve in this figure shows the

hazard measure of 1 129 with no transmutation after fuel reprocessing. In

the case of 1129 the curves shown represent the total hazard measure since

no by-products with significant radioactivity are made from the initial

loading of 1129

Figures 16 and 17 show the effect of transmutation on ti ►e total actinide

R1111i and RIGI1, respectively. Tile first section of the abscissa of these

figures represents the red► ,ction in ti ► e total actinide hazard measure resulting

from transmutation of ti ► e quantities of americium and curium waste described

in Figures 7 through 12. The three lower curves in Figures 16 and 17 are the } hazard measure of the transmuted actinides (including by-products) plus the

hazard measure of the burned nuclear fuel with 99.5 percent of the U and Pu

removed at the end of the five year irradiation. 11 ► e upper curve in both

figures represents the hazard measure of all actinides (with 99.5 percent of

the U and Pu removed) generated in 60 reactor-years of i.WR operation. The

differences in the shape of these curves represent the varying isotopic compo-

sitions of the irradiated and unirradiated wastes.

Economic Considerations for Transmutation of Radioactive Wastes

The cost of neutron-induceu transmutation of radioactive waste is pri-

marily the cost of producing neutrons. The mayor factors which govern

the cost of neutrons are: reactor fuel and associated costs; plant deprecia-

tion including interest charges; and salaries, taxes, maintenance and general Ill • 908 1V 83IVM 30 f w '1!918 3 nDnN GZ l i

v co 0 0 0 r- r r Ln _ P9 I

I I

" f

7 O •e-I

F ro w U' N a

V S $4 C) U F

A C .r4 ^ 3 I w O U- .4 d O U4 i V) c ^ c ¢ O i r w .I I 4J u

i ,O xv v ro O N ,- ro

_ I W O v 0 4 •-» ^ cc W

O ^ tf•1 ^ H M c2c

► ^n

Q^ t4 :1 w rr a .14 fV x r U p p r ^

nIM IV MTV an w wu 1 N an t vinm ' 1 11 •

A4 in

- N G^

E u v _ ro

w O C 0 •1 a-1 ro a G L G W b N Q' f-' Q 1 CT u C V) 3 A 0 .4 W O F— C Q O vi cr- a.^ Q u w } aQ7

aH bQ) A = c

u t'^ V, T_ u 1 CC 0

W F- } Q b i Ln A O ,^ H ^Q

.d Nv

^ w •„ _ rn 00 v w Q CD C) 0 r r r r

OJZi 1V HIV 30 E ' HH 1 a 30 I N I IJV IVAI - I I^^ t R

Lfi-

U) W Y N to ;S li 1 U 'U LI ro

Q w 0 C 0 y ro u E cn s; ro to E~ LL) 0 t^ s: Q z 3 U 0 .4 N .4 A 0

i Li G ^, I 0 tL Q u r u N c 'L1 d a^ w x r 5 H a a, b

,-J u ,_ vi ao d R W ro O ►e^ o 4 1— ^Y a ac

r .-4

a t.

P. r cri cu n _ O O O O O V

038 1V 831VM 30 W ' IIO I a M I N 113V 1V101 E f' . a , 175 •

operating costs. 71te cost of transmutation would also he a function of plant

output because the sum of the plant costs listed above would remain fairly

constant over a wide range of output, with fuel costa being the ma j or variant.

Plant • itput is determined by: reactor power, operating efficiency, and neu-

tron utilization.

For large-scale transmutation of radiactive wastes in the 19809 and beyond, c

the price of target material will be determined primarily by costs associated r with chemical separation from power reactor fuel residues. Estimates of unit

recovery costs extrapolated to the mid - seventies for actinide targets and

fission products are listed below 120J:

Comparison of Chemical Recovery Costs

Actinide Targets Fission Targets- , 1 17 ti$10 F r AM 241,243 ti$31/f; CS /9 242,244 Cm ,,$81/g Sr 90 ti$16 /g 129 1. ',$10/g

As available quantities increase in the late seventies, and if a need for

recovery of these materials is established, unit recuvery costs could be re-

duced further.

Aside fr^»n developrzntal costs, several economic factors are in favor

of the UF6 gaseous - core reactor for use as a duel - purpose power reactor - trans-

mutation facility. Fabrication of power reactor fuel elvnw nts accounts for

about 40 percent of the t ,)tal fuel cost 121 1 and chemical proc e!;slnf; for ahocct

another 15 percent. Elimination of fuel fabrication costs and reduction in

chemical processing costs could lower fuel costs by 45 to 50 percent of the

UF6 reactor. Increased operating efficiency resulting from continuous opera-

r tion and increased neutron utilization resulting from continuous removal of 1 16

I Isp ion protlrlrt pellnnnn cnniti re • 1111li 1$I lawn 114 , 411 11m 1 • 4111I11 1111111 oil hr Ivill 111r concepts. It r+nlable power lit grilrrat e d In till- veact or the rou111,11r41 11vut i ou cost could be lower still.

Conclusions and Recommendations

This study has attempted to ascertain the :eastbility of nuclenr trzns- mutation in a UF6 gaseous-core reactor. Some specific conclusions were obtained

in the limited time available. Transmutation of 1 129 by a five-year exposure to a thermal neutron flux of 6.44 x 10 14 n/cm2 sec results in nearly order-of- magnitude reductions in the waste inventory of this nuclide. This reduction

in inventory results in an order of magnitude decrease in the hazard potential of fission product waste 1 129 . A five-year transmutation of americium and

curium wastes produces order-of-magnitude decreases in the overall hazard

potential of actinide wastes generated in 60 reactor-years of LWR operation.

The actinide results are in approximate agreement with those obtained by

Claiborne (8). It was found in this study that increased values of the

resonance -to-thermal flux ratio, resulted in increased rates of cores/Oth' 1129 transmutation and increased reductions in hazard potential for both and

the actinides. A rough breakdown of tralramutation costa seems to indicate

that the UE 6 reactor could be competitive with other transmutation reactor con-

cepts.

Any study which attempts to determine the feasibility of a concept generally

raises more questions than it answers. Several questions were raised during

Live course of this study.

Wauld recycling previously transmuted wasters accomplish further decreases in the hazard measure of actinide wastes?

What is the effect of varying the irradiation I time on the inventory or hazard potential of actinide wastes, i.e.. would three ye:vrs irradiation accomplish the same effect as five years irradiation?

1 '1'I^e t rrhn l+luc ++I Indur+vl t rau +um+l .^l I^+n rr+lu l rra 1 nrl hrr xl udy . u^ h•r

long-lived isotopes should be examined for feasibility for Litt! tran++mut at l+m 99 process. One s l ich isotope which may prove fensible Is 2.11 x 10 5 y+ •ar 1'r .

It is present in significant quantities in fission product wastes and has a 22

barn thermal neutron capture cross section and a 92 barn resonance integral.

Further investigation into the process of neutron-induced transmutation may

lead to realization of a practical method for disposal of long-lived radioactive

wastes.

s i I7x

LIST OF REFERENCES

1. M. Steinberg, C. Wolzak and B. Manowitz. Neutron Burning of Long;-Lived Fission Products for Waste Disposal," BNL-13558, Brookhaven National Laboratory (1964).

2. W.C. Wolkenhauer. "The Controlled Thermonuclear Reactor as a Fission Prod ,ict Burner,' BNWI.-1. 695 Battelle Pacific Northwest Laboratories, pp. 33-52 (1972).

3. B.F. Gore and B.R. Leonard, Jr. "Transmutation of Massive Loadings of Cesium-137 in the Blanket of a Controlled Thermonuclear Reactor," Nucl. Sci. Eng., Vol. 53, pp. 312-323 (1974).

4. H.C. Claiborne. "Neutron-Induced Transmutation of Ili gh-Level Radioactive Waste," ORNL-TM-3964, Oak Ridge National Laboratory (1972).

5. C.E. Bemis, Jr., R.E. Druschel, J. Halperin and J.R. Walton: "Thermal- Neutron Capture Cross Section and Resonance Integral for 10.7-Year Krypton-85," Nucl. Sci. Eng., Vol. 47, p. 371 (1972).

6. J.D. Clement and J.R. Williams. "Gas-Core Reactor Technology," Reactor Tech., Vol. 13, p. 266 (1970).

7. H.A. Hassan and J.E. Deese. "Thermodynamic Properties of UF6 at Iligh Temperatures," N.C. State University, Raleigh, N.C. (1973).

8. K. Kikoin et al_. "Experimental Reactor with Gaseous Fissionable Substance (UF 6 )," Soviet J. At. Energy, Vol. 5, p. 1167 (1958).

9. J.H. Lofthouse and J.F. Kunze. "SphericLl Gas-Core Reactor Experiment," NASA-CR-72781, National Aeronautics and Space Administration (1971).

10. J.F. Kunze, J.11. Lofthouse and C.G. Cooper. "Benchmark Gas-Core Critical Experiment," Nucl. Sci. Eng., Vol. 47, pp. 59-65 (1972).

11. R.T. Schneider and M.J. Ohanian. Patent Disclosure, University of Florida (January 1970).

12. E.T. Dugan and N.J. Diaz. "Pulsed-Gaseous-Core Reactor Power Plants," Presented at 1974 ANS Winter Meeting, Washington, D.C., October 27-31, 1974.

13. J.R. Williams and S.V. Shelton. "Gas-Core Reactors for MILD Power Systems," Research on Uranium Plasmas and Their Technological Applicati o ns, NASA-SP- 236, pp. 3430350 (1970). i 1 /'1 _I 14. E.C. Gritton and Ii. Pinkel. "T hi , FeaHlbll lty of the Grlfivoun-Cor y Nuclear Reactor for EAcL tric-Power General Ion," I(M _M l SIT, '1114 . 1(aud Col poi at Ion (1969). - --

15. S. Lawroski. "Survey of Separations ProceF;ses," P_roc_. Se.ro_nd_ U.N. Conf. Geneva, Vol. 4, pp. 575-582 (1958).

16. G. Safonov. "Externally Moderated Reactors," R-316_, The Rand Corporation (1957) .

17. J.R. Lamarsh. Nuclear Reactor Theory. Addison-Wesley Publishing Com- pany, Inc.: Reading, Mass., p. 283 (1966). r 18. H.K. Clark. "Critical Masses of Fissile Transplutonium Isotopes," Trans. Amer. Nucl. Soc., Vol. 12, p. 886.

19. M.J. Bell. "ORIGEN--The ORNL Isotope Generation and Depletion Code," ORNL-4628, Oak Ridge National Laboratory (1973).

20. W.A. Rodger. "Multipurpose Plant for the Recovery of Neptunium and Other Isotopes," report prepared fo- New York State Atomic and Space Development Authority (1967).

21. G.W. Beeman, Nucl. News, Vol. 10, No. 2, p. 23 (1967).

11r t d, Cr /(1 11/ (q7'^ l, `^r7s^/lN^/?^'^, ^izorn DNS T^/^/1lSl^l^% ^!%!1^.^'^ l^fJl(IZI^ Mrtt v^^ 203 ^E U !ice Gw ^

VV 115 t E At '1 C- `.J'Jf: ENiT /'tM ® Ii h. L^4 LC 1

01,{ ^r ^ ,FUEL CYCI^ IzEPROnvciBiLrrY of THE ORIGINAL PAGE IS POOR Sponsor, cd by Nuclear Fuel Cycle Division

1, St rago Fee Analysis for C.,ammal gal active %%;v;tc disposal by nuc • Iear Lr:ulslnutalion was High-Llve Wastes, 11. H. i'r('Jrl (.•ll; C'0. considered. mtlitc(J Several critical f,ascous 111',. asscrublies h:1ec boon built to (ate` and ao advanced prolotype is presenuy Durin( I1) a study Was ronlplc l erl to till AVC to ) under con:;trurtion.' Uperatmg temperatures as high as Iecclnp ;uul Ic, a fnrnlula for d . ornlinn n •lol'a! Ice I 1800 K and neutron 'ilu..w., of Ill"' r,'wiii' sec) :ire con- to hr fu. I ,• I rorec::nr.^ I^u• 1 char;;cci 'c nulct'clal uurlr:u- • sltiored fe;tsIhIc.' Due to )Is Iligh operating temperature. storm(: their' hn: -level %castes to • vI'rltn,rnl rop,,.- t! e gaseous-core reactor would be ml effective potter INvies. The Mora:: fee formula r•: p •dwil,'don lull - f recover. to the gove-nnlcnt for It r 1wroln,,; and r:ll,ilal (;enrralin;, F%'61eIII. %%bile al the sank bite radioactive wastes could be lransamlcci ut the large irradiatuur costs du•ec• liy relate to its 'npo;c'! •, osIt , :t'll,og` volume a% ;oI; blr iii the core of the Ur,, reactor. Several operations. It is :lssunI I that ,i:nularrit.ad (-xilstcrs of . xasles generates lhruu 11 .car : IhIV aril 1,l, plat c•d i oph "Ps :ur a%aiiablr fnr In;I(kill; %taste into the reactor furnl and composition of the a lto a Hctric%•ahlc Su• f;lc Slora:,e V;Irit:l. fir ::,,Inc rlcpencting on the rhenucal r %%(re obtautrd Interim period. pussildv It :fnr; II rn 1 I'MISirrrr •dI„ a %caste. 1' le results ioportrd In this ilaix• by Ilk ing the OI(IGf:N code' in a Way Ltal it could Ile Inure permanent rellos (try. is :u soowd Ilcd a ,I nplOe dto the UF„ gaseous-coregases-core re;wior. Continuous fuel tune storage fee is c arged :it u' Li Ill tlo • Amster 1s adrlilion and renlovol of volatile fission products Was received for stur:l;_e siInitiated in these r:dc'ulaliuus. It Was found 11 t the tee nmsl rmn un I%o c„nrpom'nls: Doc to ill, inherent IoW fuel crass and hi k il nculrnn a) a rnmponen that covers the c:nu ^r's :41:11-v of coo- flux, less t'Aal r;I(Immfive write is ;'. , meratcd by the- UF,; SUuctinl; and peratilig the P 'S"iF 0 Ili- ,toll %%here it Is (;as - cure reactor. Filvrc' I sloWS the buildup of I,:• irm filled tovxpr rd c:rpacily, mill (h) :1 rnn ' po • nl rrealiril an product and actinide waste ill the L'F,; reactor :col in "cadotcmci fluid Whose enrmii!•:< cuvr.•r: I1w rolls of I a :vpwol LIVI{ of equal rowel' output as 'I fa n • nm ,d the lou- ' surveillance at the I(StiF, ( I ll I rtmsilurn q nle. Aller throe %'c;us " ron1,1IM "rs npc:'aUint, 1-140 LIE, costs r constructing the lion • per•nta nenl • I, reactor cunt;ons :shout one order of magnitude less of for G Ise canisters, of transferring: (-.muster:. (oIhal Inn(:-livcc! aetillide. Waste Lit:ul the LWil. repo itory,anrt of decunutiissionuli; the ItS5F.:Ind lit the 1c dual surv'eillmicr costs at lice nr%% rrp"ollolIt Film e 'd slows the rase w1wre tae Ull reactor has 't' s found that the most onp,n'l:ult ,Ir • Irrnulr,nll of Lr been I I I I t I j II%, loaded %%1111 sc%clal lull;- IIve(i radioactive torare fee fs the Cost of till , nueruu fa-'1111%. 'I he ill Writes (ubl,:lned frnut other rc:lut(rs) and oi,rraicd con- most Imilol-tant is till , sprvml b0movii rnNl vsr:il:llion talc tinually for several years al a neutron flux of 6.;4 • 10.4 and c•Ixtuwnnnt fund c:u'nrnhs (titicrr•st) 1%1(". A simplified I n'(cnr sec) 1;!•125 p1W(lh). 50-alit fuel ;;as pressure. ':crsit,n of the formula Is used to dcn prnsll• :dc IL•cne (i %, -` If Joel 1. In all cases the content of %vasir:, h;;s bee:; i Indima. decreased by more Ilan :Ili order of magnitude in three year's. 6 2. Nuclear Wastc Disposal Utilizing a ros- Thus. .hc UF,, gas-ruin • l'r• artur 1s aRracL:ve bolo the vic,epoud that it Las a sit;nific'a;tt1. lover .":inide cous-Core Reactor, N. J'ulr'rno slcr, :11. J. Inventor"; LLan plc :,rnl-rla% I l;;lil - v.:rtr:r rc,'AAW b. 011anian, R. T. ,Sclincidc'r (U of Fla), /I. Timm addiliun, till , dl::int:;al of radu/aeUcc W;IStCh by Inciticru (AA SA) Lr;utsmu(ali m In Ilc UF,. gas-Core reactor ran be a viable :tltc'rnaln'c to the problem of waste dlspos;ll. Public upiniun Insists nn allcrn.de twI loos."III , r Mm'. ! Lr outs presently m-cep(rd a:: Irrluclr:nll; lo:,r.11de, for 1. if. C. CIJ1MOUNE. "Neutron-Inrluvri •I'ran:

l'mao it,Acol.t 1111 ImMill! RiAl

L In/

HA. III I- tell. 4-1 t Fi :. 1. • omparisoi, of radioactive waste production of 3-12:1-mW(m) I.: -ion Otower reaclors.

10 3. \Sefczion of Signilicant Elements and Ra- dionuclklob ior Wuste kiancicj,;rnent Assess- ment. S. ii. Clena A. II han WONAI)

Assys:;Al rill .11 ur, i, IVII I ., k•; I dl " ; I c1l%* v wasw I nallage- ment volia'C l o t1dall Inal. i. of release aad I'll rmlionu- clides i s'! 'It Calctl;;ilmll1he 0IMAN

Colle' illdw ll, 1. I v 2ou rad it III/, N- illes are Pecs.

rut at mu I 1 I I'VIA•WeNSC(i It j60 dayb. A : err, .1 1 4 - mvilit " I sk- 1 1 10 deluraline the 111tost aa$ k - l " S l ivi l l. s .ukl uur litit-l• prvs•ia in

each t i l l it , dail l"1 1- s" tors, durm!.

Several wi,io• • , I . ,, * i . , ;lit I-, (lit , relative hazard of ';4 valmimelit I v., il" ' Iti dilig the .1 1 1k i li'l li

; I ( Ile U0. Ridge

Nalmmli Till:li.^: ,lIlll,.miru. ill 111111h I l l i l l'. I.% It he ral 1.) 4 _1 1 .1 h I.,t l eil^ ^ MW Iv ill ;I it quamliv of I waste 1. 0 U., . , I Z CC .Rile for ;I I- l i t- %%Aer. 14 it for mlialatt-Al. : 1 1 1 d I I . d u it Ila/.; d. resp(-t-lively. A it n • rovi t hq'. (I VI•AIIII., m l - ilwi:. int x is lwolmsed as fill lows:

1. At miv a d.r 014WI-Amil. 1 Iatrlrh I l l IlIgus- 11 0 11 1' 10-11111 ptoduc 1 1. un 'rlls wid the 1114. largest hazard.

2. Al earl I mi, oi- vi all r%ohol- • s Ifi:i I vi lowtiveh;

01mill. l.m , .v' lit, -d 111 A . fi l l' earl . d I I , three

wilit'll :11 • t • 81glillivall 11 lif e i/ ' "I Im I

54 • 1t•4• 11 l l s I ... :,w at 4 ,:wl, 11 1 1 1 1 . .1:4 st i rt , rupresel tatiol l 'l o t- • 1 011A.111 41 1"'. I l t 4 , 1 1 ; 1 . 'I u% . 4,1 ' l , %%1 1 .6 11 6 1w ralige f lI -l i m ,I , 'I Fig. 2. McMusic burnout Its Irradiation little at %*, 6.44 < :,.Cj,li ::l" I., Ic.111-Ill." lair hignifivallve ;Il, 11 - 10" n, (em' sec) in UF,j gaseous-core Mower re- fii•:,t scivillim h1q) lII't;IAl1V V0119 • l1Ii',IlWlIh in U.iht•, l ;wtor. ophill i s I' l' va • 11l l 1; i ll 'st, rourcnlrallous '1 4 rehiki l l." to

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