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membranes. By Cs permeation through the membrane, more than 99.8% of the caesium can be removed from acidic solutions. 2.3 Fuel and target technology 2.3.1 Plutonium and uranium The fabrication of MOX fuel assemblies is current technology, available on a commercial basis. The present capacity of the fabrication plants in Belgium, France, United Kingdom and Japan is about 200 tHM/year but extension of the capacity to more than 300 tHM/year in the year 2000 is foreseen. MOX fabrication is about four times more expensive than the regular uranium oxide fabrication ($275~300 /kg-U) owing to increased safety requirements which principally concern the radiological protection of the workers by performing the fabrication processes in glove-boxes and the safeguards system. On the contrary, recycling of plutonium will lead to savings in the front-end of the fuel cycle, specifically mining and enrichment, which are the major contributors to the dose rates of the workers and to the costs of the front-end of the fuel cycle. Inert-matrix fuels are a recent development to obtain optimal incineration rates. In such fuels, the compound(s) to be transmuted is (are) mixed with a neutron-inert material. This can be done as a dispersion (macroscopic scale) or as a solid solution (microscopic scale). Typical materials that are being considered as inert-matrix are oxides such as spinel (MgAl 2 O 4 ), yttrium oxide (Y 2 O 3 ), yttrium aluminium garnet (Y 3 Al 5 O 12 ), metals such as tungsten or vanadium, or silicon carbide (SiC). However, additional research needs to be done on the fabrication and characterisation of such fuels, especially with respect to the irradiation behaviour and their compatibility with current reprocessing techniques. In addition to oxide fuels, nitride and metal fuels are being considered for the incineration of plutonium. Both new fuel types are compatible with liquid sodium and can therefore be considered for fast burner reactors. Nitride fuels, as proposed for CAPRA type burner reactors, have been produced on a laboratory scale in the 1960s. As mentioned before, the present interest in nitride fuels originates partly from the fact that oxide fuels with high plutonium content do not dissolve in nitric acid, whereas nitride fuels do. Another advantage of nitride fuels is their much higher thermal conductivity, which will lead to lower central fuel temperatures and, hence, increased safety margins. The major disadvantage of nitride fuels is the 14 C production as a result of (n,p) reaction on 14 N. Enrichment in 15 N of the nitrogen used for the fuel fabrication is therefore a requirement and the recycling of enriched nitrogen would be facilitated by using pyrochemistry as investigated in Japan. In the USA, metal fuels consisting of an alloy of uranium, plutonium and zirconium (Zr) was investigated for the IFR concept and production has been achieved on a pilot scale. Like nitride fuels, metal fuels have much better thermal conductivity than oxide fuels, in addition, they have good radiation stability as a result of which very high burn-ups can be reached. In the IFR/ALMR concept, the metal fuel would be reprocessed by molten salt techniques. 2.3.2 Minor actinides For neptunium, recycling in MINOX fuel seems to be a feasible solution as it can be done in existing MOX fabrication facilities with minor adaptations. These adaptations principally concern 37
improvement of the biological shielding and enhancement of the level of automation to cope with the higher radiation levels. This will lead to an increase of about 20% of the fabrication costs. Fuels with the composition (U 0.55 Pu 0.40 Np 0.05 )O 2 have been prepared successfully at the Institute for Transuranium Elements (ITU) in Karlsruhe for the CAPRA irradiation campaign. Only very small amounts of americium can be added to UO 2 fuel due to dose limitations during fabrication. The high γ-dose is not only due to americium itself but also due to the fact that some lanthanides will be present as impurities. As a consequence, incineration of americium on a large scale can be done most effectively in an inert-matrix fuel which has to be prepared in specially designed facilities with a very high level of shielding and remote handling. Similar materials as for the fuels for plutonium incineration (see above) are being considered as inert-matrix for americium. However, the properties of americium oxide (AmO 2 ) are not very favourable: it has a poor thermal conductivity, a high oxygen potential and reacts with liquid sodium. Recycling of metal and nitride forms containing or not an inert support material might, therefore, be more advantageous for the incineration of americium and possibly other MAs. This option is being studied at JAERI. Recycling of pure curium in dedicated systems does not seem feasible at the moment because of the very high α, γand neutron radiation due to decay and spontaneous fission. One option is interim storage of curium for about 100 years, after which the relatively short-lived curium isotopes ( 242 Cm, 243 Cm and 244 Cm) have decayed to plutonium isotopes which can then be recycled as described above. However, an effective separation method of Am from Cm is a major prerequisite for some scenarios. But recycling of a mixture of Am and Cm is still being considered. 2.3.3 Target selection for fission products If transmutation of the fission products technetium and iodine is considered, they will most likely be irradiated in special targets. Research on the selection of suitable materials as well as pilot irradiation experiments have been done in the frame of the European EFTTRA collaboration. The preliminary results of these studies have shown that metallic technetium can be used as target material: a fabrication route for casting the metal into rods has been developed and irradiation experiments in a thermal spectrum (to a burn-up of about 6%) did not show any evidence for the swelling or disintegration of these rods. Iodine cannot be transmuted in its elemental forms due to the volatility and chemical reactivity. Iodide metal compounds are therefore being considered. The experiments performed in the frame of the EFTTRA co-operation have shown that sodium iodide (NaI) is the best candidate if transmutation of iodine is being considered. 38
Page 1 and 2: TABLE OF CONTENTS EXECUTIVE SUMMARY
Page 3 and 4: TABLE DES MATIÈRES NOTE DE SYNTHÈ
Page 5 and 6: PART II. TECHNICAL ANALYSIS AND SYS
Page 7 and 8: 4. IMPACT OF P&T ON RISK ASSESSMENT
Page 9 and 10: Figure II.31 Evolution of the expec
Page 11 and 12: Part II: Technical analysis and sys
Page 13 and 14: There are several scenarios which c
Page 15 and 16: eactor concepts are still in the co
Page 17 and 18: intermediate storage management, th
Page 20 and 21: 1. INTRODUCTION 1.1 Involvement of
Page 22: and natural decay play an important
Page 25 and 26: Figure I.2 A schematic diagram of b
Page 27 and 28: Instead of recycling, one could ado
Page 29: to address there is the separation
Page 33 and 34: Figure I.3 A schematic diagram of t
Page 35 and 36: Figure1.5 A notional materials flow
Page 37 and 38: A few specific regulatory and safet
Page 39 and 40: • irradiation of FR-fuel in Fast
Page 41 and 42: dispersion in the geosphere or bios
Page 43 and 44: In the meantime the burn-up of spen
Page 45 and 46: Any reprocessing campaign of spent
Page 48 and 49: 4. CRITICAL EVALUATION • P&T may
Page 50: 5. GENERAL CONCLUSIONS • Fundamen
Page 54 and 55: NOTE DE SYNTHÈSE ET PORTÉE DU RAP
Page 56 and 57: Présentation générale Cette part
Page 58 and 59: courts. L’application de cette te
Page 60 and 61: éalisable, à condition d’augmen
Page 62: PREMIÈRE PARTIE : PRÉSENTATION G
Page 65 and 66: La troisième réunion internationa
Page 67 and 68: 1.5 Objectifs du rapport Dans l’e
Page 69 and 70: Figure I.1 Schéma de principe du c
Page 71 and 72: • l ’241 Am est le précurseur
Page 73 and 74: plutonium et environ 2 m 3 de déch
Page 75 and 76: 2.3 Technologie de fabrication des
Page 77: cadre de la coopération EFTTRA ont
Page 80 and 81:
On peut voir sur la Figure I.4 les
Page 82 and 83:
Figure I.4 Flux de matières dans u
Page 84 and 85:
À court terme, les produits de fis
Page 86 and 87:
Par conséquent, au cas où l’on
Page 88 and 89:
De nombreux laboratoires dans le mo
Page 90 and 91:
usé devrait représenter environ 3
Page 92 and 93:
le nucléide le plus gênant est le
Page 94 and 95:
transuraniens. Pour obtenir un taux
Page 96 and 97:
On peut considérer des opérations
Page 98 and 99:
devrait en principe ouvrir de nouve
Page 101:
5. CONCLUSIONS GÉNÉRALES • La m
Page 105:
PART II: TECHNICAL ANALYSIS AND SYS
Page 108 and 109:
1.1.1.1 Minor actinides Americium a
Page 110 and 111:
thus preventing its dispersion in t
Page 112 and 113:
By contrast, information about the
Page 114 and 115:
DIDPA [5] (see Figure II.3) process
Page 116 and 117:
The generation of secondary effluen
Page 118 and 119:
Figure II.5 TRPO process TRPO solve
Page 120 and 121:
According to Jarvinen et al. in LAN
Page 122 and 123:
curium. • Separation of americium
Page 124 and 125:
1.1.4.4 Separation of technetium an
Page 126 and 127:
The second option is production of
Page 128 and 129:
Figure II.9 Fuel cycle actinide bur
Page 130 and 131:
Figure II.11 Flow sheet of pyro-rep
Page 132 and 133:
The metathetical reaction between L
Page 134 and 135:
This is confirmed by the radiotoxic
Page 136 and 137:
For the same burn-up as in the pure
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planned for a burn-up range of 1.5
Page 140 and 141:
On the basis of the study, it is no
Page 142 and 143:
In a given reactor system, the diff
Page 144 and 145:
- deterioration of the effectivenes
Page 146 and 147:
Table II.5 Mass balances for homoge
Page 148 and 149:
manufacture is 2 years. 12×24 targ
Page 150 and 151:
Figure II.14 MA-loading methods in
Page 152 and 153:
Table II.7 Mass balances for homoge
Page 154 and 155:
Table II.8 Mass balances for hetero
Page 156 and 157:
Core characteristics above: The fol
Page 158 and 159:
Figure II.15 Concept of double stra
Page 160 and 161:
Figure II.16 Concept of accelerator
Page 162 and 163:
In Reference [99], the sodium coole
Page 164 and 165:
In Germany, some small activities r
Page 166 and 167:
OECD/NEA programmes The OECD/NEA Nu
Page 168 and 169:
As a part of MA nuclear data evalua
Page 170 and 171:
Table II.13 Pu and minor actinide b
Page 172 and 173:
2.4.1.3 Transmutation in light wate
Page 174 and 175:
3. DESCRIPTION OF CURRENT TRENDS IN
Page 176 and 177:
The SPIN programme studied various
Page 178 and 179:
• the RP1-2 scenario is compared
Page 180 and 181:
Mass balance The MA mass balance, f
Page 182 and 183:
4. IMPACT OF P&T ON RISK ASSESSMENT
Page 184 and 185:
For deterministic effects, in the c
Page 186 and 187:
• accounting for a number of safe
Page 188 and 189:
Table II.17 Effective dose coeffici
Page 190 and 191:
MOX fuel and recycling of recovered
Page 192 and 193:
Figure II.22 Potential radioactivit
Page 194 and 195:
Moreover, plutonium recycling and m
Page 196 and 197:
Curium is assumed to be stored for
Page 198 and 199:
Figure II.25 shows the radiotoxicit
Page 200 and 201:
decrease is obtained in the radioto
Page 202 and 203:
4.4 Risk and hazard assessment over
Page 204 and 205:
4.4.2 Radiological impact of waste
Page 206 and 207:
0.191 man-Sv/TWhe and the RFC with
Page 208 and 209:
• conventional reprocessing does
Page 210 and 211:
The maximum inventory of reactor co
Page 212 and 213:
pursued. Since 137 Cs and 90 Sr are
Page 214 and 215:
Figure II.26 Surface facilities. Ge
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Doses are controlled by 36 Cl up to
Page 218 and 219:
Figure II.30 Overview of the canist
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surface. The reference repository i
Page 222 and 223:
Figure II.34 Cavern-convection scen
Page 224 and 225:
Figure II.36 Normal evolution scena
Page 226 and 227:
considerable energy release as a re
Page 228 and 229:
• taking into account the potenti
Page 230 and 231:
• on the subject of transmutation
Page 232 and 233:
[12] Arnaud-Neu, F., et al., “Cal
Page 234 and 235:
Fuel”, Proc. Int. Conf. on Future
Page 236 and 237:
[64] Arai, T., Suzuki, Y., Handa, M
Page 238 and 239:
[88] Murata, H., and Mukaiyama, T.,
Page 240 and 241:
[117] Gudowski, W., “Accelerator-
Page 242 and 243:
[146] D’angelo, A., Marimbeau, P.
Page 244 and 245:
[172] OECD/NEA and IAEA, Uranium: R
Page 246:
[202] Schmidt, E., Zamorani, E., Ha
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