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Conditioning of cladding waste

De
126 pages
Press compaction and encapsulation by low melting metal alloys
Nuclear energy and safety
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ISSN : 0379-4229
CODEN : EARRDF 4(2) 305-560 (1982)
REPRINTED FROM
VOLUME 4
N UMBER 2 EUROPEAN I 982
APPLIED RESEARCH
REPORTS
A Journal of European Science
and Technology
NUCLEAR SCIENCE AND TECHNOLOGY SECTION
Published for the Commission of the European Communities,
Directorate-General for Scientific and Technical
Information and Information Management
CONDITIONING OF CLADDING WASTE : PRESS COMPACTION AND ENCAPSULATION
BY LOW-MELTING METAL ALLOYS
By J. Broothaerta et al. (EUR 7876)
Harwood Academic Publishers
ï* ¿13
European Appi. Res. Rept.-Nucl. Sci. Technol.
Vol. 4, No. 2 (1982), pp. 413-534
0379-4229/82/0402-413 $ 18.80
Printed in France
CONDITIONING OF CLADDING WASTE :
PRESS COMPACTION AND ENCAPSULATION BY LOW-MELTING
METAL ALLOYS
(Final Report)
J. BROOTHAERTS, L. DE WILDE, F. CASTEELS, P. DE REGGE,
J. VANDERSTEENE, H. VANBRABANT
Studiecentrum voor Kernenergie I Centre d'Etude
de l'Énergie Nucléaire, S.C.K. /C.E.N., Mol, Belgium
This work was performed as part of the European Atomic Energy Community's Indirect
Action Programme (1975-79) on "Management and Storage of Radioactive Waste" — Action
No. 2 : Decontamination and Conditioning of Cladding Hulls from Irradiated Fuel Elements,
Contract No. 030-76-11 WAS Β
EUR 7876
©ECSC, EEC, EAEC, Brussels and Luxembourg, 1982 ¿u
TABLE OF CONTENTS
PREFACE
1. INTRODUCTION
2. EXAMINATION OF FILLING ALLOYS
2.1. Metallurgical data
2.2. Literature review with respect to corrosion
2.3. Experimental corrosion studies
2.4. Leaching of embedded compacts in Antwerpian ground water
2.5. Interaction of lead alloys with cladding waste components and
canister materials
2.5.1. General
2.'5.2. Experimental studies
2.6. General conclusions and selection of the embedment alloy
3. LEACH TESTING OF ACTIVE HULLS
3.1. Introductory remarks
3.2. Sample preparation
3.2.1. Embedded hulls
3.2.2. Bare hulls
3.3. Experimental set-un
3.4. Analytical techniques
3.5. Results
4. TECHNICAL STUDIES
4.1. Compaction of unirradiated cladding waste components
4.1.1. Experimental
4.1.2. Results
4.1.2.1. Effect of compaction parameters
4.1.2.2.t of hydrogen
4.1.2.3. Effect of other cladding waste components
4.2. Embedment in low-melting alloys
4.2.1. Experimental procedure
4.2.2. Results
4.2.2.1. Effect of the casting technique
4.2.2.2.t of other parameters on rest porosity
4.2.2.3. Hydrided zircaloy
4.2.2.4. Shrinkage of filling alloy during solidification
5. CONCEPTUAL STUDY
5.1. Cladding waste arisings and activity data
5.2. Conditioning 415
5.2.1. Reference process
5.2.1.1. Design basis
5.2.1.2. Process description and facility lay-out
5.2.1.3. Processing capacity
5.2.2. Press design
5.2.2.1. Technical characteristics
5.2.2.2. Press operation and maintenance interventions
5.2.2.3. Lifting carriage and ejection slide
5.2.3. Embedment oven
5.2.3.1. Heating power
5.2.3.2. Oven construction
5.2.3.3. Canister cooling
5.2.4. Canister and loading system
5.2.5. Welding station
5.2.6. Secondary waste
5.2.7. Cost estimate
5.3. Product characteristics
5.3.1. Thermal conductivity of product matrix
5.3.2. Radiation aspects
5.4. General : availability of lead
5.5. Hot-cell test unit
6. CONCLUSION
REFERENCES
ACKNOWLEDGEMENTS ¿16
SUMMARY
Experimental and conceptual studies have been carried out with a view to
develop a conditioning process for cladding waste based on press compaction
and lead matrix formation.
Potential embedment alloys, including Pb,Pb-Sb, Pb-Te, Pb-Sn, Pb-Ag and
Pb-Sn-Ca formulations, and samples cut from compacted zircaloy tubing
embedded in Pb 1.5. Sb-alloy, have been submitted to long term corrosion
tests in demineralized water, in groundwater from the aquifer covering the
Boom-clay formation in the Mol-region and in an atmosphere containing
nebulized clay-water. Whereas strong corrosion was observed in the deminer­
alized water, most of the alloys exhibited excellent resistance to the
groundwater and to the clay-water mist on account of the build-up of
protective surface layers. Pure Pb and the Pb 1.5 Sb-alloy were found to
present the best overall corrosion resistance. Based on the specimen weight
changes measured, then rate zas 0.1 to 1 Um y -1 in the ground­
water, 1.5 to 2 Um y-1 in the humid atmosphere and of the order of 100 Um
y-1 in the demineralized water. The nature and composition of the reaction
products formed have been determined by structural analysis techniques.
Interactions of the proposed embedment alloy, Pb 1.5 Sb, with the cladding
waste components and with candidate canister materials have been examined.
Effective wetting of zircaloy was only obrained after long contact at
higher temperatures. Deleterious effects on the canister material were
found to be due to impurities present in the embedment alloy used in the
preliminary tests and were avoided by using an alloy composition of better
purity.
Samples of compacted, inactive, zircaloy tubing embedded in Pb 1.5 Sb
showed excellent resistance to leaching by Antwerpian groundwater. The
product quality with respect to the immobilization of radio-active contam­
inants could not be demonstrated however, as excessive contraction of the
embedment alloy occurred in test samples prepared with uncompacted hulls
from an irradiated fuel rod.
Appropriate work conditions for the conditioning process have been deter­
mined in technical-scale tests with inactive materials. The volume reduc­
tion achived, and product density obtained, were measured at different
compaction pressures, waste compositions and press loading conditions.
Densification to 70 7, of the density of zircaloy is obtained at 245 MPa, to
60 % TD at about 140 MPa. No special effects were observed in the case of
hydrided zircaloy. The compacts formed were satisfactorily embedded in Pb
1.5 Sb at a temperature of 723 K. Casting of the alloy at atmospheric
pressure resulted in a product with rest porosities corresponding to 15-18
vol. Z, whereas this was only 3-5 vol. 7, after a preliminary degassing of
embedment container to 150 Pa. 417
Examination of product sections indicated that excellent penetration of the
embedment alloy occurs even in highly densified zones, and that adequate
joining with the waste components is obtained, although no soldering
interaction with zircaloy is found at this embedment temperature. The rest
porosity was found to consist of isolated voids, scattered throughout the
matrix.
Conceptual work has been devoted to the industrial applocation of this
conditioning method, and a suitable process lay-out for the waste arisings
from medium to large-capacity reprocessing plants has been developed. With
a press force of 15 MN, waste batches of 20-25 kg would be compacted to 60
% TD in a die of 340 mm diameter. This corresponds to a volume reduction
factor of about 4.4. The briquettes formed would subsequently piled-up in
closely fitting storage canister which, after evacuation to 150 Pa and
heating to 723 K, would be filled-up with the embedment alloy. At a total
height of 1500 mm, these canisters would contain the cladding waste aris­
ings from about 1.4 tU. A basic design of the press, adapted for remote
operation and maintenance, and a functional lay-out of the process cells
and equipment have been worked out.
Cost estimations made for an application in a 300 tU/y -plant indicated
that the cladding waste produced could be conditioned at an expenditure of
80 000 BF per tU processed.
In completion of this programme, preparative work has been devoted to the
definition of the equipment to be installed in a hot cell test facility for
further work with active hull materials. 418
PREFACE
The mana.geme.nt o¿ cladding umte. deienvzi ierioui concern on account o(¡
the voluminoui amounti Involved and the high leveli o(, β/γ- and a-activitieA
aiiociated. The current practiie o(¡ itonage, ¿η water-(¡tiled baiiim οι
iiloi, ¿6only provisionally acceptable ; economic and ia(¡ety requinementi
call {¡on. convention o{¡ thti law-demiXy metallic ¿map into a mote compact
and betten, isolated (¡onm.
With the view o& making iuitable conditioning pnoceA&e& available at the
time o& large-icale {¡uel reprocuiing in the member itateli, R i V-wonk in
thti (¡ield hai been initiated by the European Atomic Energy Community.
Presi compaction followed by embedment o& the demi^ied material in loui-
melttng lead alloy ¿i pnopoied by the SCK/CEN. Tkii procesi would achieve
coniidenable volume reduction and aaure immobilization o(¡ the (¡tiiion
products and TRU-elementi present in a matnÂx o{¡ high integrity and long
tehm itabiliXy.
The development pn.ogh.amme undertaken included :
- examination and ielection o{ appropriate embedment alloyi with regard to
the conAoiion and leach resistance ;
- detenmination o{¡ the open.atX.onal conditiom ion compaction and embedment,
in tests with inactive materiati ;
- conceptual itudies referring to an application o& thii method in medium -
to lange-capacity nepnoceiiing pianti. 419
1. INTRODUCTION
In large-scale chop and leach processing of spent LWR-fuel, the assemblies are
sheared to small sections which are subsequently immersed in nitric acid for
dissolution of the contained fuel. If the massive end nozzles and adaptor
plates are preliminary sawed-off and discarded saparately, the waste material
discharged from the dissolver afterwards will consist of short lengths of hulls,
guide- and instrumentation tubes, mixed with fuel rod plugs, springs and spacer
grids. It is highly radio-active and contains significant levels of fission
product- and TRU-element contaminations.
This low-density metallic scrap, almost completely composed of empty, thin-walled
tube sections and honeycomb structures can be readily reduced in volume by
mechanical compaction, a technique currently used in the nuclear industry for the
densification of low-level waste.
Press compaction in particular is attractive ; densification up to 70 \ of the
theoretical density of the component materials can be accomplished, and the
material is consolidated into briquettes offering sufficient isolation from the
air so that pyrophoricity risks during further handling are minimal.
The corresponding volume reduction will result in substantial savings in the
transport-, storage- and disposal costs. It brings however no protection against
dispersion of the contaminants into the environment in the event that the inte­
grity of the canister would be lost, as the remaining porosity offers considerable
surface for leaching.
Immobilization of the long-lived fission products and TRil-elements should bo assured
by filling-up the voids with an appropriate encapsulation material. The radiation
dose to be endured t109-1010 Rads) excludes the use of organic resins. Glass and
cemont on the other hand would nnt penetrate into the narrow pores obtained after
compaction.
Low-melting lead alloys present useful properties. Appropriate viscosity is
obtained at reasonable temperatures to allow the densified material to be properly
filled-up. On account of the corrosion resistance offered by the embedment alloy,
this should securely isolate the waste during long periods of time. The tech­
nology of application is well suited for hot-cell work, a practical advantage
being the reversibility of solidification, and the operational temperature is
sufficiently low to maintain volatilization of fission products at a negligible
level.
Press compaction followed by embedment of the densified material in lead alloy
presents therefore potential interest for the conditioning of cladding waste.
The experimental and conceptual studies reported have been undertaken with a view
to acquire basic data for the development of this process and to evaluate its
suitability for industrial application. 420
2. EXAMINATION OF FILLING ALLOYS
2.1. METALLURGICAL DATA
Based on literature rinta with respect to potential interactions of the filling
alloy with disposal environments, and with the cladding waste and canister
materials, a few lead alloys have been selected for detailed corrosion and
material compatibility studies.
The chemical compositions are given in Table I.
Relevant data with regard to melting points and viscosities are given in Table II.
The Pb-Sb phase diagramme is of the eutectic type. The eutectic temperature is
525 K for a Pb 11.1 wt % Sb composition. The solubility of Sb in lead is very
low at 373 K ¡ (0.44 wt %). The metallographic structure represents two phases :
Pb(Sb) and Sb(Pb). The Pb-Sn phase diagramme is also of the eutectic type with an
eutectic temperature of 456 K at the composition Pb - 61.9 wt % Sn. A rather
high solubility of Sn (% 2 %) in Pb exist at room temperature.
Ca is insoluble in Pb. The addition of Ca to Pb considerably increases the melting
point of the alloy. The structure of low alloyed Pb-Ca is composed of Pb and the
intermetallic compound Ca Pb3.
The Pb-Ag system is of the eutectic type. The eutectic temperature is 577 K.
The eutectic composition corresponds with 95.3 wt % Pb - 4.7 wt \ Ag. As Ag is
insoluble in Pb, the phases in equilibrium beneath the eutectic temperature are
Ag(Pb) and Pb.
Te is known to be insoluble in Pb. Pb and the intermetallic compound Pb Te are
the phases found to be in equilibrium in a lead-based, low-alloyed Pb-Te alloy.
2.2. LITERATURE REVIEW WITH RESPECT TO CORROSION
A bibliographic study has been made to identify the factors influencing the
interactions between cladding waste and the filling alloy (lead), between the
filling alloy and the canister material md between the filling alloy and the
aqueous or atmospheric environments to be met during storage and disposal.
The composition of the aqueous environment has a considerable influence on the
corrosion resistance of lead and its alloys.
Some impurities of natural waters protect lead and lead alloys against corrosion.
Lime, Chromate, sulphate, phosphate and silicate even in very low concentrations
act as corrosion inhibitors (1).
The influence of chloride ions is complex. The corrosion rate is maximum in
aqueous solutions containing approximately 60G0 ppm Cl~. Solutions containing
very large quantities of chloride ions, on the contrary, give small corrosion
rates (2). This is due to the fact that chloride ions, as well as sulphate ions,
act as anodic inhibitors for lead in neutral solutions.
Since the corrosion resistance of lead in water is mainly determined by the surface
oxide film formed before or during immersion, all factors breaking down or
improving the protective film will influence the corrosion rate.
Carbon dioxide present in the water, or in the atmosphere above it, will reduce
the corrosion of lead due to the protective nature of lead carbonate. Oxygen
Dn the other hand will increase the corrosion rate ; the corrosion proceeds up to
the moment that 20 mg of lead is dissolved in 100 ml of water, resulting in an
alkaline solution. The corrosion rate in water containing oxygen is complex.
It decreases with increasing temperature, in the temperature range between 268 and
333 K (3).

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