IT8347900A1 - PROCEDURE FOR THE RECOVERY OF PLUTONIUM FROM AQUEOUS NITRIC ACID SOLUTIONS - Google Patents

Description

DOCUMENTATION

BOUND

DESCRIPTION . wygujBa

? in support of a patent application _for_ invention _ _ having as _title loj _

''Procedure for the recovery of plutonium from aqueous solutions of nitric oxide <n >_ _

on behalf of: EUROPEAN ATOMIC ENERGY COMMUNITY

(EURATOM)

o. ci? ?* SUMMARY 3 c<-Procedure for the recovery of pi u ton io CD a j from aqueous solutions_ of nitric acid by extraction- _ :

?5<" >jliquid-liquid reaction of such solutions with a solution of tributyl phosphate in an organic solvent immiscible with water, in which organic and aqueous solutions are formed as waste streams ^Freed for the most part from plutonium, charac- _

| ?

_ jrized by the fact that these waste streams are still treated with oxalic acid

DESCRIPTION

[ _ The present invention relates to a

i

Improvement of a process for the recovery of plutonium from aqueous solutions of nitric acid by liquid extraction of such solutions with a solution of tributyl phosphate in an organic solvent immiscible with water. Usually this solvent is kerosene. The process is known

to under the ^ex_ignation of process Pur ex^

The invention contains the combination of

certain processes partially known for the

regular post-treatment of waste streams

which are formed in the Purex process, to give a global process fully integrated with this latest one.

The additional procedural stages? necessary

O-sari, which present a maximum measure of identity and consequently of compatibility with the process _ <?>

or Purex? allow you to use it optimally '

sS in 'finally a series of parameters of the .Process

oS disengaging! from the obligation to limit the losses of fissioned material to low values_

or required by the current "open" procedural principle cr* from the waste side. Losses of fissioned material are practically eliminated, by means of these post-treatment processes, in the current

of liquid waste, for which the formation of

Medium-active waste emitting alpha radiation is drastically reduced, while highly active waste is split into two fractions.

first fraction contains heat-generating fission products with negligibly low transuranic actinide content (defined as free from

Malfa'! )__while the second fraction contains the deeds->

! long-lived nests together with the rare earths_as a chemically_homogeneous group of elements._

The industrial reprocessing of burned nuclear fuels occurs through the process

Purex, in which uranium and plutonium with tributyl phosphate (TBP) dissolved in kerosene are further

continuously extracted in a largely selective manner in a countercurrent process from the starting solution in nitric acid also containing the fission products (SP) and are subsequently separated from each other by reduction of the Pu to:

<"? It was trivalent, insoluble in the organic phase.

The separate currents of uranyl nitrate,

the ? O' of Pu are subjected to further cycles of CT extraction

tion with TBp/kerosene to a high purification before their conversion into the corresponding oxides. As an inevitable consequence of the Purex process, waste of multiple physical and chemical nature is generated in the preparation of these final products.

chemical as well as with variable product contents

of fission and ?ctinides. Since these wastes, in particular the ?ctinides contained in them, represent not only an economic loss (Fu) but also

; a source of danger, special handling and treatment procedures are necessary to be able to

to exclude _so darkly, even in the_long term a diffusion in the biosphere which in turn means an additional burden on the economic viability of the ci- _

of nuclear fuel. In practice? _ci? _sign i

fig the need to keep as low as possible. p_o p

how much waste there is in general? and the costs?

<?>^r ?. are you talking about fission material? in particular P u . t/j _ To achieve this goal? neces- a OS

O

it is possible to reach compromises in the choice of appropriate process parameters? which may be more or less distant from the optimal solutions for

individual problems? since their specific requirements often go against the optimisation

tion .

For the case that in this case _p r o c e d ira eh t o ai

for t a on the waste side still come super ?i ...

What limit values are considered tolerable in the wastewater treatment plants? Is recycling necessary?

c 1 addition of the e_ comments of waste not corresponding to the- .

the_ norms? usually and _imp tying the same unit of !

,l

<?? >.. j

exercise? being_ inevitably linked to such!

I use an interruption of the normal progress; _

of production. The consequences are a reduced ?i-|

availability of the system and as a final consequence a

greater production of waste, as well as the need to install redundant systems in particularly critical or large storage capacity points. ! _ .lung_cotie .between the individual stages of the process.

_l._Highly Active Waste (HAW) _

JSel__ treatment_..of.the_refining_co.rr ent_!?.

-to_Jbr_ac i do....n i t i co del_ fimo_ciclo_di_estrazione

.(_-HAW_)__with_formic_acid. ...bols ent e_ per.er_ 1 a. degr adazion.e_r idu t iv a._dell_!_e.c c.e s so__di . nitric acid, __si__ arrives o_ l/J in the presence of ps s alic acid at a _p_r.e_c.ipit ation e.

oel ipraticamente q u an t i tativa (99-99 ?.9.7?). _d_e_ll_a_.fr action.

of pr? do t ti _di_f iss ione._( SP ).-..eh irai cam ent e...orno gene a_ _

<? >? [of rare earth elements ( SE) and yttrium as well as of the tran-O actinides

j suranic homologous _(An: Np,__Pu? Araj Ora? etc. . . ). The_ easily sedimentable crystalline precipitate contains as its main impurity one of the Ba and

<?>of Sr from HAW? in elements such as Zr ( Nb ) _ and_. Mo ? _

! . j <? >.

.that other with two pno_ to the formation of .precipita-. _ ^_ <:>gelatinous precipitates that are difficult to separate?

mind in the presence of low concentrations of acid? _

<? >? <? >. . i ? . . . . remain in this case as corapies of oxalate in the above aqueous phase (see table 1). In contrast to post-treatment extraction processes, with the precipitation process, a qualitative separation of all insoluble and non-ionic (polymer) forms of Polymers is also achieved, from the cor4

r waste_entity_._ _ _ __ 1 : j.

The optimal separation factors of SPs

remaining from the SE and the An are made to concentrate-<'>- ;- - r

tractions of the SE below 0? 1 M, i.e. in the trat-

? |

direct treatment of the HAW or at most after its pre-concentration for 2-3 times? for which it is also

the risk of spontaneous formation of unwanted precipitates of another nature has been largely excluded.

_ _ Reductive denitration of HAW with aci-_

!i

c?. &o _formic also has the consequence that the Pu is- <? depolymerized in a polymeric condition? i.e. not O..

There is ionic: and is it consequently not extractable? and that up to

80% of nitric acid is degraded to give the inert, non-condensable laughing gas (N 0).

2. Medium-Life Waste (MAW)_ _j_ _ r

Refined streams in nitric acid! <~>

-CT* i <' " . >- T!

coming from the purification cycles of U and Pu i

(<.>=MA<.>W) are<..>concentrated and rectified<'>according to?l.o T|

state of the art on nitric acid and its reconduction

tion_of_water_as_chemical_process_products^ The

concentrate free of inert salts with the residual quantities of SP_ and the process losses of U _ and Pu vie-. . <? >. <? " >. <? ? >?<'>

added to the HAW, i.e. the An,_ essentially Pu

J

and Np, are separated by the same .stage as j

pr e c p az on e .. e .

According to the state of the art, the extraction medium TBP/keratosine, after the re-extraction of the product 11 i Pu_ e d U, is subjected to a regeneration process by layering with alkaline solutions of NaOH/Na CO , to remove the retained 2" J ' ? <. >co SP and Ali as well as degradation products, <?>2? or mainly dibutyl phosphate (HDBP). The currents- OA OC

aqueous residues _ containing salt that are formed

r-j what do they contribute to? following the presence of the

-CD non-volatile Na ions, sensitive to ZZ formation.

atc of the quantity of io-active waste OQ CD

By experimental tests it has been ensured according to the present invention that only oxalic acid ions (for example 0.8 M) are capable of

? remove in one of its washing phases more than 99_^9$ i ' , ? ^of the retained portions of Pu and U, even from the irradiated extraction medium (v 250 watt/l).

From the SPs, which contribute substantially to the pollution of the extraction medium, Zr (Nb) and Mo are quantitatively detected, while Ru (Rh) and Cs are quantitatively detected in quantities of 95% and 90%, respectively.

The HDBP, which thus remains in the organic phase, can be quantitatively removed in a second conventional alkaline washing stage, where

the waste stream containing sai and which is formed

.what?. you might come back. _ inde t erminate _c.ircp.st an z e

in the weakly active category (LAW) and in any ca

so_can? be conducted as a waste without "alpha", ___

|

_ _____ _ By oxidative degradation of IIP a-_ _

oxalic acid in the waste stream generated in the -- - ? - - | ? ? cL the first washing phase? a salt-free aqueous solution is obtained again? whose concentrate with _ o* * the SP and An can be added to the HAW^ before _the j_

<" ' >. <.>. . <">i

^?? its treatment. The laying takes place more appropriately: _ "?4

3. by introducing the oxalic acid washing solution into the bottom phase of the MAW solvent evaporator, since oxalic acid is degraded quantitatively by nitric acid at higher concentrations and temperatures.

I 3. Treatment of waste currents "caused by operation <11 >(=Management) (IL1W)

A non-negligible portion of the waste

|_medium activity containing inert salts?_ which_ are_ formed-I - - ?

j but in the reprocessing plants?, returns to ppe-\

'ad-hoc' rations not from Reatine? Like for example ^

i . . ?

the rinsing and decontamination processes. This

Is ILLW characterized by high uncertainty? both for

I

as regards its formation over time both for<' >;

as regards its composition. It must therefore be assumed that it contains for example strong

/

complex-forming compounds such as citric acid and tartrate co-or EPTA. Its presence, for example, makes it impossible to successfully apply traditional procedures for the separation of radioactive products. Both the traditional co-precipitation process

that the English flocculation process is ineffective in solutions of the composition indicated in Table 2? as indicated by specific experimental analyses. On the contrary, it has been possible to demonstrate that by using a precipitation technique of the OXAL type, even in the presence of such complex formers, the

fraction of An pu? be separated with superior yields_

et al with the result that the content of active- CO <. .. ' ~ >|

t? alpha in the ILLW has been reduced to a non-si- z c's5 value

significant .

To carry out the precipitation process, in this case it is necessary to add to the waste stream not only oxalic acid but also a certain amount of SE<? >to produce the chemical carrier of the precipitate OXAL. The reduction of the excess free acid can take place here by slowing down the process by adding an alkaline solution such as NaOIL.

In the absence of complex formers, a setting of the LLLW current is sufficient.

with concentrations of SE (represented by Ce) and acid

?3 ~2

oxalic acid of 5 x 10 and 2t 5 x 10 M respectively

and \xn subsequent neutralization at pH 1. - -A pretreatment of this type of ILLW?

or CX does not affect in any way the possibility of subsequently sub-' -e?-ex rre the solution so and to a

-3 r-4 further co-precipitation type treatment for

O

the separation of the remaining beta-gamma activity. _

ex In the presence of complex formers? the _C?5.

CO

precipitation must take place at pHs lower than ??? ,

!to exclude their effectiveness. With a solution

. <? ' " >? - <' >j

of IL1W typ i c ? as indicated in table 2j ?_a Pu accuracy of 99 ? 5% me? has been achieved?

-2

during 1* setting^ of the solution^ to 5.. x_1 0_ jn

-1

IF and 2? 5 x of oxalic acid ? i.e. al de-j

cupio of the first values. It has been demonstrated that it is not?

then it is practically more necessary to add smooth-;

<.. >. <" . ">T<?? >-?

ways for the partial neutralization of the excess of

nitric acid. As a consequent manifestation are ]

jstati_ also realized better factors_ of separation!

{? ?

tion of impurities that form simultaneously.

j .

|4 . Method for including other types of waste

The previously mentioned possibility of separating transuranic actinides in high yield in a simple manner even from solutions containing complex formers can help to facilitate the decision of which reagents should be indicated.

for the decontamination of solid waste of the most varied kind, such as those formed in the reprocessing

or irradiated nuclear fuels and especially in the

-cr reprocessing of isolated Pu. The need for methods _

! efficient for the treatment of such materials contaminated by Pu will be particularly accentuated <'>

r-J intense mind in the future, when the rfconduction of the ?

t . ; CD jPu in the nuclear center and especially _in reat-- !

rapid self-fertilizing crops will have reached 1st industrial maturity.

\ * Another process for dissolving ceramic PU oxide consists of a process of oxidative leaching of solid waste (for example, . ..

I <? ? " ? >- <? ? >i- <? ? >?pious filter material from the manufacture of _ _ .fuel elements* incineration residues

from combustible waste^ residues from the solution of combustible elements)_meaning a mixture of

!

!nitric acid_and anionic Ce nitrate complexes. The Ce nitrate solution containing An which ?

4+

jis it formed like this? (after jreduction of Ce<4 >to Ce _)__could be taken into consideration in any mo-

i

mento_ as_ a setting solution for the treatment-'^

ILLW's chin.

5. Diaphragm of the processing cycle in the post-treatment!

waste disposal; integration with the diagram of the

Purex production cycle

The post-treatment processes mentioned in point 3 and the Purex process may be affected.

integrated, in the manner represented in simple form in figure 1? in a global process?_ eh e |

can be fully combined with the technological state *

<' ' ' . ? >i

current state of the art of reworking and treatment

j waste? and brings clear advantages for both ._gl_i or areas. ' H CQ

The overall process is based on the conventional Purex process steps with only the va- .Ij

plans to use oxalic acid as a solution of 1

travel in the first stage, of washings -<iei_me <?>

zzi _d <">H la i_.i_<">

extraction . ? - ' |

. --? ? ? ; <' "? >. j

The "process-specific" waste streams and respectively these 11 e_ ^e-specific-<'' >; <' >, <' ' >-I

"service" are converted by the OXAL precipitation process described in waste streams

highly active "alpha-free" (FP) or medium-active

. <' ' >. . <? ? ? >. i

(ILLW) ? that after sedimentation of the precipitate cry??

{

stallino can be removed from the pre--i cipitation reactor and separated from the solid substance residues

dragged. The precipitate? Sometimes after washing

coarse? is subsequently decomposed (for j

- - as much as possible in the same reactor? to avoid the

<? . . >? CO needs to transfer highly active solid substances- ^

\ j1 ^ tive) with nitric acid at concentrations and tempera- o<.

11 00 upper tures, possibly with the addition of pe- 1

! ? hydrogen oxide, by oxidative degradation ^

j ?? of the oxalates? and the solution is separated from the !

solid residues after setting the 1i ^2*

i or? 1 concentration of nitric acid. i

j By means of an "exhaustive" extraction!<1 >type !

11 Purex with TBP at $30 in kerosene, the actinides Pu ?

| High-value Np are extracted together with select-j

? sively from the decomposed solution and subsequently

- ! separated. The product stream containing. Pu vie-^ .

! traced back to the Purex process, better for example the

in the Pu product from the separation of U-Pu (in a re-

"precedent" partition), while the means of extraction!

i - -- =_ consumed tion can be subjected to a washing together with that of the Purex co-decontamination cycle. <v >|

! joint with extraction medium. The separate Np current is normally fed to the ultra- ? treatment.

further expected for "alpha" waste, but it could pa-'

<'>j<?' >

?

The Irimenti are good to serve as a starting product for a j

possible commercial use of Np,

The absence of the SPs Zr(Nb) and Ru which distur-!

ibano more? in the Purex scheme excludes a? .i.nfluence' i.

!negative on the degree of action dx this extraction "e-!

isaur lente" despite the low saturation of solve Tn-Ite. Regardless of this, the expected scheme is

oO ]to conduct process currents beyond the pernette

O

to tolerate mutual impurities (for example Pu in C t ?

the current of Np and vice versa, or An in the medium

? of extraction), without compromising the overall objective of the operation.

Due to the simultaneous partial precipitation of Sr(Y) and Ba, the aqueous raffinate stream _JSL_ ' from the exhaustive extraction is still chemically inhomogeneous and the thermal performance of the SP mixture is relatively high (see curve number 2 in figure 2).

By separating the alkaline earth SPs, on the contrary, a chemically homogeneous solution of An and SE would be obtained, whose thermal performance after starting the activity of Ce/Pr-144 is

:an order of magnitude smaller than that of the original HAW (see curve number 3 in figure 2)? both factors by which possible

lit?_for _noticeable improvements in conditioning!

! and the final storage of long-lived <,,>alpha" waste.

By setting the ionic concentration by adding concentrated aluminum nitrate solution with acid deficiency and sodium hydroxide to 0.5-0.6 M and 1-1.2 M respectively (salting effect) and wherein simultaneously the free nitric acid is reduced to approximately 0.2 M,

or the An and SE can be selectively extracted with the Purex extraction medium (30% TBP in kerosene) and consequently can be separated from the impurities carried over by the OXAL precipitation

?

in the form of other SPs. For their re-extraction?;

r-. sufficient dilute nitric acid. The raffinate with <' >^ [the other SPs and the inert salts can? without further Ss be reunited with the above solution of SP of the preci-? ? - - ;

t OXAL pitation? and for example be conducted to: vitrification, given that due to the limited volume

3+

of this refined product the inert ions Al and Na are tr<1>

they are still below the quantities necessary for the production of glass.

In this case too the means of extraction

: consumed can be sent to solvent washing-<. . . "" . ' >. <. >j <' >te of the co-decontamination cycle. Together with the cycle.

"exhaustive" extraction.corresponding to. 20-25. and -the flow rate. _in. .comparison.. cycle conditions.__Pur ex.

According to the current law, the P.urex process is an "open" process as far as rx-

I look_at_them. losses, of .material _ di_f.ission._ne.lle _cor-_r.ent i^di_jri fiu t.i.._La_l imi t action ...d.i_que s t e__pe.r dit e ai

y alo ri ..1 imi t e__economie ameni e_ac ce 11 ab il e . avyiene.._from _

cL .lat.o lla_progett action ^ell_!_plant.to__me-__L ?A , _..prog.e.tto__ne (

j

of an t e_l a_l imi ta zione .de 1 degree of s ur a t ion.e_d.el ; 3S or. solvent with heavy metal ions (U Pu) less or

of 70# of the theoretical value and before 1 ' in s t allat ion 2S o3 equipped with capacity? of e xtraction_suppl emen t ar y . j _ ...

1

_ From the point of view of the exercise we are looking for. 3

QO

through instrumental and analytical surveillance _

intensive process of recognizing the _as soon as possible and correcting deviations from standard values

<; ? >i

If this? Despite this? in exercises not in orderj _ they are formed. ..cari eh and with .perd ite __ in amiable i ssibili of

| fission material _ (mainly Pu ) , these _ <?>

;Do they go up to the mind about the c1 at e_ in the 1 1st an it? of p :P-and s know > ci? which has as a consequence an interruption of the routine exercise and therefore a reduction... of- .

_1a available? What is the capacity of the J plant.

?: The low saturation of the solvent and the over-_ !sizing of the extraction capacities _ with

lead to a greater extractability of some SPs, in

(particularly Zr/Nb-9 5 and Ru as well as contact times

i? lung <'>

hi <~>

d <.>

in the middle <.>

d <.>

i <.>

extraction <..>

with the <.>

and <.>

_ sol <.>

u <. >j p z_ions I_ <? ~ ~ " . ' ' . . ' . " . >? highly active and therefore to an increase in irradiation damage of the extraction medium. The main degradation products of THP, dibutyl phosphate<' >and monobutyl phosphate (HDHP and H MEP) form with Zr

<^ >? ? insoluble precipitates (cruds) in ?- OL -extraction apparatus

With the increase in residues this phenomenon ; _ has made it necessary to drastically lengthen the pe- I_

? . i J radioactive riodes for nuclear fuel irradia- j

I

diato (from J_50 to_ 180_giom i as 1 normal original level, _ ?

(coginal at least 3 years old).

While this fact may still be acceptable for the current light water supply line, it represents a serious burden for the future concept.

i

to the fast self-breeding reactors. j

| <' >Hot cell tests on combusted fortemeiite material have demonstrated _that the "crud?

? i j can? be practically suppressed even_with short te<'>m-. !

! more cooling _and consequently becomes controllable, if_ the loading of the extraction medium with

' !

I

heavy metal ions in the highly active section

I

Purex scheme extraction is increased from the current <' >standard values below 70$ adr at least

$90 of theoretical value

To counteract the escalating danger of greater losses of Pu, however, it should be kept

available in the current concept of the <M>open process " several extraction stages, whereby following the

; consequent increase in the irradiation of the medium

extraction the desired advantages would be lost again.

in a "closed" concept of the type of the post-treatment procedure described with recovery of

Pu, in the Purex process phases, however, Pu losses higher than those are perfectly acceptable

? The currently tolerable, and consequently high degrees of saturation can even be combined

re-.<? >;to a reduction in the currently usual sizing of the extraction section (number of stages or length of the columns).

i This concept of bleeding (limit-I .... . <' >?? i

<?>to) ? naturally also applicable to the pu-I cycles

advanced fission material refining <1>

Decoupling from process parameters

whose optimal conditions are opposed to one another

to the other, made possible by the introduction of the

post-treatment process of the type described here,

provides a number of advantages for the Purex process

of exercise such as:. _ _ _

_ - major decontamination factors per process stage?, _

_ - least reported amount of extraction medium/least amount of fission material? _

- drastic reduction in the need to interrupt normal operations for the purpose of recycling ? already <.>p., of _"waste_charges" <.>? i.e. <" . >in _last <.>a_ analysis 7;^ <?r<?>

- greater availability of the <1>system. s From the construction point of view, this gives rise to the possibility of setting up the system in a more

: <" .>. <. ' .>. <. . ? ' >r t??

i

| o9 compact the design^of the <1 >plant for a de- _ii_._- . desired net annual flow rate? by: I i ? <" >. . . . . .. <'>[

I

_ _ - reduction in the number of stages and

? then ! shortening of the columns<.>or <' >batteries<'>in the<.>extraction section? _ _ : _ ^ _ _ _ _

_ ___ - _reduction of capacity reserves due to_shorter operating times? and therefore for example.?_ _

.- smaller diameter, of the extraction columns

i _ - reduction of lung capacity necessary-I <? >' i pi between the individual stages of the procedure . . _ _ J.

I |

The facts mentioned lastly facilitate the technical safety planning for

the control of criticality for surveillance,

of the flow and inventory of fission material

Reduction of radiation damage by means of e-

extraction and the formation of "crud" that is created

in parallel with _it^ _realizable reduction of media -

te 1 <1 >e 1 high solvent saturation with simultaneous

in and limitation of the extraction stages? opens

! or 11 r and there? the possibility of reducing the radioactive waste of irradiated nuclear fuel again an-O ??, which in the case of high residues? without having to employ- :

We are dealing with highly corrosive reagents such as fluorine ions <?>

?-4 as complex formers to suppress the O-form

tion of precipitates, or without having to wait for technological advances in technical extraction devices j- * c??o i <' ? _ >;

!usually very delicate of the type of central contactors.

refugee for a drastic reduction of the rd- load!

diation of the extraction medium. I

While the integrated post-treatment de-

written above of the rejection currents increases the

: degrees of freedom for the optimization of a series of

i !

process parameters of the procedure Pur ex tole-

causing greater temporary losses of Pu? leads to;

And I?

'ultimate analysis to a drastic reduction of e_ per-.; _

! <? >i

'Pu's words that must be considered definitive.

As regards the intermediate-active waste, the following are obtained first by washing the extraction medium with oxalic acid and then by oxidizing the oxidized waste streams:

<. . ' . . >i <. >containing inert salts practically .."alpha-free"* _ so the potential danger to the environment that.. _ derives from this category_ of waste is two orders of magnitude at least from the temporal point of view. In this way the problem of long-term stability is simultaneously resolved as a point-<? >. . . <" " >. <"" ' >. . <""" >i Q.

i O? to__critical in the <1 >use of ?cement_or bitumen ? usual j _

1 ? or according to the current state of the art} which matrices or a. ci for the consolidation of this waste. _

What about JlAW? They contribute

?? sensitive to risk in the long term and therefore;

<?>; <. ' . >i<"' >to the necessary final shelf life in addition to

aa Pu also contains other transuranic actinides^ _

\

in it Np y Am_e<' >Cm, while the heat production is determined by the SPs (see figure _2). JL a _ sud-j

; i d ' iv "isi "on .e of the HAW " " i .n two . fr .az . i .on - i} gl * .i SP . and gl <.>i A..n .# . ? .

I <' >? ?

! ?<' >so again deemed desirable to disac- ' explode the parameters that hinder each other _

!'for a more efficient final conditioning and storage J?

? <? ' ' ? . >[ <" >imitated and more flexible than these two groups. ] j <' >? <. >i <.>

i Due to its chemical behavior as

_ - _ 'plexus , the Np occurs in different proportions ! ! . <' ? " >. j<" >_ jin all the specific primary waste streams. ; _ of the process, .of the Purex process. The_ process . <" " ? ?>. <? >. <'>. . i .

_ of.. post-treatment described, however, allows its _ _ quantitative separation in practically pure form _ in a single product stream . The two super-actinides Am and Cm L..behave analogously to the '.... _ I Idi SE group chemically related in its type ,_ and can consequently be isolated with it from the _HAW_ as f r a-.

i \ _ homogeneous tion at the end of the post-treatment procedure^ _ _ v _ _ _ I | _ The p r q p q r tion_. in p e s q.. of the SE in the SP .

_ <!>corresponds to about 1/3,_ while the An are negligible. The heat production of_ this group corresponds, after a radioactive life of one year, to about 30 fo of the total HAW and_ is mostly due to Ce/Pr- 144? After _10 years practically_ only the An remain as a heat source. i 1 The use of the rare earth Gd as a poison

i ? i j for neutrons dissolved homogeneously for _the control : ? ! <" " .. . >? <.. " . >of the criticality? in the solution medium (of increasing _ ! , ?Importance with 1 <T >increase in size of the plants), i <? >t i would no longer represent a burden for the conditioning of the SPs in a global scheme of the type described here. Its presence in the HAW would on the contrary even be advantageous to favor the control of criticality in the 0XA1 precipitation process as well as in the formation of homogeneous crystalline matrices and

(unlike glass)_ thermodynamically stable for the consolidation of the An ._

As for the other SPs, it also appears more convenient to incorporate them in glass Respect-I ci.

G?

to the current state of the art, they also result

c for this fraction advantages through the post-treatment procedure exi HE5 Ct<" >ment Thus? through the denitration ex reducing treatment with formic acid the volatility of Ru in the calcination or in the? 3*:

or direct vitrification, which significantly reduces the exhaust gas treatment effort.

By <1>the absence of the Ans furthermore the debated question of the long-term stability (at least 1000 years) of the glass matrix becomes important, and the composition can be oriented if necessary mainly towards the immobilization of the most important SPs such as Cs and Sr. <' >! 6. Summary of the inventive features 1'

1?

. i The novelty of the present invention concerns the integration of the procedural steps previously in use

I

unknown parts, described in point 3, namely;

I

- washing of the extraction medium in the first stage with oxalic acid solutions? _ j ..

Yes

- denitration and OXAL precipitation

. . . .

of the specific waste currents of the Pu<1>-jrex? process, i.e. of the JHAW? to which previously we were:

the concentrate of the refined products of the cycles of

T <.>i<.>fication of <" >Ui <. >e Pu e del ..primq_s <.>stadio d <">el <.>_lavag I <'>pur -<' >!g io de 1 _ me z zq d i e s t razione ? _ _

- the treatment of waste streams;

,me_dio- active of the te cn ic a of and service (container! t i

ci <?>inert salts and compound-formers ss i)_by means of a precipitation process of the <? >type <">OXA <_>L <.>, _ _ ? i _ c - post-treatment of the precipitation product for the recovery of the fission material Can _ .. r-<1 >and for the isolation of the remaining transactinides, together with the rare earths as fission products? _ ... _ i ?ti - . ? ? CO

In the. -- pr ?o ?ce - s -s*o -- P ? u - re .x i -used industrially for the so-called J

reprocessing of irradiated nuclear fuels according to the current state of the art. !

r. - . . - . . <' ? ' >. . <' " >?

? !

j_ _ _ The technology required for the execution

f

]

-Idi these additional procedural stages and realizes;

[bile _ according to the current state of industrial experience and always refers to solutions which are identical to the techniques already applied in the Purex process

i

'same, both from a chemical and a physical point of view

ideal equipment.

raze to this en r on to a size

- ? *' high compatibility between all procedural stages, so that all the advantages offered by recycling techniques can be fully exploited.

The resulting benefits can be summarized in abbreviated form as follows:

GL

- possibility of reprocessing after shorter radioactive periods even for com-1| c? material

/

- :- highly combusted nuclear fuel (of vital importance for future self-ferring reactor technology)

O

rapid users)?

- widening of the range of limit values permitted for the individual process parameters<? >c and therefore <' >j

- reduction of requirements - to process management and monitoring systems?

1

j - reduction of noCl charge formation

! ! compliant with the regulations and the interruptions resulting from continuous operation? and therefore

- lower need for capacity reserves?

!

do you have lung reserves? and as a consequence of this

' 1 j 1 j - reduction of the inventory of fission material? ? |

?

- increased load factor and possibility of: |

i

. !1

.

! i

__ - a more compact design of the overall system? _

- reduction of the extraction medium flow due to high saturation? then <?>. .. f - reduction of the design dimensions !j of the extraction equipment (e.g. smaller diameters and lengths of the columns) and

- shorter contact times from the extraction medium to the highly active extraction part?

- reduced radiation damage to the mining medium

- improved decontamination and separation factors per process stage and - suppression of crude oil formation - virtually quantitative recovery of Pu ? , j (also non-ionic forms) from all waste streams of the reprocessing plant - and thus

- reduction of waste containing "alpha"? !especially in the medium-active sector through "non!

i e g>enerazione" (solvent washing with oxalic acid) or by post-treatment and con-?

<: >consequence , ]

? I I

i - reduction of the long-term danger potential associated with it

consider as "alpha-free" with products

?

of heat-generating fission and an "alpha" fraction chemically homogeneous with rare earths

- -- reduction of Ru volatility in vitrification,

- extension of the degrees of freedom for the conditioning and final storage "according to -?t ci. u? misura" of the first fraction *

- ditto for the "alpha" fraction,

- possibility of isolating in a practical form-I

I ! pure and quantitatively Np as a product of the , maximum lifetime of transuranic actinides, l ? for any forms of treatment and storage -s1a cO final even more? specific.

TABLE 1

Content of some important radio-elements in the above solution after OXAL precipitation of

' Authentic HAW, obtained from nuclear fuel with a residue of 33,000 MWd/t. j element U Pu No Am/Cm Ba/Sr Ru Zr/No SE <?>+ Y

! . !

?dissolved r^99?7 r\**0.1 ?/?? 1 r>^0.1 r^lO /v97 ?>-99 />^-0?5

(continued table 2) - ?

1

? j

i<

! ?

TABLE 2

iComposition of a "ILLW ica rejection current"

jConmosto/ Mass Concentration

!element -- M - - - - 1- kg/t fuel

?HNO-<'>? ' i -|???03 4 340 510

]<SE>7?7<~~ >2, 2E-1 <?>3677<" >55

NBCI - -2? 5E--1 -15 - 22.5 - .. ....?^1..

cn <? >IN? SO 1, 8E- 1 25.7 38.5

<j >? ?- -4

H01 1 , 0E- 2 0.4 0.54 oc CD

Al <?>872E-3 0723 0.35 <~ >a Ca- -3.8E-2. : - - 2.3 -Cr _1.5_E- 3 _ 0.08 0, 12

.._sa iCu 2, 4E-3 0, 15 0.23 or Fe <'>4, 6 E- 2 S ?K -2 , 7E- 3 - .0,008 0, 1 2 00

CD

J? 1E-2 _ __0.75 1.1

jMn 1 , 0E- 2 0.6 | 0.9

<">*Mo 470E-3 07 38 <; >i t o, 57<'>

there. -1,4E~3 ? <? >-0.08 -?]-0.12

jFu_ L_ _ 1 , 5E-3 0, 15_ _| 0.23

| Zn 2.3E-3 0.15 I 0.23

: Zr <. ? >8, 8 E- 4 <" >0.08 <' >! 0, 12

}

^Na.-oxal ato? --3, FRI- 2 - - 5.0 - . the? 7.5--i

Ja Tartrate _ 2, 2E-2 _ 5.0 7.5

|Na citrate 1.9E-2 5.0 7.5

NaF <">T73E-2 <~ >1.0 <' >T, 5

?EDTA ??? ?2, VE- 3- - - 1,0-i

.TBP __ 7, 5E-4 0, 2 ! 0,3

"T

<?>HDBP 9 , 5E-5 0.02 0.03 .

3, 2E-5 <" >0.005 <' " >0.007

1 Kerosene . 0.02 _

: Can <? ?>1 , 0E- 4 0.024 4,OE-2

./?

x

Claims (3)

_S ?a 1 . Process for the recovery of plutonium from aqueous solutions of nitric acid by OQ liquid-liquid extraction of such solutions with a solution of tributyl phosphate in an organic solvent immiscible with water, in which organic and aqueous solutions form as waste streams liberated for the most part by plutonium, characterized in that these waste streams are further treated with oxalic acid 2. Process according to claim ?1 , characterized in that the organic type stream is continuously washed in countercurrent with a 0.5-1molar aqueous solution of oxalic acid 3. Process according to claim 1, characterized in that aqueous waste streams containing lanthanides are treated with a-<' >oxalic acid? and lanthanides precipitating as oxalate- you with simultaneous precipitation of actinides transuranic elements present. Rome, \ k, MAR.1983 LX ?? pp.: EUROPEAN ATOMIC ENERGY COMMUNITY (EURATC <)>M)_ _ _ \?~fter if and for the others ?*- x i mTfOY/Y/?r <' >TC/g$.-8928 ** a ?? . r?*.; -o ?NJ you -C - fg cd _ <' >/?F* <? >CU thesis? c? /} you V<4>* -??. ? >S> ?7 /O <Nv>^ajt?sV^ Ir . ; - - . - ---- - . ----- - - - -