grouping of hlw in partitioning for b/t (burning and/or transmutation) treatment with neutron...

Pergamon 0306-4549(95)00004-6

Ann. Nucl. Energy ¥ol. 22, No. 11, pp. 697-709, 1995 Copyright © 1995 Elsevier Science Ltd

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G R O U P I N G O F H L W I N P A R T I T I O N I N G F O R B/T ( B U R N I N G A N D / O R T R A N S M U T A T I O N ) T R E A T M E N T

W I T H N E U T R O N R E A C T O R S B A S E D O N T H R E E C R I T E R I A

M U L Y A N T O a n d A S A S H I K I T A M O T O Research Laboratory for Nuclear Reactors, Tokyo Institute of Technology, O-okayama, Meguro-ku,

Tokyo 152, Japan

(Received 31 December 1994)

Abstract A grouping concept of HLW in partitioning for B / r (burning and/or transmutation) treatment by fission reactor was developed in order to improve the disposal in waste management from the safety aspect. The selecting and grouping concept was proposed herein, such as Group MA1 (Np, Am, and unrecovered U and Pu), Group MA2 (Cm, and trace quantity of Cf, etc.), Group A (Tc and I), Group B (Cs and Sr) and Group R (the partitioned remains of HLW), judging from the three criteria for B/T treatment, based on (1) the concept o f the potential risk estimated by the hazard index for long-term tendency based on ALI, (2) the concept o f the relative dose factor related to the adsorbed migration rate transferred through ground water, and (3) the concept o f the decay acceleration factor, the burning and/ or transmutation characteristics for recycle B/T treatment.

ALI A BTF~ [B/T rate] E A h HIALI llmq n I l,, M, M N /CA N,(t) NI N~ nR

R ~ P~ t

#i

Oi

NOMENCLATURE

= annual limit o f intake = total annual input o f MA or LLFP into B/T reactor ( = Z~ A~) = B/T fraction of i, defined by equation (13) = averaged Bfl" rate, defined by equation (12) = neutron energy = fraction of change by neutron absorption of k to i = effective height o f reactor core = ingestion hazard index based on ALI = total inventory of MA or LLFP in n-th cycle = total input of MA or LLFP into B/T reactor ( = Zi Ii) = fraction of change by radioactive disintegration o f j to i = molecular weight o f nuclide i = number of component (MA, FP and U&Pu) = number of component (MA and LLFP) = Avogadro's number = atomic density of MA or LLFP = atomic density of MA or LLFP produced by MA or LLFP = atomic density of MA or LLFP produced by fuel components = recycle number in one reactor life = total production of MA or LLFP ( = Z ~ ) , defined by equation (11) = production rate of MA or LLFP ( = F/T) = total remains of MA or LLFP after n-th cycle ( = x~ RT) = radius of reactor core = time, defined in every cycle period = neutron flux ~ r ) of energy group-g = averaged neutron flux in r and E = radioactive decay constant o f i = decay acceleration factor of i in thermal region, defined by

equation (14) = neutron absorption cross-section of i

697

(Bq/man yr) (g/yr)

(--) (egyr)

(eV) (--) (cm)

(man/GWe) (g)

(egyr) (--)

(g/mol) (--) (--)

(atom/mol) (atomlcm 3) (atorrdcrn 3) (atom/cm 3)

(--) (g) (g)

(egyr) (cm)

(s) (n/cm 2 s) (rgcm 2 s)

(yr -I) (--)

(barn)

698 Mulyanto and Asashi Kitamoto

17

%re

B/T P&B/T R&P U&Pu

= recycle period = reactor life

= burning and/or transmutation = partitioning and burning and/or transmutation = reprocessing and partitioning = unrecovered U and Pu by reprocessing

(yr) (yr)

l . I N T R O D U C T I O N

Many studies in the field of P&T (partitioning and transmutation) of HLW were carried out by several research groups, such as partitioning (Blomeke and Croft, 1982) of long-lived radionuclide in HLW and transmutation (Prunier et al., 1993) by nuclear reactor. However, there was no reasonable concept for partitioning combined with the effective transmutation or the mass balance in B/T (burning and/or transmutation) treatment by fission reactor, but there were only some partitioning models considered by the possibility of chemical separation. Partitioning and grouping of HLW should be done for B/T treatment by neutron reaction. In this study, the terminology "B/T" was used to describe the neutron reaction, because it is not so meaningful to distinguish the nuclear reactions of fission (i.e. burning) and capture (i.e. transmutation) in a reactor. Although "P&B/T" (partitioning, and burning and/or transmutation) treatments of long-lived radionuclide has an important mission for self-completed nuclear system as an acceptable harmonized energy source, the total capital investment required in P&B/T technology should be strictly restricted for the economic reality of nuclear power industry in the future.

In this study, the combined evaluation based on three criteria for B/T treatment was introduced for selecting and grouping of long-lived radioisotope in HLW (Kitamoto and Mulyanto, 1993). The three criteria in partitioning for B/T treatment are related to (1) the concept of the potential risk estimated by the hazard index for long-term tendency based on ALl, (2) the concept of the relative dose factor related to the adsorbed migration rate transferred through ground water, and (3) the concept of the decay acceleration factor, the burning and/or transmutation characteristics for recycle B/T treatment, introduced in this study.

Here, the first and the second criterion are related to the selection of long-lived radioisotope in HLW that has to be removed by P&T treatment. The third criterion is related to the grouping concept for B/T treatment by nuclear reactor. According to P&B/T concept, or formerly P&T concept, the B/T treatment of long-lived radionuclides may be a promising option for improving the geological disposal, if the selection in grouping based on three criteria is appropriate.

The purpose of this study is to find the necessary and sufficient requirements for reasonable selection of radionuclides in P&T treatment, in order to perform the reasonable B/T treatment in a limited space of thermal and/or fast reactors. In this paper, the concept for grouping of HLW in partitioning will be made clear for B/T treatment, and also the closely linked requirements for P&T treatment of long-lived radionuclide, in addition to the basic performance of thermal and fast B/T reactor, will be discussed for advanced waste management.

2. N U M E R I C A L P R O C E D U R E S

In order to evaluate the performance of the recycle P&T system, the B/T rate and the total inventory of minor actinides (MA) and/or long-lived fission products (LLFP) have to be estimated by reactor criticality equation and depletion calculation. In this study, B/T reactor was supposed to be loaded homogeneously with B/T fuel. B/T fuel was fabricated by MA or LLFP blended with enriched 235U or plutonium in order to keep high B/T rate under the critical condition of B/T reactor. The loading fraction of MA or LLFP in B/T fuel was examined from 0.05 to 0.20 in weight fraction. It was considered here that the blending of enriched 235U or plutonium in B/T fuel has an advantage to keep the criticality of reactor with high B/T rate and high B/T capacity for B/T treatment.

In thermal B/T reactor, the criticality, the cycle period and the neutron flux were controlled by the fraction of MA or LLFP in B/T fuel blended with enriched 235U into it. In a fast B/T reactor, Pu was used in B/T fuel for controlling the criticality of B/T reactor, the cycle period and the neutron flux. The specifications of thermal and fast B/T reactor, the evaluation model, the cross section data and the composition et al. are

Grouping of HLW in partitioning for B/T 699

Table 1. Specifications of the model and the nuclear data for the evaluation of the recycle B/T treatment by thermal and fast B/T reactors

Parameters Thermal B/T reactor Fast Bfl" reactor

(1) B/T reactor 3 GWt-PWR, cooled by H20 Core geometry: Ro = 170cm and h = 370cm, respectively, with reflector

(2) Evaluation of performance 1) neutron diffusion model 2) neutron energy 3) cross section 4) neutron flux level

(3) B/T fuel of MA or LLFPs 1) loading pattern

2) isotopic composition

3) materials of B/T fuel (a) for MA (b) for LLFPs (c) blended uranium

(4) TRU chains for depletion calc.

3 GWt, sodium cooled reactor Core geometry: Ro = 185cm and h = 100 era, respectively, with reflector

1) one group 2) thermal energy (~ 600 K) 3) BNL neutron cross-section data 4) approximately 3 x 10 ~3 (n/em2s)

Two regional loading: - i n n e r region for B/T fuel - outer region for normal fuel

1) four groups 2) 10- 2eV-5 MeV 3) processed based on ENDF/B by

shielding factor method or collapsed from ENDF/B-V

4) approximately I x 1015 (n/cm 2 s)

Homogeneous loading in active core, with no blanket

Standard isotopic composition (the same values of the discharged fuel from 3 GWt-LWR, bum-up: 33,000 MWd/Mg (HM), cooing time: 150 days) (a) MA: oxide type (UO2-MAO:) (a) MOX type (U-Pu-MA)O2 with 20% Pu (b) LLFPs: Tc in metal form and iodine in (b) no estimation

Cel3 or 12 (c) natural uranium (c) MA or LLFPs was blended with

enriched "SU to keep the criticality

Isotopic chain reactions for U, Np, Pu, Am, Cm, Cf, Bk and Es, caused by (n, n), (n, 2n), (n, y), etc. and fl-, -,-decay

listed in Table 1. The schematic of the recycle B/T reactor system is shown in Fig. 1. The t ime delay for processing a t R & P (reprocessing and par t i t ioning) process was no t in t roduced here for simplicity. Here,

(1) A ( = EIA~) means the input o f mass of M A or L L F P f rom the discharged fuel, (2) P~ ( = E~ ~ ) means the mass of M A or L L F P produced by fuel c o m p o n e n t in the B/T reactor, (3) R ~ ( = Ei RT) means the mass o f unfissioned M A or unreac ted L L F P o f n- th cycle o f per iod ~.

A, was assumed to be cons t an t for every cycle. ~ means the per iod o f recycle for P&T t rea tment , and is equal to the residential t ime o f M A or L L F P in BFF reactor, such as 1, 2 or 3 years. In the per iod ~, the input masses to the B/T system become zA~. R" and P" are the mass flow f rom B/T reactor to R & P process at the end o f n- th recycle wi th per iod ~.

The total inventory of M A or L L F P in the B/T reac tor in n- th cycle, [INV]n(t), is defined by

[INV]"(t) = ( n R ~ ) ~, N~(t)(M/NA) (i = 1,2 . . . . . N) (1)

where N~(t) means the a tomic densi ty o f M A (Np, Am, Cm, Bk, C f or Es) or L L F P (99Tc, 1~9I, ~35Cs, ~37Cs or 9°Sr) pu t into the B/T fuel. The total a m o u n t o f [INV]~(t) should be limited, due to the limit of core volume for loading under the critical condi t ion. Therefore, the selection o f long-lived radionucl ide tha t will be loaded into a B/T reactor, based on three cri teria men t ioned below, is i m p o r t a n t for keeping this condi t ion.

Now, the behav io r of the a tomic densi ty o f M A or L L F P in B/T reactor, N,(t) for n- th cycle, can be es t imated by M-s imul taneous differential equat ions ,

dN/d t = ]Rj l~2dV i + ~ ~,kf~trkNk -- (2, + trd~,)N~

( i ,j ,k = 1,2 . . . . . M)

(2)

where

N,(0 = ~ ( t ) + N,~(0 (0 ~< r ~< Ro) (3)

N~(t) is the a tomic density of M A or L L F P produced by M A or L L F P loaded in to B/T reactor. N~(t) is the a tomic densi ty of M A or L L F P produced by the fuel c o m p o n e n t in BFF reactor. Time t is defined f rom 0 to


(Bfl" fuel)

R" = El R~'

p n= Zi Pi n

Discharged Fuel

I A= ZiAi

_fl Reprocessing

(HLW) I1-1

R~'I+PI +x A i

Partitioning ] : ~

(LLFPs and MA)

n = Rin-IF n-I li Pi +x As

~ ~ 1 ordinary fuel region

B/T Reactor [ ~ [ "" B/T fuel region

(discharged fuel)

A P R

B/r reactor SLFPs LLFPs

: input amount of IVlA or LLFPs from LWR : amount of MA or LLFPs in discharged fuel from B/T reactor : remain of MA or LLFPs after B/T treatment for x : recycle period

: burning or transmutation reactor : short-lived fission products : long-lived fission products

Fig. 1. The schematic of recycle B/T reactor system, in which the time delay at R&P (reprocessing and partitioning) process was not introduced for simplicity.

in every recycle under consideration. N~ and N~ were assumed constant in those regions, respectively, fi,v in equation (2) is the average neutron flux in space and neutron energy, E,, can be given by the following relation,

1 fro Z Aug f'~ = ~ Jo 2nr , ~ f ' (r) dr (0 <~ r ~ Ro) (4)

where ~zg Au g ~(r) means the summation of Au ~ fig(r) with respect to g. The neutron flux fig(r) can be obtained from a set of the mult igroup neutron diffusion equations for the B/T system under critical condition.

The initial condit ion of N~(0) in n-th cycle can be given by P~, and the final remains at the end of n-th cycle, RT, can be given by N[(z), respectively, such as

N~(O) = ~I(nR~)(MJNA) (5)

R7 = OrR~h)N~(x)(MJNA) (6)


where P~ is the initial input mass of nuclide i in n-th cycle expressed by

P~ = A , (forn = 1) (7)

Pi=RT-I+P~-I+zA~ (forn = 2, 3 . . . . . nR) (8)

where nR = (zl~fdz) means the total recycle number in one reactor life. The total input of all nuclides in the initial n-th cycle can be given by

P = E, B ( = [INV]"(t = 0)) (9)

The total mass of remains in n-th cycle R" can be given by

R" = Z , R7 [=(rrR~)E,N[(z)(M, INA)] ( i = 1,2 . . . . . N) (10)

and the total production rate of mass of MA or LLFP produced by fuel component can be given by the production at the end of n-th cycle, P~ = Ej ~ , and [P] = P~/z,

P" = (nRZoh) E, ( M, INA)N~(z) (11)

Here, the time average B/T rate is defined by the equations,

[B/T rate] = (nR~) E, (M, INA){N~(O) -- N[(z)}& (12)

The time average B/T rate, [B/T rate], and the production rate of mass from the fuel component in a BIT reactor, [P], were defined by equation (11) and equation (12), respectively, as seen in Fig. 1.

Equation (2) for i = 1, 2 . . . . . M are a set of M-simultaneous differential equations in depletion calculation could be obtained numerically by the Runge-Kut ta-Gi l l method. The inventory evaluation for B/T reactor in the recycle P&B/T treatment was calculated for the mass of the grouped HLW in the discharged fuel from 10 units of 1 GWe-LWR at 33,000 MWd/Mg(HM), and 150-day cooling (Benedict et al., 1981). The inventory evaluation for B/T reactor was done for 40 years, because the effect of B/T treatment should be remarkable in one reactor life. The numerical results were obtained by the use of supercomputer operated at the Computer Center, Tokyo Institute of Technology.

3. RESULTS AND DISCUSSION

3.1. Three criteria for selecting and grouping of H L W

In this study, it was thought that the partitioning for Bfr treatment consisted of the concept for selection and the concept for grouping. There was no unified concept for "grouping" of HLW for B/T treatment, but there was only two concepts for "selection" of radionuclide for geological waste disposal, such as (1) the concept of potential risk caused by the hazard index based on MPC or ALI, proposed by Pigford (Benedict et al., 1981), and (2) the concept of relative dose factor related to the mobility or retardation in ground water penetrated through geologic layer (Pigford, 1991). According to the concept for selection in geological waste disposal, 9°Sr, 137Cs and MA, etc. were selected as the risky radionuclides from the first concept, while n9I, 99Tc and ~3~Cs, etc. were selected as the risky radionuclides from the second concept. On the other hand, Pigford (1991) pointed out that the minor actinides were not critical nuclides from the standpoint of geological waste disposal, because they can keep sufficient retardation in the ground water for a long period due to their strong adsorbability to geological layer.

Here, the grouping concept based on the combination of three criteria was introduced in partitioning of HLW in order to group the long-lived radioisotopes for Bfr treatment. Three criteria for grouping in partitioning are essentially important concepts related to (1) the potential risk for long term tendency estimated by the hazard index based on ALI, (2) the relative dose factor related to the mobility or retardation


,9 ° 6

105

103

101

10 - I

10 -3

244Cm 237Np Total "N ~r 243Am \ \ \ "~

- \ . . . . . . . .... : _ _ 3

_~ ~ . _ 3 - - - ~ - - ~ - - - - - . . . . . 2 4 6 c ~ : r ~ .~.__~ Z41Am \ .m

" um ~ _ _ . _ - ~ - - _,._..._-------~ 242- i . _.,-- / Am __=

! /#" ~ . -- ~ . . . . - - - - - 4 e ., \ / ' ~ _ _ _ _ _ _ - - - - - ~ ~ / ~ " X248r- m - - . - . . - - - - ' - - ' - W ~ - - ~ =

, , ~ - 252Cf 250Cf 251Cf 249Cf -~

10 20 30 40

10-5

10-7

Burning time (year)

Fig. 2. The inventory of MA in thermal BfI" reactor with time, with homogeneous loading of B/T fuel of MA (3.0 wt %) blended with enriched U (8.0% 235U), under the conditions of ~t, = 3.0 x 1013 (n/cm 2 s) and the isotopic composition of MA of the discharged fuel of the burn-up, 33,000 MWd/Mg(HM) and

150-day cooling.

in ground water penetrated through geologic layer, and (3) the burning and/or transmutation characteristics for recycle B/T treatment, depending on fission or capture cross section with neutron energy, and the decay acceleration ratio and chain reaction, etc., pointed out in this research.

The criteria (1) and (2) are similar to the concept mentioned above for geological waste disposal, and can be regarded as the criteria for P&T treatment. The criterion (3) is the grouping standard for B/T treatment, that is, a group which is meaningful for B/T treatment, a group which is loaded into thermal B/T reactor, and a group which is loaded into fast B/T reactor. The criteria (1) and (2) are the concepts for selecting of HLW, while the criterion (3) is the concept for grouping of HLW in P&T treatment.

According to the first criterion evaluated by the ingestion hazard index based on ALl, the incentives for reducing the potential risk caused by the disposal of HLW for about 100-200 years are closely linked with the reduction of the hazard index due to MA, 137Cs and 9°Sr in HLW. The radionuclides of MA, ~37Cs and 9°Sr have the incentives for P&T treatment. On the other hand, if the second criterion based on the release scenario of radionuclide from a canister through geologic media is considered, 99Tc, 129I and 135Cs have the potential risk higher than those of actinides (Pigford, 1991). Therefore, the radionuclides of 99Tc, 1291 and "SCs have also the incentives for P&T treatment.

3.2. BIT characteristics of MA

Here, the calculations of recycled B/T treatment for 40 years were carried out under ~th = 3.0 x 1013 (n /

cm 2 s) with the variation of (RslRo) as a parameter. Figure 2 shows the relation between inventory of MA in thermal B/T reactor and time under the conditions of (R,IRo) -- 1.0 (--) and MA loading of 3.0 weight %0 in B/T fuel. The self-shielding effects due to thermal neutron in thermal B/T fuel was not so serious under the condition that the blending fraction of MA was less than 10 and 8% 235U. So, the results could be approximately sufficient for the analysis of B/T treatment. MA of about 253 kg/yr, produced in a year by 10 units of LWR, was supposed to be treated conceptually by the B/T reactor in a year, in which the inventory of MA of about 2460 kg was contained in equilibrium state, as shown in Fig. 2. In this calculation, the reactor criticality, the cycle period and the neutron flux could be controlled by the blending fraction of MA with enriched 235U. The results were obtained for the B/T fuel of 3% MA blended with enriched uranium of 8% 2"U. From Fig. 2, it can be seen that the total inventory of MA in thermal B/T reactor was dominated by 237Np, 244Cm, 246Cm


Table 2. B/T fraction o f M A burned by thermal or fast 13/1" reactor for 40 yr, in which M A of standard isotopic composit ion was assumed to be loaded with the discharged fuel from 10 units o f 1 GWe-LWR in every year

Loaded MA Remain after 40yr (kg)t B/T fraction for 40yr (--):~ Isotope in mass for 40 M A years (kg) Thermal B/T Fast B/T Thermal Fast§

237Np 8.16 x 103 1.3 x 103 4.8 x 103 0.84 0.41 24tAm 5.28 x 102 1.2 x 101 6.4 x 10 z 0.98 -0 .21 242mAre 4.76 x 10 ° 1.4 x 10 ~ 2.5 x 10 ° 0.97 0.48 ~3Am 9.92 x 102 3.3 x 102 1.0 x 103 0.66 -0 .01 243Cm 0.0 1.6 x 10 -1 3.0 x 10 -I - - - - ~ C m 3,64 x 102 5.3 x 102 1.5 x 102 - - 0.60 245Cm 2.22 x lO ~ 3.5 x lO ° 1.1 x lO -~ 0.84 0.96 246Cm 0.0 3.0 x 10 ~ 2.3 x 10 ° - - - - 247Cm 0.0 1.9 x 10 -j <10 ̀ 9 - - - - 24aCm 0.0 1.4 x 10 -I <10 -9 - - - - z~gCm 0.0 1.4 X 10 9 <10-9 __ __ 249Cf 0.0 1.0 x 10 4 <10-9 __ __ 25°Cf 0.0 1.5 × 10 4 <10-9 - - __ 2~Cf 0.0 1.2 x 10 -4 <10 -9 - - --- 232Cf 0.0 5.0 × 10 -4 <10 -9 - - - -

Total 1.01 × 104 2.46 x 103 6.69 x 103 0.76 0.34

t Remain indicated the mass o f inventory at the final stage of one reactor life. B/T fraction is defined by equation (13).

§ Minus sign of B/T fraction means the accumulated mass of M A , Notes: Characteristics of thermal B/T reactor were calculated under M A fraction of 3.0%, ~sU of 8.0%, (R,/Ro) = 1.0 (--) and ~ =

3.0 x 10 ~3 (n/era 2 s). Characteristics o f fast B/T reactor were calculated under M A fraction of 8.0%, with homogeneous loading and ~f = 1.0 x 10 ~s (n/cm2 s).

and 243Am. The burning of MA was effective in case the inventory changes just like sharp cut teeth of saw or shows high achievement of B/T fraction as seen in Table 2, while the burning of MA was not effective in case of smooth variation or of low value of B/T fraction. It can be understood that Np and Am could be burned in a steady state by thermal BT reactor, but Z~Cm, 246Cm, 248Cm and other isotopes of Cm and Cf, etc. could be burned with difficulty by thermal B/T reactor, so their inventory for the latter case increased continuously.

Figure 3 shows similar results of MA burned by fast B/T reactor. The calculation of recycle B/T treatment for 40 years was carried out under fir = 1.0 x 1015 ( n / c n l 2 s) with (P~IRo) = 1.0 ( - - ) and MA loading of 8.0 weight % in B/T fuel. The self-shielding effects due to fast neutron did not influence so much the characteristics of fast B/T reactor as compared with the characteristics of thermal B/T reactor, under the condition of blending fraction of MA less than 10%o. So, the results could be approximately sufficient for the analysis of B/T treatments. From Fig. 3, it can be seen that the inventory of all isotopes of Cm and higher mass of MA did not accumulate in fast B/T reactor.

In Table 2, it was seen that Np and Am could be burned for 40 years by thermal B/T reactor more than 80%, and higher than by fast BFI" reactor. Here, the B/T fraction for 40 years is expressed by BTF, which is defined by

BTFI = [~,if~A~- R,(t = l:life)l/'ClifeA i (13)

The build up chain reactions for isotopes of U and TRU, i.e. U, Np, Pu, Am, Cm, Bk, Cf and Es, were taken into account in the computation for 40 years. On the other hand, the isotope of Cm could be burned effectively by fast B/T reactor and the heavier atomic mass than Cm, such as Bk, Cf, etc. did not accumulate in fast B/T reactor. According to those results, it can be seen that the characteristics of neutron reaction in thermal region were quite different from those in fast region. So, it is important to separate MA into two groups, such as Group MA1 of Np, Am and Group MA2 of Cm (and the higher mass of MA).

3.3. BIT characteristics of LLFP

The effectiveness of transmutation of LLFP by capture reaction could be understood by the decay acceleration factor #i, which can be defined by the ratio of total decay rate under neutron flux ~b to the spontaneous decay rate,


o

106

105

104

103

102

101

10 ° ~f f

10 -1 ~ "

10 .2 _--- m

10_ 3 -- ___-- 10 -4 "'

0

243A m 2 3 7 ~ Total

" .

~- ,-.~--~--~__.__._..-_--~ ~X,~--~4 I ~ - ~ ' - - 245Cm \ Am

r - - - " - - ' ~ ' 242Am 246C m 243C m ~ _

I I I - 10 20 30 40

Burning time (year)

Fig. 3. The inventory of MA in fast B/T reactor with time, with homogeneous loading of B/T fuel of MA (8.0 wt %), under the conditions of ~f = 1.0 x 10 ~5 (rdcm 2 s) and the isotopic composition of MA of

the discharged fuel of the burn-up, 33,000 MWd/Mg(HM) and 150-day cooling.

/z, = (2,. + a,ff)12i (14)

/zl approaches to unity when ad,<<21 or for radionuclide with small trj or under the lower neutron flux 9b. If #~ is approximately one, the decay acceleration cannot effectively act on the radionuclide. /t~ for 137Cs and 9°Sr are nearly equal to one even under high neutron flux 9~, so it was considered that the decay acceleration by neutron irradiation cannot be effective and the transmutation is meaningless, if the decay acceleration of nuclide i,/~, was not far larger than one. On the other hand, 99Tc, 1291 o r 135Cs have a high value of/zg in the thermal region, so the transmutation can be effectively reasonable by thermal B/T reactor.

The inventory of 99Tc, 129I, 135Cs, 137Cs and 9°Sr in transmutation by thermal B/T reactor is shown in Fig. 4. These results were calculated for thermal B/T reactor under ffth = 3.0 X 1013 (rdcm 2 s), in which B/T fuel was loaded under the condition of LLFP of 5.0 weight % blended with enriched uranium of 5.0% 235U. Here, the evaluation was carried out under the standard composition of LLFP, equal to the discharged fuel of burn-up at 33,000 MWd/Mg(HM) and cooled for 150 days. In these results, the reactor criticality and the neutron flux were adjusted for the recycle period of 3 years. Now, the member of Group A' was assumed to be 99Tc, 1291 and 135Cs, and the member of Group [A" + B'] was assumed to be 99Tc, 1291, 135Cs, 137Cs and 9°Sr. After 40 years, B/T fraction of 137Cs and 9°Sr were only 0.19 and 0.27, respectively, as seen in Table 3. On the other hand, B/T fraction of 129I, 99"I'c and J35Cs were 0.68, 0.57 and 0.25, respectively. In the case of Group A', the total B/T fraction was 0.53 at the end of 40 years. Total B/T fraction of Group [A' + B'I decreases to 0.41, if Group A' is mixed with 137Cs and 9°Sr. Therefore, it is better to exclude t37Cs and 9°Sr from Group [A" + B'], and also J37Cs and 9°Sr could not be burned effectively by thermal neutron reactor.

Since 135Cs and 137C5 were included in different groups, therefore the isotope separation must be done for separation. Unfortunately, it was very difficult to separate the intermediate mass number, i.e. 135Cs, from the radiocesium of four components, i.e. 133Cs, 134Cs, 135Cs and 137Cs. On the other hand, the presence of highly active isotope of ~37Cs in radiocesium makes the isotope separation of 135Cs almost impossible. Therefore, it is realistic to exclude 135Cs from Group A'.

3.4. Grouping in partitioning

According to the evaluation described in Sections 3.2 and 3.3, three characteristics can be derived as follows: (1) MA could be generally burned by thermal B/T reactor, but Cm was burned effectively by fast B/

Grouping of HLW in partitioning for B/I" 705

10 5

10 4

"0 ~o

10 3 ..9,0

..o

~o 10 2

I I I

(4)90Sr (2)137Cs ( l )To ta l

_

- - , - . . d .--,___2 " - ' - - ° J

(3)99T C (5)13SCs (6)1291

~ ( 6 ) (5)

l 0 t~ ,, , I I I 0 10 20 30 40

Irradiation time (year)

Fig. 4. The inventory of 99Tc, 129I, 13SCs, mCs and 9°Sr in transmutation by thermal B/l" reactor, under ~th = 3.0 × l013 (n/cm 2 s), under the condition of LLFP (5.0 wt %) blended with enriched U (5.0% 23sU), the recycle period 3 yr, and the isotopic composition of MA of the discharged fuel of the burn-up,

33,000 MWd/Mg(HM) and 150-day cooling.

Table 3. B/T fraction of LLFP of Group A' and Group [A' + B'] in thermal reactor for 40 years

Loaded LLFPs mass Isotope in LLFPs for 40 years (kg) Remains (kg) B/T fraction ( - - )

ml 2.78 x 103 0.9 x 103 0.68 9~Tc 1.15 x 104 5.0 x 103 0.57 mCs 3.86 x 103 2.9 x 103 0.25 J37Cs 1.60 × 104 1.3 x 104 0.19 9°Sr 8.04 x 103 5.9 x 103 0.27

Total (A' + B') 4.22 x 104 2.5 x 104 0.41 Total A' 1.81 x 104 8.5 x I(P 0.53

Note: Characteristics of thermal Bf r reactor were calculated under LLFPs fraction of 5.0%, 2"U of 5.0% with homogenous loading and ~b = 3.0 x 1013 (rdcm e s).

T reactor, and the heavier mass of MA, such as Bk and Cf, etc., did not accumulate nor were produced so much in fast B/T reactor, therefore (i) Np, Am and minor component of U&Pu can be burned and/or transmuted by thermal B/T reactor (the group was named Group MA1), and (ii) the isotope of Cm could be burned effectively by fast B/T reactor, and heavier mass of MA did not accumulate in fast B/I" reactor (the group was named Group MA2), (2) 99Tc and 129I was transmuted by thermal Bfr reactor (the group was named Group A). On the other hand, (3) it was difficult and meaningless to transmute 9°St and/or ~37Cs in thermal Bfr reactor and, of course, fast Bfr reactor, too, so they should be treated by the other application such as beta or gamma ray source in industrial scale, if possible (the group was named Group B).

Based on these characteristics and the presupposition of no application of isotope separation process, the grouping of HLW in partitioning could be proposed as follows:

(1) Group MAI: Np, Am and U&Pu* (2) Group MA2: Cm (and other heavier MA)


1 GWe-LWR

~ l Reproeessing ~ UandPu )

t ( H,W )

( 1-129,rc-99,CS-13~ ) ( MA, C$-t37, Sr-90 ) | I

II [ I ,o,.,-u- II I I

*) B/I" characteristics; lxi, off, oai, gi, Ni, etc.

' Remain

Glass I Solidification

Interim Storage I

Fig. 5. Conceptual grouping of HLW in partitioning based on three criteria for B/T treatment.

(3) Group A: Tc and I (4) Group B: Sr and Cs (5) Group R: all of the remains in FPs.

U&Pu*: minor fraction of U and Pu in HLW, unrecovered by reprocessing.

The conceptual grouping of HLW in partitioning based on three criteria for B/T treatment is shown in Fig. 5.

3.5. Remains of the partitioning

Several trends of HIAu variation for the discharged fuel of burn-up at 33,000 MWd/Mg(HM) and for the fractional remains assumed in the partitioning till 108 years are shown in Fig. 6. The keys for these trends and the fractional parts are listed in Table 4. A trend marked by (a) and others marked by (b), (c), (d) and (e) correspond to the partitioned remains of HLW removed by five different recoveries. In this study HIALJ for the partitioned remains of HLW were compared with HIAL~ of uranium ore, the processed mill tailing (pmt) and the standard mill tailing (smt).

Now, HIAu of uranium ore in Fig. 6 was assumed to be equivalent to HIALI of the total mass of natural uranium ore of 2.83% U3Os, which can produce 3.5% 2~5U enriched uranium fuel consumed in a year by


1014

1013 I

1012

1011

lolO

109

108

107

106

105 100 101 102 103 104 105 106 107 108

Storage time (year)

Fig. 6. Variation of HIAu for MA, LLFP and the partitioned remains of HLW with storage time, which was compared with HIAL~ of the processed mill tailing (pmt), HIAu of the standard mill tailing (smt),

HIAu of the natural uranium ore due to the remained crude mill tailing.

Table 4. The keys for the partitioning fractional parts, t where (a), (b), (c). (d) and (e) correspond to the partitioned remains of HLW removed by different recoveries

Partitioning U and Pu Group MA Group A Group B Group R Key§

Discharged fuel 1 1 1 1 1 (a) Mode 1~: 5 x 10 -3 1 1 1 1 (b) Mode 2 10 .-4 10 -4 1 1 1 (c) Mode 3 10 -4 10 -4 10 -4 1 1 (d) Mode 4 10 -4 10 -4 10 -4 10 4 1 (e)

Remarks: Fractional part means the ratio of partitioned remains of HLW to the grouped component in discharged fuel.

:~ Recovery of U and Pu in the reprocessing was assumed to be 99.5%. § Key corresponds to the mark in Fig. 6.

1 GWe-LWR (Benedict et al., 1981). Natural uranium ore of 2.83% UaOs contains 298 Bq/g (from 238U), 314 Bq/g (from 226Ra), 348 Bq/g (from 2~Th), 291 Bq/g (from zt°Pb) (Dreesen et al., 1982).

HIALI of the processed mill tailing (pmt) in Fig. 6 was equivalent to HIAu of the remained crude mill tailing including 226Ra, after uranium metal that is equivalent to 3.5% enriched uranium fuel consumed in a year by 1 GWe-LWR was recovered from the total mass of natural uranium ore. The processed mill tailing (pmt) contains 1.77 Bq/g (from 23sU), 58.4 Bq/g (from 23°Th) and 66.0 Bq/g (from 226Ra).

On the other hand, HIALt o f the standard mill tailing (smt) in Fig. 6 was equivalent to HIAu of the purified mill tailing, removed z3su, 23°Th and 226Ra chemically. The standard mill tailing (smt) was derived from the water contaminat ion standards related to the mill tailing disposal, and it contains 226Ra of 0.115 x 10 -3 Bq/ g (5.0 pCi/l) and 238U of 0.23 x 10 -3 Bq/g (0.03 mg/l) (Dreesen et al., 1982).

Due to a small amount o f ~26Ra, the processed mill tailing (pmt) and the standard mill tailing (smt) were thought to be the better standards than uranium ore. Toxicity o f uranium ore could not be a suitable criterion as an acceptable risk of radionuclides (Pigford, 1991), because the substantial toxicity of uranium ore is 226Ra which is not quite similar to the property o f MA-isotopes. Some of the measurements of 2~rRa in ground


H-3 C-14 Kr

t t t

solid ~ T W&SII~ /

discharged (dissolution in PUREX fuel Nitric acid)

n~ elmh

O.Tcl TRUEX

t ~Am-On

CURE I precipimim I (Amine separaao.) [ I~AmO2(CO3)3 ]

Notes: Q grouping products

low-level w a s t e

t

Fig. 7. Schematic flow sheet of the chemical separation processes for grouping in partitioning, with the modified flow sheets of PUREX process, TRUEX process and CURE process, and double carbonate

precipitation.

water contaminated by uranium ore show that human use of contaminated water could result in radiation doses far exceeding the level established for the public (Pigford, 1991).

Figure 6 shows the relation between typical trends of HIAu and time, with the change of the mode of partitioning. The partitioning by Mode 4 will reduce HIAu of the partitioned remains of HLW below HIAu of the processed mill tailing (pint) or uranium ore after about 10-100 years. If HIALI of the standard mill tailing (smt) is selected as a criterion of radioactive safety for biosphere, the partitioning by Mode 4 will be able to reduce HIAu for the partitioned remains of HLW below this criterion after about 300 years. If the partitioning by Mode 4 may be developed in the future together with B/T treatment, the interim storage may be meaningful for the disposal of HLW.

3.6. Separation method

In the practical meaning, the separation method to realize the grouping concept mentioned above could be made by the implementation of the combination of the modified flow sheets of PUREX process, TRUEX process (Schulz and Horwitz, 1988), CURE process (Westinghouse, 1990) and double carbonate precipitation, KsAmO2(CO3)3 (Groh et al., 1965). Figure 7 shows schematic flow of chemical separation processes proposed by this study for grouping of HLW in partitioning.

CONCLUSIONS

(1) The three criteria in partitioning for B/T treatment are related to (i) the concept of the potential risk estimated by the hazard index for long-term tendency based on ALI, (ii) the concept of the relative dose factor related to the adsorbed migration rate transferred through ground water, and (iii) the concept of the decay acceleration factor, the burning and/or transmutation characteristics for recycle B/T treatment. (i) and (ii) are related to the criteria for geological disposal, but (iii) is related to the criterion for B/T treatment.

(2) Grouping of HLW in partitioning was closely linked with the transmutation method and the scenario

Grouping of HLW in partitioning for BtT 709

of geologic disposal. According to the grouping concept based on three criteria, the long-lived radionuclides can be grouped, such as (a) Group MA1 of Np, Am and unrecovered U and Pu, (b) Group MA2 of Cm (and other heavier MA), (c) Group A of Tc and I, (d) Group B of Sr and Cs, and (e) Group R of all of the remains in FPs.

(3) Grouping of HLW proposed by this study conceptually can be realized by the implementation of combined processes of PUREX, TRUEX, CURE and double carbonate precipitation.

REFERENCES

Benedict M., Pigford T. H. and Levi H. (1981) Nuclear Chemical Engineering, 2nd ed. McGraw-Hill. Blomeke J. O. and Croft A. G. (1982) Nucl. Tech. 56, 361. Dreesen et al. (1982) IAEA-SM-262/56. Groh H. J. et al. (1965) Nucl. Applic. 1, 327. Kitarnoto A. and Mulyanto (1993) Proc. o f GLOBAL'93, 1027; Proc. o f SAFEWASTE "93, 286. Pigford T. H. (1991) Trans. A NS 81, 63. Prunier C. and Salvatores M. et al. (1993) Proc. o f SAFEWASTE '93, 299. Schulz W. W. and Horwitz E. P. (1988) Sep. Sci. Technol. 23(12 & 13), 1191. Westinghouse Hanford (1990)CURE: Clean Use o f Reactor Energy. WHC-EP-0268 WH.

grouping of hlw in partitioning for b/t (burning and/or transmutation) treatment with neutron...

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