LU84016A1 - METHOD FOR THE RECOVERY OF PLUTONIUM FROM NITROUS ACID SOLUTIONS - Google Patents

Description

ï / Ii'J ilû î »EURGPAEI3GHB ATOMGEtëEIHSCHAFT (EURATOM)

Paternanael dnng

Seducer, recovery of plutonium from nitric acid. -aqueous solutions j

Inventor; None Ι; · /; 0Ιί30ΚΑΚ

! Brian A. HURT

Francis M0U3TY

* Francesco MAKNOKE

iae Erfiiidang Lexrifeine improvement of a method sur Rüchrewin - :. uuî of plutonium from nitric acidic aqueous solutions by liquid i'Jüsaig extraction of such solutions with a solution of tributy] phospha; into a. '; i, :: iit i.asser ment: r.iser. t aren ccvahi scner lecture notes · ueDui · .- ·. he - be rui'feAVfejt’j 'nren he .η · n l.

_ o _ jr r

The invention includes the combination of certain, in part, known processes for the regular treatment of the waste streams which occur in the Purexvr ferry for an overall process which is fully integrated with the latter.

Me required additional procedural steps, never a maximum. of Icentit & t and thus compatibility with the Purex process, allow optimal use of a pinch of frozen parameters? . through their decoupling from the Swang, the spalti material process

Losses due to the current "open" waste | ". principle to limit the low verte. There will be | through these cachet treatment processes, the pu fission material losses in the; final liquid waste streams are virtually eliminated, making the | The amount of alpha-active medium-active waste is drastically reduced, while the highly active waste is divided into 2 fractions. The first contains the heat-generating fission products with neglected low transuranactinide content (defined as "alpha" -free) and the second contains the long-lived actinides together with the rare earths as a chemically homogeneous group of elements.

The industrial v’i rather inherited! Burnt-out nuclear fuels are processed using the Purex process, whereby uranium and plutonium with tributylphosph (TBP) dissolved in kerosene are initially largely selectively extracted from the nitric acid starting solution, which also contains the fission products (SPj), in a countercurrent process and then separated from one another become i-valued by reducing the Pu, in the organic «

Phase insoluble, Pu.

The separate urar-yl and Pu-kitrate streams are subjected to a cleaning in further extract !, oracyclia with fBp / nerosir, before they are converted into the corresponding oxides. As an inevitable consequence of the Purex process in the preparation of these end products, waste from a wide variety of physical and chemical cultures is lost. he. Held on fission profiles and actinides. Since this waste and, in addition, the actinides contained in it, in addition to an economic loss (Pu), also represent a source of danger, special treatment and treatment processes are required to ensure that they are spread in the long term

sure 5.U £ 3 .’Γ. * ^ $ 3 6L SU ΚΟΓ_Γ.βΓχ | > »2-3 Vr2 0QSr * USL 0 2 Γ> 0 Σ Λ 2CVTi £ ÜS.1 StSVilXjj * C r T

* _ J _

Economics -’ea nuclear fuel cycle represents. In practice "this means the need to keep the amount of waste in general and the fission material losses, especially Pu, as small as possible.

In order to achieve this goal it is necessary to make compromises in the selection of the suitable free sparex, see below, which may be more or less far from the optimal solutions for the eir-selbpro! Eme! ', Since their specific requirements are often opposed to each other for optimization.

In the event that this ** open ** process towards the waste side nevertheless exceeds the acceptable limit values for fissile material losses, it is usually necessary to recycle the non-standard waste streams using the same operating unit, which inevitably involves an interruption of the normal production process The consequences are reduced plant availability and, ultimately, increased production of waste, as well as the need to install redundant systems at particularly critical points or large Larer capacities as a buffer between the individual process steps. 1 -C lu - & - * still active waste (HAW) After the treatment of the nitric acid raffinate strc-ms of the first Exir-u ·: - • xi nrscycle (= HAK; with boiling formic acid the reductive degradation of the excess nitric acid occurs in the presence of oxalic acid to a practical quantitative! _qq _ gpfo ^) Precipitation of the chem ish homogeneous spa product- (3?) - rare earth fraction fSS) and yttrium and -sr homologous Tr an sur an — actinides (An: No, Pu, Am, Cm, etc). I »he easily settled crystalline precipitate contains a file of Ba and Sr from the HAN as main impurity, while elements such as Zr fNb) and Ec, which are otherwise particularly low acid concerate! Or.er. lead to the formation of gelatinous, difficult to separate precipitates, in this case as oxalate complexes remain dissolved in the supernatant aqueous phase (egg i '& c. ij. Contrary to extractive Î * eight treatment procedures \ * - 4 - i \ eî " You can use a cem P ai] la gsver process to quantitatively separate all insoluble and nient ionic (t.olvr.eren) Pu forms from the waste.

The best separation actuators for the remaining 3P from the 3E and An. "achieved at concentrations of SE below 0.1 K, i.e. either with direct treatment atu HAW or maor. after its 2-3 times pre-concentration, which largely eliminates the risk of spontaneous formation of unwanted strains of other types of ketus.

Lie reductive lenitration ces HAW with formic acid also has the consequence that Pa in the polymer, i.e. non-ionic and therefore not extractable. State is depolymerized and that up to 80fo of the nitric acid is broken down to the inert and non-condensable laughing gas (N ^ O).

2. Hi-active Abi'al 1 i HAW}

The nitric acid ftaffinate streams from the unification cycles of U and Pu ^ = HAW) are according to the state of the art for nitric acid and water recirculation as rrczessehemikaiien aucononsentriert and rectified. Las inertsul zi'reie concentrate with the. Fixed margin to SP and the process loser. he. U and Pu is added to the HAW, i.e. the An, irr: essential Pu and Dp are separated by the same OXAL precipitation step of the HAW.

according to the prior art, the extraction agent TBP / fcerosin is extrautirr after the return; the products 1¾ and D are subjected to a pegeneration process by washer, with alkaline solutions from KaOE / ^ FagCO ^, in order to remove zurüokgehalter.e SP and An and degradation products, primarily bibutyl phosphate HLBF), cu. The resulting saline water drills. AbiallctrSme trarer. · "Due to the presence or not of T. i - rorer weser.tiict to /. / Î 'car ntte-iantiver A /. · N ..... _e.

Lurch experizerteile 1. :: tersuchunren has been successfully checked that Co. al saurai öszrgen (e.g. 0, · - * Q ir. Your location smc, do sir.-r.

· - · - r '- 5 -' eïruriger. Remove> 99.9 $ of the residual Pu and U portions of-of the irradiated extractant (· ~ · 2 ^ 0 Katt / l).

From the SP, which contribute significantly to the contamination of the extractant, Zr (Kb) and Mo are washed back quantitatively, Ru (Rh) and Cs to $ 95 and $ 90 ι · respectively.

HDBP, which remains in the organic phase, can be removed quantitatively in a second traditional alkaline washing cycle, whereby the resulting saline waste stream could possibly fall under the weakly active category (LAW) and in any case be classified as "alpha" free waste can be.

By oxidative degradation of the oxalic acid in the waste stream generated in the first wash cycle, a salt-free aqueous solution is again obtained, the concentrate of which should be added to the 3P and An der HAK prior to its treatment. This is done most sweetly by feeding the oxalic acid solution into the bottom phase of the evaporator of the MAK raffinates, since oxalic acid is broken down quantitatively by nitric acid at higher concentrations and temperatures.

- * Treatment of "operational" waste management (ILLK)

The not insignificant proportion of the inert saline medium active waste that falls in gravel processing plants is due to non-1 routine "aö-hoc" operations, e.g. Flushing and decor, minimization processes. This ILLK is characterized by a high degree of uncertainty, both in terms of its timing and in terms of its composition. However, Kan has to assume that he e.g.

svrke contains complexing agents such as citric and tartaric acid or SDTA.

Your presence makes e.g. a successful application of traditional procedures for the separation of radioactive products is impossible. Both the French Cc-precipitation method and the English I Ploc method are ineffective at readings with the I composition shown in Table 2 like own experimental investigators. he gave. In contrast, inclined -erttn knew that with the help of a j. Fall technique even if there are such Eouplexbiidner the t »_ _

A practice with yields> 9¾ can be separated using the result | nis to have the content of alpha activity in the ILLW at a non-significant value.

i i

To carry out the precipitation process in this case it is necessary to add a certain amount of SE to the waste stream in addition to oxalic acid to produce the aacrochemical carrier ces OXAL precipitation. Reducing the free. Excess acid can be simplified here by adding an alkali such as KaOH.

-

In the absence of complexing agents, one setting of the ILLW

Current on SE (represented by Ce) and oxalic acid concentrations of -3 -2 5 x 10 or 2.5 x 10 M and a subsequent neutralization to pH values around 1.

Pretreatment of this type of ILLW in no way influences the possibility of subsequently subjecting the protruding solution to further treatment of the Go-Fresipitstion type in order to separate off remaining beta-gamma activities.

S

J

f If the complex bioners are present, they must be filled in at pK values below 0 to exclude their effectiveness. With a | typical ILLW solution, as given in Table 2, an I 9 °, 5 / -pu-precipitate was achieved by adjusting the solution to 5 5 x 10 * "M 3n and 2.5 x 10 ^ V. Oxalic acid, ie to 10 times the first values. It has been shown that it is then practically no longer necessary to add alkalis to partially neutralize the excess nitric acid. As a result, there were also better separation factors Impurities achieved from initial 1 ends.

i ···· · :: -? - - 7 -

Nuclear fuels and especially in the further processing of the Pa obtained. The need for efficient methods for the treatment of such Pu-contaminated materials will be particularly accentuated in the future when the Pu return in nuclear power plants and especially in fast breeders will achieve industrial soap.

Another method of solving ceramic Pu-Gxid consists in a. oxidative leaching flags of solid wastes (e.g. filter material from the manufacture of fuel assemblies, ashing residues from combustible waste, residues from the dissolution of the fuel assemblies) using a mixture of nitric acid and anionic Ce ^ + hitrato complexes. The resulting An-containing Ge-Kitrat solution (after reduction of Ce ^ + to Ge "') could be used at any time as an all-in-one solution for the ILLW treatment.

5th issue of waste treatment: internment with the Purex—

- -> "~ '1 nm ^ · Μ ·· ι— · ιιι · ιιιιι m I II ———— I · ι · ΐ4ΐ --------------- -,, N r Γ Ί Γ IIIII

Ess11esschema

The aftertreatment processes listed under point 3 and the Purex process can be integrated in the simplified manner shown in Fig. 1 to form an overall process that is fully compatible with the current technological status of refurbishment and waste treatment technology and provides clear advantages for both areas.

The overall process is based on the conventional steps of the Purex process with the only variant of using oxalic acid as a washing solution in the 1st stage of extractant washing.

The "procedural specifiscben" or the "company-specific" waste streams are converted into "alpha-free" highly active (FP) or moderately active (iLLW) waste streams using the OXAL precipitation process described, which according to Absexzer.i assen ses crystalline creditor scarfs vas iex Pälireafcxor attracted. and before. mi f led. Pest material residues can be re-separated, from precipitation, possibly after rough washing, then (if possible, the same reactor, in order to avoid the need for higher concentrations and temperatures, possibly with the addition of Z before · - "· 3- '. - 8 - v - Ô -

Wasserstol’l i-üroxvd disrupted by oxidative degradation -of the oxalates and the disruption was separated from solid matter residues after adjusting the nitric acid concentration.

Kittels of a Purex-like "exhaustive" extraction with 3Q / o TBF in kerosene will then be of higher quality. Akiirdden Pa and Kp are selectively extracted together from the digested solution and then separated. The product syrup containing Pu is returned to the Purex process. into the Pu product from the U-Pu separation (with "earlier" partition), while the used extraction agent can be subjected to a common extraction agent wash together with that of the Purex code coniasination cycle. The stripped lîp stream is normally fed to the further treatment intended for "alpha" wastes, but could just as well be the starting product for a possible! serve commercial use of the Kp.

j The absence of the most disturbing SP Zr (Kb) and I Ku in the Purex scheme indicates that the efficiency of this "exhaustive" extraction is impaired despite the low solvent saturation. Independently allowed ·; it accepts the scheme of mutual impurities (e.g. Pu in the Kp stream and vice versa, or Ar. in the extractant) intended for the weather management of the process streams, without compromising the overall objective of the operation.

Due to the partial i-litrecipitation of 3r (X) and Ba, the aqueous baffinate stream resulting from the exhaustive extraction is still chemise :, inhomogeneous and the thermal performance of the SP mixture is relatively still (see curve Kr. 2 in Fig. 2 ).

r By separating off the alkaline earth SP, on the other hand, a chemically homogeneous solution of An and SE would be obtained, the thermal performance of which would be * Abizlin. the Ce / pr-344 activity ca one order of magnitude rzedrirer i \ P_ r C 3 X; _k- '- "· v'i h _' .uTt 6 it Σ '» Î Γ. λ. DD »~ j} also files with which there are considerable improvements for the conditioning and storage of the long-lived "Alpha" waste from the SS.

20 Pf - 9 -. Ci 'y

The kin concentration of the ion concentration by adding an acid-deficient concentrated aluminum soil solution and sodium hydroxide solution to 0.5 - 0.6 M or 1 - 1.1 K (salting-out effect), at the same time reducing the free salting stones to approx. 0.2 M. the An and SE selectively = extracted with the pure extracting agent (30 / .. IBP in kerosene) and thus separated from the impurities from the OXAL precipitation in the form of arid 3P. For their back extraction, it is sufficient • '- dilute nitric acid. The raffinate with the other SP and the inert salts can easily be combined with the transferring SP solution of the OXAL precipitation and e.g. be added to the glazing, because due to the restricted volume of this raffinate, the Al "* and Na insertions are still below the amounts required for glass production anyway.

In this pall, too, the used extractant can be added to the solvent wash of the code contamination cycle. Together with that of the "exhaustive" extraction cycle, it corresponds to 20-25 / t of throughput compared to the Purex cycle.

According to the current state of the art, the Purex process is an "open" process with regard to the loss of fission material in the waste streams.

These losses are limited to the economically acceptable limit values by the project when designing the plants, by limiting the degree of oil saturation with heavy metal ions (g + Pu) to <70% of the theoretical value and the precautionary installation of additional extraction capacities.

Intensive instrumental and analytical process monitoring is attempted by the aer company to identify and correct deviations from the standard values as early as possible.

If batches with inadmissible fissile material losses (primarily Pu) occur during non-standard operation, they are usually recycled in the same process unit, which leads to an interruption of routine operation and thus a reduction in availability, i.e. the capacity of the system.

JjI & - 10 - [- ÏO -

The meuri, · j ', \ vatsattung and the oversizing of the extraction K & pc'-i leads leads to increased extractability of some SP, in particular Zr / iD-9ü and Ru, as well as to longer contact sides of the extractant with the highly active solutions and thus to an increase the Sirahlensenaiigur.g res extract i or.sni tt els. The main breakdown products of the IBP, Bibutjl- and Vonobutvlphcsrbut (HIE? And KJö?) Form with Zr non-soluble K3 eoerschl uge (C-ruds) in the extractors.

With rising abortionists, this phenomenon has necessitated a drastic increase in the decay times for the irradiated nuclear fuel I (from 150 - 1 & 0 days as the original standard to at least 3 years).

i While this fact may still be acceptable for the current light water reactor line, it will be a heavy burden for the future fast breeding reactor concept.

Experiments in hot cells on highly burned-off material have shown that the "crud" phenomenon can be suppressed practically even on short cooling sides and is therefore manageable if the extractant, when mixed with heavy metal alloys in the squat-active extract! Purexschema is increased from the current standard values <701 to ^ 9θ £ of the theoretical value.

r Ua. To counter the increasing risk of higher Pu losses as a result, jeucch i: ii should currently be "more open." Process-sept more extraction stages are provided, which by the associated increase in I Extr & KÎionsrr.: Ttelbeirahlung the desired. Benefits lost again i gener. wurcei. ·

L

B ^ i a "closed" concentration of the type of the described 15 ach- i 'treatment process with Pu-Puckgewimur.g, on the other hand, are in the purer- C, r ^ 2! ^ - SSCJÄi * i X "- - it-iS 5.I2.tr I-7752 _ 6 Γ "ΙΙ — v'SsTMUS'üS 'ΊΙλΓΟι- £ 7 * 3: armenxbsr ur.d thus high degrees of saturation even with a reduction I of the current. usual dimensioning of the extraction section (number of! stages or length of the · columns) compatible.

! This nonce :: of ^ limited) bleeding is of course also applicable to the SnaLtraierialf'einreinigungs cycles.

- 11 -

With the introduction of a post-treatment procedure of the type described here, the decoupling of process parameters, the optimal conditions of which contradict each other, gives the Purex process a number of operational advantages such as: - higher contamination factors per process step - lower extractant throughput / fissile material unit - drastic reduction in the need to have to interrupt the KormaT operation in order to recycle "reject batches", ie ultimately - higher availability of the plant.

From a structural point of view, this results in the possibility of making the design of the plant more compact for a certain desired net annual throughput by - reducing the number of stages and thus shortening the columns or batteries in the extraction sections - reducing the capacity reserves due to the lower level of soaps and thus e.g.

- smaller extraction column diameter - reduction of the buffer capacities required between the individual process steps.

The latter facts in turn facilitate the safety-related design for checking criticality and for monitoring the fissile material flow and inventory. The reduction in the steel damage of the extractant, which can be achieved by the high solvent saturation with a simultaneous limitation of the extractor suffles, and the "Orud" formation occurring in parallel, also reveal the delicacy, the decay time. of the irradiated, "burned off" it again, even at high levels of aroma, without having to use highly corrosive reagents such as fluorine ores as complexing agents to suppress the formation of precipitate or on the technological breakthrough of technically very high delicate centrifugal! .cntäkioren type Extrcuzticr.sapparatursr for. drastic. Reduction of the radiation exposure of the Srtrskiicr.sait waiting for .'fisser ..

- 12 - While the tes · hrî «;. ~ Re integrated cachet treatment of waste streams, the degree of freedom i '-' ir the optimization of a number of process parameters of the Pure - procedural increased by accepting current higher Pu losses, it leads the last In the end, a drastic reduction in the Pu losses, which are considered to be final. With regard to the medium-active waste, one obtains on the one hand by washing the extractant with oxalic acid and on the other hand by the CJULL precipitation of the ILLW 5 * - practically "alpha-free" inert salt-containing waste streams, which at least increases the potential hazard to the environment from this waste category reduced in terms of time by orders of magnitude. This also solves the problem of long-term stability as a critical point in the use of cement or bitumen as matrices for the solidification of this waste, which is customary in the current state of the art.

In the case of HA.W sluggish, the other transuranic actins contained therein, Kp, Am and Gm, contribute significantly to the long-term risk and thus to the required final storage period, while the heat generation is determined by the SP (see Fig. 2). The division of the HAW into 2 fractions, the 3P and the An, has therefore repeatedly been felt to be desirable in order to decouple the mutually obstructing parameters for a more targeted and flexible conditioning and final storage of these two groups.

Due to its complex chemical behavior, Dp occurs in unique proportions in all primary process-specific waste streams of the Purenprosesees. However, the tile treatment processes described allow its quantitative separation in practically pure form in a single product stream. The two higher actinides Am and Gm behave analogously to the chemically related SE group and can be isolated together with this on the HAW as a homerere fraction at the end of the hochbfchHndlnr.rsverianren3.

The weight of the SE to the SP is about a third, while the An are of no importance. The heat generation of this group corresponds to approx. 3θί · of the Gsssst-HAW after one year of operation and goes back to the Orcts part on Ge / Pr-1-2. After 10 years practically only the heat remains.

- -t - 13 - -îi-

The use of the rare earth Gd as a homogeneously dissolved keutron venom for criticality control in the resolver (of increasing importance with increasing size of the plants) does not represent a burden for the SP conditioning in an overall scheme of the type described here. His presence in the HAK would be in On the contrary, it is even an advantage to support "criticality control during the OXAL precipitation process and the formation of crystalline, homogeneous and (in contrast to glass) thermodynamically stable!" Matrices for solidification.

«- As far as the other SP is concerned, it seems to be the most sweet thing to embed them in glass. Compared to the current state of the art, this fraction also has advantages from the tiling process. The reducing denitrification with formic acid drastically reduces the volatility of the Ru during calcination or direct glazing, which considerably reduces the burden on exhaust gas treatment.

Due to the absence of the An, the question of the long-term stability (^ 1000 years) of the glass matrix loses importance and, if necessary, the composition can be focused primarily on the immobilization of the most important SP such as Cs and 3r.

c. Summary of the inventive features of the present invention relates to the integration of the previously unknown process steps described under point 3, i.e.

- the extraction agent washing in the first stage with oxalic acid solution - the denitrification and CXAL precipitation of the Purex processes s specific waste -. of the HAU, which was previously the concentrate of the raffinates of the U and

Pu cleaning cycles and the first stage of the extract! OnsmittelWaesch "" has been treated - the treatment of the medium-active operational xeroxic (inerxssls and plexing agent-containing) afofall streams by means of a GXÂL-like precipitant pi-ozesses: .irnductts for the hüokgev.ir.nung of the gap material I Pu and the isolation of the other îrans ai: t in! the outside with the rare earth gap products

I A

- 14 ~: into the Purex process industrially used for the reprocessing of "irradiated nuclear fuels according to the current state of the art".

The technology required to carry out these additional process steps is feasible based on today's industrial experience and consistently uses solutions that are identical to the techniques already used in the Purex process itself, both in terms of chemistry and apparatus.

This identity "results in a high degree of compatibility between all procedural steps, so that all the advantages that deepen the techniques of pecification can be fully exploited.

The resulting. The advantages can be assessed in short form as follows: - Opportunity to refurbish the Lie after a shorter cooldown even for highly consumed learning fuel (of 1 essential importance for the future fast breeder technology) - Extension of the bar. '. Range of the permissible limit values for the individual process parameters , •••With ; - Reduction of the requirements for process control and monitoring [systems - Reduction of the response to batches not in accordance with the intended purpose and the associated. ! · .Interruptions of the continuous operation, thereby I- less need for f- capacity and buffer reserves and connected with it - reduction of the spatial inventory - increase of the loadirator and possibility of a - more coherent off] er 'ί-' the total Sn5¾¾ - reduction of the E: - ..r & ktior.smittslfiusses your nc * re saturation, thereby - reduction of the size of the extraction apr, disclosure (a.3. smaller column:.; ·, -ohmeters and lengths) and - shorter contact list ; r of the Ezvrakiiorsnitiels in the highly active extract izr.s-thus divides .....

- 15 - - Reduced radiation damage to the extractant - improved contamination and separation factors per process step and - suppression of "crud" formation - practically quantitative recovery of Pu (also non-ionic forms), from all waste streams from the reprocessing plant and thus - reduction of "Alpha" -containing wastes, especially in the medium-active range either by "non-generation" (solvent washing with oxalic acid) or by post-treatment and consequently - reducing the associated long-term hazard potential - division of the EAW into a fraction to be regarded as "non-alpha-containing" the heat-generating fission products and a chemically homogeneous "alpha" fraction with the rare earths - reduction of Ru volatility in the glazing and - expansion of the degrees of freedom for conditioning and final storage "after barrel" of the first fraction - idem for the nalph & " -Fraction - Possibility, Kp as longest possible product of the Transursnak Isolate tinides quartitatively in practically pure form for possible even more specific forms of treatment and disposal.

4 _ Ti __ I '- le- - i i I Intel 2 e 1

Content of some other endioel emenia in the supernatant solution - after the Γ \ AL precipitation · of real HAN, obtained from: Kernbr eiu.c τ <- i'z i: .with a margin of 33-000 KWd / t. Element U Pu Kp Am / Gm Ba / 3r Ru Sr / Mo SE + i / grelös., Yy.7 «0.1« 0.1 <0.1 «70> 97 ^ 99 <0.5

E

*! r- \ _ 17 - 17 -

Table 2

Together, they are typical "ILLW waste streams"

Connection / concessional checkout

Element M p / 1 kg / t fuel ENT, 2 ό

NaNû _ 4 540 510 K a PO, 2.2E-1 3o.7 55

NaOi 2.5S-1 15 22.5

N / A. 30 L. ÖE-1 25-7 38.5 HCl 1.0E-2 0.4 0.54

Al 8.2E-3 0.23 0.35 V 'Pv O' E, Ol

Cr 1.5E-3 0 * 08 0.12

Cu 2.4B-2 0.15 0.23 each 4.6B-2 2.06 4 i. 2.7E-3 0.006 0.12

Kg 3.1E-2 0.75 1.1 iin 1.0E-2 0.6 0.9 'Mo 4.0E-3 0.36 0.57 io 1.45-3 0.08 0.12 hu 1.55-3 0.15 0.23

Zn 2.3E-3 2.15 0.23

He 8.65-4 0.ü £ 0.12

Well oral or. /o—".0 7 * 5

Na tartrate 2.2E-2 -.0 7 * 5

Na zixrat 1.9E-2 ~ .0 7.5 515Ό. 2.75-} _.G 1.5 TBP 7-55-4 0-2 0.3 HI) BP 9.55-5 0.02 0.03 H, KBr 3.25-5 1.005 0.007 nerosin J.C2! - 18 -

Vc-f hζ? 11 fe * -1 location S1 '' 4 Γί £

Connection / Koneentretien checkout

Elentent V. c / l kg / t fuel, Pu 1.03-4 0.024 4.CS-2 U t.72-3 1.6 2.4, Np 1.25ε-5 0.003 4 * 53-3; On 1.43-5 0.003 5 * 03-3 "‘ Cm 5 * 53-6 1.33-5 2.03-5

Radioactivity: 0.5 Oi / 1 essentially through

Ce, Cs, Eu, hu / Rh, Sb / T'e, Sr, Y, Zr

Volume: 1 * 5 m_ / T · processed nuclear fuel * x ~; .3, »- Ij -

Fig. 1 ^ --—— γ —- Æ) i Ui ο /! 0Ιβ ™ ,, \, -J.j «5 ______» JF r-. 1 ^ j 2; · "“ “*“ ·)

ou. ; ^ ce “N. AV

Δ · -! » -,: | #V ^ ί i lr ί ^ * £ tu 5.Î2 f ^ £

LifJ rir ^ ·; s £ 2 / T \; if "* t * I ^ axusfeawi * imeuaelagsE% ^ tu · ë" | Uj ^ J · vc Si I Q; ,. ·· * J., · 2 W 5 ................ ... .. · * ..... tu "*“ * -gj · 'î /> ¢: ce ^: [z J / ί έ ^ |: i Tl / ^> \: ʧ I / 2¾ Ό § fesS W | Sf | £) § f ™ «v *

\ fsj /; i i i \ J § / - i I

L, L-jrJ Ç d I is ’S:; · v 1 lu | o’ i. > 0: i L · «ois / ^ T \ *" * ---- yes d I, “{a £ --- 1 Ό S Æ y, - | J t - the» \ D i /. Lad! ^ ts, st I.

v y · i ^ ^; 11, - *; * «S α * Β £ Ϊ4 '1ΦΜ? *** e t * L * ti? : i h *: i j p ^ .. v «» χ> pvi «.. ta • tël. l · · L ._.- / T \ - ίεr- ”............ ~~ -) 1¾i ..... · \} y

Li ^ J lli v- ^ Ψ l f i-i, ·.:; · .. j; ..5 ·! 5 j fi '^ \

h ^ a ^ -rateMBUMTBMnr mm ....... ri. '.> a ^ rWTrri' ^ iBrTW ^ ilîW ^ 1riri ^ 'tt «^ λ: "tf VV Si I

j b J '<zP

• f 1 i

Fi «. ?, 'C'Y. r ....., - .... -. , _______________ τ_. _______: N! .............. .......... 'T f ------

Watt '\. '--f ----- \ ------- 1 GtSAMT-HAW J__________

I; N! i CXAL REPUT

______ 3 Am / Cm-SE FRACTION J _______ ^ ^ j "; ________________..... j ^ x J__VL_______________________

! ::: \ i \\ J

! : r ..... r ........... "χΐΤΓΤΊ’ Β: ϊ. | “£ ΞΪΕΕ:

j, *; Il I I

t '3 5 10 50 100 300 500 1000

Closing time (years)

Claims (1)

1. Extension of the plutonium to the nitric acidic aqueous * solutions by liquid-liquid extraction of such solutions with a solution of Tr i du Ly 1 ph o sph at in a water-immiscible crga- "" f ni see solvent Waste streams freed from plutonium are obtained from the organic and aqueous solutions, characterized * t - η -π - by the fact that these waste streams are each treated with oxalic acid. Burning according to claim 1, characterized in that the organic waste stream is washed continuously with a 0.5 to 1 molar aqueous oxalic acid solution in countercurrent. ... exporters. according to claim 1, characterized in that lanthanide-containing aqueous waste streams are treated with oxalic acid, the lanthanides precipitating as oxalates, co-precipitating the transurans present. ê * * * À fi p