be& selective separation of actinides and long lived fission products from aged liquid wastes ...

22 Pages·2012·0.9 MB·English

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ISSN10393-6252
BE&
COMITATO NAZIONALE PER LA RICERCA E PER LO SVILUPPO
IDELUENERGIA NUCLEARE E DELI-It ENERGIE ALTERNATIVE
SELECTIVE SEPARATION OF ACTINIDES
AND LONG LIVED FISSION PRODUCTS
FROM AGED LIQUID WASTES PRODUCED
BY THE EUREX PLANT AT'SALUGGIA1*1
Paper presented at the IAEA/CEC International Symposium on Management of Low
and Intermediate Level Waste, Stockholm, Sweden, May 1988
L. PIETRELLI, G. GROSSI
ENEA - Dipartimento Ciclo del Cornbustibile, Centro Ricerche Energia Casaccia
F. TROIANI
ENEA - Dipartimento Ciclo del Combustible, Centro Ricerche Energia Saluggia
RT/COMB/89/8
Testo pervenuto nel maggio 1989
Progetto ENEA: Trattamento condizionamento residui radioattivi (CH)
I contenuti tecnico-scientifici dei rapporti tecnici dell'ENEA
rispecchian6 'opinions degli autori e non necessariamente quella dell'ente.
3
ABSTRACT
I
The chemical process for the selective separation of actinides
and long lived fission products fromaged MTR liquid wastes is
described.
To perform t'his selective separation, some chemical procedures
such as precipitation and ion exchange, both in acidic and
alkaline media, have been investigated, mainly at the laboratory
scale, using simulated and traced waste solutions. -
To confirm the results obtained with imulated or traced
solutions, additional tests with true MTR wastes are in progress
in the 'analytical hot cell of ENEA EUREX pilot plant in
Saluggia.
With all these results, it is possible to perform a preliminary
selection of the reference process.
Further larger scale information will be available after
experimental runs on a cold pilot plant, named SERSE, designed
and built at Saluggia for engineering scale demonstration of the
chemical separation process.
This R & D work was performed and partially funded in the frame
of the research programmes of the.European Commun ties.
RIASSUNTO-
Nei presente lavoro, presentato ad un congresso congiu nto
IAEA/CEE tenutosi nel 1988,' vengono riportati i risultati
preliminari della sperimentazione effettuata con soluzioni
simulate e tracciate del rifiuto 1AW-MTR stoccato presso
l'Impianto EUREX di Saluggia.
II trattamento chimico del rifiutQ consiste di una separazione
selettiva della components radioattiva ed i processi scelti per
tale separazione prevedono sia tecniche di scambio ionico e/o
adsorbimento sia.tecniche di precipitazione selettiva.
Viene inoltre riportata una descrizione sommaria sia
dell'allestimento effettuato nella cella analitica
dell'Impioanto EUREX sia dell'Impianto Pi-Iota "freddo"
denominate SERSE destinato ad una sperimentazione in scala
significative dei processi scelti per il trattamento di tali
rifiuti.
Parte del lavoro stato finanziato con i contribute della
ComunitA Economica Europea.

5
1. INTRODUCTION
The liquid wastes produced by the ENEA reprocessing pilot
plant EUREX (Saluggia), dring campaigns with MTR type
irradiated fuels, contain very large amounts of inactive salts
such as aluminium nitrate > M) and significantamounts of
other chemicals such as iron, mercury and sulphates.
In addition, these liquid wastes were produced in the early
19701s, so that at present essentially all their radioactivity
is due to Cs-137, Sr-90 and transuranium actinides.
The selective separation of actinides and long lived fission
products from aged liquid wastes containing large amounts f
inactive chemicals may simplify very much their management,
expecially in terms of overall production of active glass,
because by this separation only the radioactive fraction is
vitrified, the larger fraction containing the inactive chemicals
being conditioned with the usual procedures for intermediate and
low level wastes, such as cementation.
Such an approach has been adopted by DOE in the USj for the
management of the very large amounts of the I'd' efensell HLW at
Hanford and Savannah River, as well as for the "commercial" HLW
at West Valley 12,3].
2. EUREX MTR LIQUID WASTE COMPOSITION.
The EUREX MTR liquid wastes consist of 85 m3 of 1M IIN03
aqueous solutioncontaining aluminium nitrate (>1M) and other
minor costituents, including fission and activation products and
reprocessing chemicals. This wastes are currently stored in two
stainless steel tanks (F-710/D and F710/C) . Table I gives the
estimated (by ORIGEN code and/or process analytical data)
chemical composition of the total 'waste, and the real
composition (partial) of the solution stored in the first tank
(F-710/D).
The significant differences between estimated and found
chemical composition may be explained considering that:
- the estimated composition refers to the average composition of
all MTR liquid wastes tanks F710/D and F710/C);
-the found composition refers to liquid waste contained in the
tank F710/D, where the liquid effluents coming from the first
"cold" runs with inactive fuels (carried out at EUREX plant
before the hot startup) were collected.
As a consequence, a significant increase of aluminium
concentration and a dilution of radioactive components, with
respect to the average estimated composition, is to be expected.
As soon as possible, the liquid waste the second tank.
(F-710/C) will be sampled and analysed. In this case, it has to
be expected to find higher values for radioactive components and
lower values for aluminium.
6
TABLE 1. ESTIMATEDa AND REAL' COMPOSITIONS OF THE MTR
WASTE SOLUTION
Estimated F-710/1) Estimated F-710/1)
concn concn concn concn
Al 27.01 g/L 34.42 g/L Sr 7.8 mg/L 5.8 mg(cid:1)l-
Hg 1.34 g/L 0.79 g/L Sm 7.3 mg/L
Fe 0.56 g/L 0.43 g/L Y 5.0 mg/L
Si 0 14 g/L - Np 1.8 mg/L
Zr 35 mg/L - Pu 2.5 mg/L 0.27 mg/L
Nd 35 mg/L - PM 0.07 mg/L
MO 27 mg/L - Am 4E-3 mg/L
Ce 20 mg/L - W IAM 0.92M
Cs 17 mg/L 11.8 mg/L 6E-3M 1.7E-2M
S04-
Ru 13 mg/L - NO- 4AM 4.65M
3
Ba 12 mg/L - 1.23 g1ml, L242 g/mL
d24'C
u 6.5 mg/L 4.8 mg/L
Radiochemical conTosition
Estimated F-710/13
Pu 1.4 mCi/L' 0.074 mCi/L
Np-237 1.3E-6 mCi/L n.d.
Am-24i 7.8E-3 mCi/L n.d.
Cm-244 1.2E-5 rnCi/L
Sr-90 0.6 Ci/L 252 mCi/L
Cs- 37 0.6 Ci/L 290 mCi/L
Cs 134 0.11 rnCi/L
Co-60 0.05 mCi/L
Eu 154 0.52 mCi/L
Sm 15 1 1.2E--(cid:1)2 CUL n.d.
Sb-125 n.d.
Pm-147 6.3E-2 Ci/L -
Gross - 1060 mCi/L
Gross -y 372 mCi/L
Estimated from ORIGEN code and/or process analytical data.
From F710/1) waste tank.
I Ci = 37 Gl3q.
n.d.: not detectable.
not yet determined.
7
3. CHEMICAL SEPARATION PROCESS AND EQUIPMENT.
The main purpose of the chemical separation process is to
obtain the following two liquid streams:
- a low activity level, high salt content stream, (MTR
"declassified" solution);
- a high activity level, low salt content stream of much smaller
volume, containing essentially all the fission products and
actinides.
The high-level fraction must be vitrified, and the
low-level one must be suitable for cementation; both
conditioning processes should give rise to the smallest solid
waste volume.
In Fig. is shown a schematic flow-sheet for the management,
of the high-level MTR liquid wastes.
To meet this objective ENEA has extensively studied and
investigated, until now mainly at laboratory scale, the
following hemical processes 45,6]:
-Acidic Declassification Process separation of caesium
by precipitation with Phosphotungstic Acid (PTA),
and removal of -Sr and Actinides by adsorption on
Polyantimonic Acid (PAA) or other chemicals
- Alkaline Declassification Process nol: alkalinization
of the acidic waste by Sodium Hydroxide--to pH > 13 (Sr and
actinides. co-precipitate with ferric hydroxide) and cesiulm.
precipitation by Sodium Tetraphen .ylbbrate;
-Alkaline Declassification Process n2 alkalinization
of the acidic waste by Sodium Hydroxide to pH > 13 (Sr and
actinides co-precipitate with ferric hydroxide) and cesium
removal by adsorption using zeolite.
The larger part of the experimental tests were performed
using simulated and traced MTR waste solutions.
In order to verify the performances found in the previous
tests with simulated. and traced solutions, 341 of real MTR
waste solution were sampled from the F710/D stainless-steel
waste tank. All the before mentioned processes under study w_
being verified with the real waste.
The experiments we beingcarried out in a previously
decontaminated hot cell of the EUREX plant. The small pilot
plant mainly onsists of:
- a glass reactor (11 useful capacity) equipped with
a mechanical st'irrer and temperature and pH control;
- a peristaltic pump with changeable head and flow-control;
- a centrifuge;
- a set of glass column (I.D.=8 mm, H=200 mm) provided of a
safety containment "tank" to avoid any unwanted discharge of
the radioactive solution in the cell;
usual laboratory equipment (balance, magnetic stirrer,
dryer, tc.).
The hot small pilot plant will be operated by rmote
controls and telemanipulators.
AL Lhf, ame ime i order Lo allow the engineerinq r(cid:1)cale
demonst-raLion of the chemical separation process, a cold
plant alled SER,((cid:1)E (SEp'arazione Rifiuti Serhatoi Eurex) wa
designed [71 and constructed by scaling p the preliminary
results bained with traced solutions durin t atory
tests.
The main goals to be prsued are te fol--](--)wjnq:
- to ostruct a flexible pilot pant for an ay punc(-75,
demonstration o a significant scale, in order, Lo a I ow
future process modifications to optimize the waste management;
- Lo verify the most important engineering parameters, in order
to obtain useful information f o r the. design o f 1-111(-
future hot facility;
- Lo selec eipment, materials and components wch would llow
an easy and safe operation of the future hot facility.
r
DIRECT VITRIFICATION
Active glass: 10-15 3
Cemented LLW: 700 drums
LA I I SLUDGEan
WASTES (Cs, Sr, actinid
VITRIFICATION
rums (I 1.5 M3)
WLASLT ORAGE FINAL STORAGE
IN SUPERFICIAL IN DEEP UNDERG.
GEOL FORMATION CLAY FORMATION
FIG. Flow sheet for he management of HLW at present stored at the EUREX plant.
9
4. PROCESS DESCRIPTIONS AND PERFORMANCES.
4.1.ACidic process.
4.1.1.Caesium separation.
The caesium precipitation process using Phosphotungstic Acid
(PTA) was studied during the first step of.the current research.
According to the results of the tests with simulated and
traced solutions, with the addition of one mole of PTA per mole
of Cs in the acidic waste solution without any pH adjustment, a
Cs Decontamination Factor (DF) of 100 can be achieved at room
temperature.
The addition of KNO3 to the -initial feed 01-0.2 molll)
proved to be beneficial: a 'significant increase of the Cs DF
values was thus -obtained. This is due to the fact that the
co-precipitation of potassium-PTA adduct together with te
cesium-PTA adduct improves the overall Cs recovery (1]
Long contact times >15) were required at 25'C to reach high
Cs DF values, as well as to obtain a particle size suitable to
perform an efficient solid-liquid separation 4.
In order to ensure the highest Cs DF, two subsequent
precipitation s'trikes are suggested.
4.1.2.Strontium and Actinide separation.
The. separation of Sr and Actinides from Acidic waste by
adsorption on Polyantimonic Acid (PAA) was studied. PAA is an
inorganic compound which can give high decontamination factors
both for Sr and Pu i acidic media [8). The adsorption on PAA is
characterized by very slow kinetics (contact times of more than
15 are usual) In addition, the strong competition due to the
high salt content in the MTR liquid waste drastically reduces
the Pu removal from the solution (DF=10 after 20 hours) 45].
For the same reason, the Sr adsorption capacity is also
reduced from 70 mg/g PAA 9 to =mg/g PAA. In any case the Sr
adsorption on PAA is very slow and multiple columns were needed
to increase contact time between the solution and PA,-in order
to achieve useful Sr DF.
4.1.3.Overall process performance.
According to the laboratory test results 45,6] the assumed
process conditions and performances of the Acidic process for
the 85 m3 of EUREX-MTR waste are as follows:
Precipitation of CsPTA
two precipitation strikes and separation by filtration
temperature: 250C
contact time: >15 hours for each strike
Cs DF >100
PTA requirement: 22.1 mol (as H3PW12040)
KNO3 requirement: 8500 mol to increase the concentration
to 0. M.
- Sr and Pu adsorption on PAA
PAA requirement: approximately 700 kg PAA is required,
10
with a loading of mg Sr/g PAA.
flow rate: 0.5 bed volume per hour, with three columns
to increase residence time, and thus DFs, and
utilized capacity
Sr DF: >100,
Pu DF: low
The Cs-PTA precipitate and the loaded PAA ion exchanger are
sent together with the glass formers to the vitrification system
for immobilization. The decontaminated acidic supernatant is
then neutralized and blended with cement to form a stabilized
LLW product.
The main advantages of this process are that all mercury
goes directly to the LLW reducing the troubles due to its
volatilization during glass melting, and that a relatively small
quantity of HLW is produced.
The major disadvantage is that the DF of the Pu and other
actinides is very small.
The exact loading capacity of Sb into the glass must be
better studied, although some experiment have already been
carried out at ENEA laboratories [10] in this field. According
to these studies, no more than 17 wt% (as Sb203) may be loaded
into the vitrified product. In addition, Antimony is a rather
volatile element (melting point 670'C), and it will probably
cause some difficulties with melter operation and piping
occlusion.
To increase the low Pu DF obtained with the acidic process
using PAA some experiments have been performed using an actinide
specific reagent normally used in the TRUEX solvent extraction
process, developed at the Chemistry Division of Argonne National
Laboratories (Chicago-USA)[111, and adopted by ENEA, with proper
adjustments, for the treatment of alpha contaminated liquid
wastes coming from plutonium facilities 12].
This reagent,Octyl(phenil)-NN-diisobutylcarbamoylmethyl-
phosphine oxide (CMPO) has the peculiarity that it is water
insoluble and it is effective even in acidic media.
Tests have been performed using CMPO adsorbed on PAA 6,
this product was then used to fill a column.
The results were not satisfactory: in fact 3after only 26
column bed volumes the Pu DF decreased from 10 to 33. This
behaviour is probably due to the limited impregnation capacity
of PAA with respect to CMPO, as the specific surface of PAA is
low 20.9 m2/g).
The comparison of the results obtained with PAA and PAA+CMPO
shows that, although at the beginning very high Plutonium DF may
be obtained with PAA+CMPO, after a relatively small amount of
waste solution treated (compared with the total amount of
available CMPO), the significant decrease of Pu DF does not
permit an adeguate decontamination. 'On the other hand, the use
of a different inorganic support with higher specific surface
(at present some experiments are in progress using CMPO adsorbed
,on silica) could increase the efficiency of adsorption by CMPO.
However, in this case, an additional specific adsorbent for the