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UO2 Sintered Pellets Microstructure for Advanced Nuclear Fuel

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					                13th Romanian International Conference on Chemistry and Chemical Engineering



                UO2 Sintered Pellets Microstructure for Advanced Nuclear Fuel
                                                      D.Ohâi
                                  Institute for Nuclear Research, Pitesti, Romania


       Keywords: UO2 sintered pellets, nuclear fuel cycles, CANFLEX, SEU, RU, DUPIC
       Abstract
       In the world the new realities in the nuclear field are the unexpectedly slow growth of
nuclear energy, the escalation of back-end costs, the delay in the implementation of fast
reactors, the end of the Cold War. The consequences of these realities are a surplus of
uranium, a continued debate over the choice of fuel cycle, a surplus of separated plutonium,
the demilitarization of plutonium and high-enriched uranium weapons
       An international collaboration between Korea Atomic Energy Research Institute
(KAERI), Atomic Energy of Canada Limited (AECL) and British Nuclear Fuel plc (BNFL) to
use RU was developed. Since 1991, KAERI and AECL have introduced the Canadian Flexible
(CANFLEX) fuel concept. CANFLEX fuel bundle can be associated with many nuclear fuel
cycles being used like vehicle in CANDU pressure tube for different fissile material. A very
attractive alternative to use RU in CANDU Reactors appears.
       The DUPIC (Direct Use for spent PWR fuel in CANDU) and Mixed Oxide (MOX) are
other fuel cycles compatible with CANDU Reactors.
       The Romanian preoccupations to develop advanced fuels for high burnup compatible
with CANDU 6 reactors are presented.
       1. Introduction

       In the world the new realities in the nuclear field are the unexpectedly slow growth of
nuclear energy, the escalation of back-end costs, the delay in the implementation of fast
reactors, the end of the Cold War. The consequences of these realities are a surplus of
uranium, a continued debate over the choice of fuel cycle, a surplus of separated plutonium,
the demilitarization of plutonium and high enriched uranium weapons [1].
       The nuclear fuel fabrication industry is characterized by a continuous upgrading of the
end product, the fuel assembly, driven by the needs of utilities and by feedback from
operating experience. Meanwhile, the safety authorities are increasing their requirements to
license higher burnup fuels, asking for more representative material tests under accidental
conditions, more feedback from experience with lead test assemblies, etc, before granting a
licence. Meeting these more and more stringent constraints requires heavy equipment



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programme which must encompass a large amount of feedback from engineering and
operating experience, the use of heavy testing equipment, and careful step by step licensing.
        Increasing burnup, which allows the utility to get the same kWh output with a reduced
tonnage of fissile material, provides a saving not only in the cost of fuel fabrication but also in
the cost of disposal of the irradiated fuel. The cost of disposal of the irradiated fuel is two to
four times higher than that of fuel fabrication. Reducing the quantity of irradiated fuel also
has a positive impact on the environmentally acceptable solution adopted for its disposal [2].


        2. Technologies For UO2 Large Grains Obtaining
        One of the ways to increase pellets grain size without increasing sintering temperature
and time is the addition of small quantities (< 1% wt M/U) of sintering additives (aliovalent
metal or rare earth oxide).
        By the addition of certain dopants in the UO2 powder (TiO2, Nb2O5, Cr2O3, CaO,
V2O5,) the grain size, porosity and the mean free diffusion path are increased, whereas the
grain boundary area is reduced [3] - [7].
        At the Institute for Nuclear Research (ICN) - Pitesti a technology for obtaining large
grains size UO2 pellets using dopants [8] was developed.
        The UO2 non-free flowing powder, manufactured by ADU route, was mixed with
Nb2O5 in a Y - con master mix. The blended powder was pre-pressed and granulated using a
0.5 mm sieve. The resulted granules were mixed with Zn stearate as lubricant. The green
pellets were manufactured by bilateral pressing. The compacts were directly sintered (4 hours
at 1700°C) in standard continuos sintering furnace with a dewaxing step at 900°C.
        For the production of niobia doped UO2 fuel the “direct pelletizing process” which has
been developed in relation with the AUC powder technology can be applied without any
change beside the admixture of niobia to the UO2 powder. In a master mix UO2 and Nb2O5
powders are added and homogenized. The blended powder is directly pressed without the
addition of a lubricant. The green pellets are sintered in the sintering furnace. Under the same
sintering conditions, the density of the pellets can be adjusted by U3O8 addition, UO2 - Nb2O5
pellets with densities between 9.9 - 10.75 g/cm3 and grain size between 2 - 50 µm being
obtained [9].
        Many other methods to obtain uranium dioxide pellets with large grain sizes are [10],
[11], [12], [13].




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                  13th Romanian International Conference on Chemistry and Chemical Engineering



       3. Obtaining of UO2 Sintered Pellets with Controlled Porosity
       The technology for manufacturing of UO2 sintered pellets with controlled porosity is
relatively simple using pores former. UO2 sinterable powder is mixed with pores former and
thereafter the sintered pellets are elaborated by usually methods[14].
       The common pores former can be organic compounds, UO2 powder precursors (ADU,
UO3, U3O8) or powder resulted from sintered pellets calcination.
       The powder homogenization is very important to obtain uniform porosity. If the mix
powder – pores former is defective appear connected pores and grouping porosity. Therefore,
first step is to make a preliminary admixing of pores former with about 10%wt of lot powder
and thereafter this is mixed with the rest. The manufacturing is continued with compacts pre-
pressing and granulation, granule homogenization with Zinc stearate, green pellets pressing
and sintering in hydrogen atmosphere. The sintered pellets are ginded in a centerless grinding
machine. The washing, the drying and the quality control of the ginded pellets are the last
steps in the obtaining process.
       The pores structure and volume are dependent on pore former percent, pre-pressing
and pressing pressure and thermally treatment conditions. If the technical conditions will be
good choused, desired microstructure is obtained.
       UO2 sintered pellets with homogenous controlled porosity can be obtained directly
from the UO2 powder. UO2 sinterable powder is treated thermally in hydrogen at exceed
reduction temperature (few hours) and thereafter the sintered pellets are manufactured by
usually way. Desired pores structure and volume are obtained in terms of the technical
conditions.
       4. UO2 Pellets Manufacturing from Recovered Uranium.
       Recovered Uranium (RU) resulted like byproduct from conventional reprocessing of
LWR spent fuel for Pu obtaining. The enrichment of RU is about 0,9% U235 and it contain
Uranium daughter products and traces of transuranic elements (Table 1).
                 Table 1. Uranium daughters and transuranic elements in RU.
     U daughters               Contents                 Unat.             Nuclide          Range g/gU)
          U232               0.15 - 1 ppb             0.0055%                Np                  3.10-6
          U234             0.014 - 0.018%                                    Pu                  3.10-6
          U235               0.85 - 0.95%              0.711%               Am                   1.10-8
          U236               0.28 - 0.4%                                    Cm                   1.10-9
          U238             98.856 - 98.632             99.289



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                     13th Romanian International Conference on Chemistry and Chemical Engineering



            The quantity of RU in Europe and Japan is about 25.000 t.
            RU [15] resulted from reprocessing is Uranium Nitrate (UN). UN is converted in UO3
by denitration (BNFL) or in U3O8 by ADU route (COGEMA). The manufacturing of UO2
pellets is usually. UO2 powder is manufactured by reduction of UO3 or U3O8 with hydrogen.
If UO3 is converted to UF6 (for enrichment), the Integrated Dry Route (IDR) is applicable.
UO2 sintered pellets can be manufactured anywhere it exists facilities, devices and
technologies for UO2 no free flowing powder manufacturing.
            The difference between powder and pellets manufacturing from natural uranium and
RU is the radiological inventory of RU, dependent of the fuel history (reprocessing, aging
stage, burnup). In every process volatile fission product can be released. That imposes
supplementary measures for human and environment protection. During sintering the release
     137
of         Cs and other volatile products was detected. But AECL concludes than no significant
field in a commercial fuel manufacturing plant would build up due to release of volatile
products.
            An international collaboration between Korea Atomic Energy Research Institute
(KAERI), Atomic Energy of Canada Limited (AECL) and British Nuclear Fuel plc (BNFL)
to use RU was developed. Since 1991, KAERI and AECL have introduced the Canadian
Flexible (CANFLEX) fuel concept. A very attractive alternative to use RU in CANDU
Reactors appears. Theoretically the quantity of 25.000 t of RU would provide sufficient fuel
for 500 CANDU reactor years of operation, knowing that the annual refueling requirement for
a RU fuel burnup 13 MWd/KgU is around 50 t/an comparatively with 85 t/an for NU.
            5. Sintered Pellets Fabrication for Dupic [16]-[18]
            The DUPIC (Direct Use for spent PWR fuel in CANDU) is the fuel cycle where the
spent PWR fuel is dry processed into CANFLEX fuel for the additional burnup of about 15
MWd/KgUHE (Heavy Element) in CANDU Reactor. AECL, KAERI, USDOE and USDOS
proposed the DUPIC fuel cycle concept.




                                                Fig. 1. OREOX Process


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                 13th Romanian International Conference on Chemistry and Chemical Engineering



       The PWR spent fuel, having a nominal burnap of 35 MWd/KUHE contain 0.9wt%
fissile uranium and 0.6wt% fissile plutonium, to much like natural uranium with 0.711wt%
fissile content. UO2 powder, from PWR spent fuel, manufacturing is based on OREOX
(Oxidation Reduction of Oxide Fuel) process. UO2 oxidation to U3O8, the crystallographic
system is changed elemental cell volume increase, the stresses appear and pellet is broken
themselves. The process is repeated to obtain a very fine powder.




                                             Fig. 2. DUPIC Process
       The powder resulted from OREOX process is conditioned by milling to increase the
sinterability. The sinterable powder is pressed into pellets and green pellets are sintered to 95
- 98 % TD (Theoretical Density). The sintered pellets are centerlines ground (dry process
only) to final diameter and surface.
       The DUPIC being a dry processing technology to manufacture CANDU fuel from
PWR spent fuel without separating fissile materials and stable fission products from the spent
fuel need the experimental works to verify the performance. In Korea, the simulated sintered
pellet where manufactured, capsule and iradiation condition are established. The irradiation
will be started in 2000 year.
       6.MOX fuel fabrication [19],[20]
       The purpose of the Mixed Oxide (MOX) fuel manufacturing is the following
                        to reduce the present stockpile of plutonium
                        to keep the stockpile of plutonium at a minimum in the future
                        to use the weapons plutonium
       The MIMAS (MIcronized MASter blend) proces for MOX fuel obtaining (for LWRs)
was developed by BelgoNucleaire (BN) in the Dessel plant. The MIMAS MOX pellets are a
solid solution of UO2 and PuO2, homogeneously dispersed in a UO2 matrix [61].
Schematically the process is presented in the Figure 3.


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                13th Romanian International Conference on Chemistry and Chemical Engineering




                         Fig. 3. Sintered pellets obtaining by MIMAS process
       The powder preparation have two blending etaps: the primary (master) blend obtained
by ball-milling (micronization) and secondary (final) blend. The powder obtained is
pelletized. The green pellets are sintered in H2/Ar atmosfere and the sintered pellets are dry
grinded.
       The MIMAS process can lead to excellent isotopic homogeneity of the Pu in the
product even whith Pu of various origins, forms and batch sizes because it have double
blending.
       In Canada, AECL MOX fuel fabrication activities are conducted in the Recycle Fuel
Fabrication Laboratories (RFFL).
       The process consists in blending of UO2 and PuO2 powders in Master-Mix blend
followed by usual manufacturing of sintered pellets from non free flowing powder. Therefore,
the powder is pre-pressed, obtained compacts are granulated. The granules are pressed and
green pellets are sintered in hydrogen atmosphere. The sintered pellets are grinded in a
centerless grinding machine.
       7. SEU 43 Romanian Fuel Bundle Concept [21]

       This fuel bundle type is destined to a high burn up and it is compatible with CANDU
6 Reactor systems. To this purpose, the basic overall dimensions of SEU-43 fuel bundle were
designed to be the same as those of the 37-element bundle.


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                 13th Romanian International Conference on Chemistry and Chemical Engineering



       The type of assembly is welded bundles of 43 elements in circular array with brazed
appendages. The major feature of SEU-43 bundle is an increase in the number of fuel
elements from 37 in the standard CANDU-6 bundle to 43 elements. The SEU-43 bundle
consists of 2 fuel element sizes: the 11.50 mm diameter elements (35) in the outer and
intermediate ring, and the 13.50mm diameter elements (8) in the inner and center rings. The
small-diameter elements in the outer ring allow the peak element ratings in the bundle to be
reduced comparatively with the standard 37-element bundle. The larger-diameter elements in
the inner rings of the bundle compensate for the fuel volume lost due to the smaller-diameter
outer ring elements.
       The fissile material is UO2 sintered pellets with 96-97%TD and enrichment about
0.9% U235. The intention is to use Recovered Uranium (RU) resulted from LWR spent fuel
reprocessing.
       Remarks
       In the nuclear word, many efforts for fuel manufacturing resistant to extended burnup
were make. Many new technology offer the posibility to use the available manufacturing
equipment and quality assurance programs from comercial production without specially
modifications
       Increasing burnup, which allows the utility to get the same kWh output with a reduced
amount of higher enriched fissile material, provides a saving not only in the cost of fuel
fabrication but also in the cost of disposal of the irradiated fuel. This latter cost is two to four
times higher than that of fuel fabrication. Reducing the quantity of irradiated fuel also has a
positive impact on the environmentally acceptable solution adopted for its disposal.
       This is true not only for fresh uranium fuel, but also for mixed oxide (MOX) fuel. For
MOX fuel, the goal is to achieve the same higher burnups as with uranium fuel. Higher
burnups make MOX fuel more competitive in comparison with fresh uranium fuel, because
the latter requires more uranium feed and enrichment to achieve higher burnups. Meanwhile,
the safety authorities are increasing their requirements to license higher burnup fuels, asking
for more representative material tests under accidental conditions, more feedback from
experience and testing, before granting a license.


       References
        1. Peter Jelinek, Noboru Oi, “Nuclear Fuel Cycle and Reactor Strategies: Adjusting to
       New Realities”, The Uranium Institute Twenty Second Annual Symposium, September
       1997


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        13th Romanian International Conference on Chemistry and Chemical Engineering



2. Jean-Paul Lannegrace, “Long Term Strategy and Nuclear Fuel Evolution” The
Uranium Institute Twenty Third Annual Symposium, September 1998
3. K.C.Radford, J.M.Pope UO2 Fuel pellet microstructure modification through
impurity additions - J.Nucl.Mat, 116, (1983), 305-313
4. M.El.Sayed, R.Lorenzelli, Kinetics of initial stage of sintering of UO2 and UO2
with Nb2O5 addition - J.Nucl.Mat, 87, (1979), 90-96
5. D.Diaconu, D.Ohâi, V.Bãlan, Microstructural Aspects of grain growth induced by
the dopants in UO2 sintered pellets, Mat.Sci.Forum 126-128 (1993), 427
6. D.Ohâi, D.Diaconu, V.Bãlan, I.Mirion, Sintered pellets having large grains
structure for advanced nuclear fuels, Proceedings of National Energy Conference
CNE’94-Neptun, 1994, România
7. K.W.Song, S.H.Kim, S.H.Na, Y.W.Lee, M.S.Yang, Effects of Nb2O5 addition on
grain growth and densification in UO2 pellets under reducing and/or oxidizing
atmospheres - J.Nucl.Mat, 209, (1994), pp.280-285
8. D.Ohai, Contribution to fuel manufacture technologies for advanced reactors, Ph.D.
Thesis, Bucharest-Romania, 1997
9. H.Assmann et.al.Doping UO2 with niobia - b eneficial or not? - J.Nucl.Mat. 98,
1981, 216
10. I.Tanabe, M.Oguma,m H.Masuda, JP patent document 3-287096/A/, JP patent
document 2-87578
11. K.W.Lay et.al. US patent document 4, 869, 868/A/
12. K.W.Lay et.al. US patent document 4, 869, 867/A/
13. P.Dehaudt, A.Chotard Nuclear Europe Worldscan, No 11-12, Nov.-Dec. 1997, 65
14. D. Ohai, I. Dumitrescu, D. Benga, V. Balan, M. Mihalache, Correlation Of Uo2
Sintered Pellets Porosity With The Fabrication Technological Parameters,
15. H.C.Suk, J.H.Park, B.J.Min, K.S.Sim, W.W.Inch, T.G.Rice, “Technical Aspects
And Benefits Of The Use Of RU In CANDU Reactors”, Proceedings of Sixth
International Conference on CANDU Fuel, 1999 September 26-30, Niagara Falls,
Canada, vol.1, pag.444-459.
16. J.D.Sullivan, D.S.Cox, “AECL’s Progress In Developing The DUPIC Fuel
Fabrication Process”, Proceedings of 4th International Conference on CANDU Fuel,
1995 October 1-4, Pembroke, Canada, pag. 4A_49-4A_58




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        13th Romanian International Conference on Chemistry and Chemical Engineering



17. J.D.Sullivan, M.A.Ryz, J.W.Lee, “AECL’s Progress In DUPIC Fuel
Development”, Proceedings of 5th International Conference on CANDU Fuel, 1997
September 21-25, Vol.1, pag. 300-310
18. K.Bae, K.Kang, M.Yang, H.Park, “Irradiation Plan Of DUPIC Fuel At Hanaro”,
Proceedings of Sixth International Conference on CANDU Fuel, 1999 September 26-
30, Niagara Falls, Canada, vol.1, pag.460-469
19. D.Haas, “MOX Fuel Fabrication Experience At Belgonucleaire”, IAEA-TECDOC-
941, May 1997, pag. 77-89
20. F.C.Dimayuga, “AECL’s Experience In MOX Fuel Fabrication And Irradiation”,
IAEA-TECDOC-941, May 1997, pag.373-385
21. Dumitru Ohai, Gheorghe Andrei, Romanian Program for Seu/Ru Fuel
Manufacturing at Nuclear Site Pitesti, “7th International Conference on CANDU
Fuel”, September 23-27, 2001, Kingston, Ontario, Canada




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Description: UO2 Sintered Pellets Microstructure for Advanced Nuclear Fuel