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3.2.3.5 Quantitative recycling of U and Pu into LWR-MOX fuel Plutonium separated during reprocessing of LWR-UO 2 fuel is in some countries transferred to the reactor as LWR-MOX. However, the irradiation of LWR-MOX increases the radiotoxic inventory of the fuel because of the increased production of MAs. Multiple recycling of plutonium in LWR-MOX is not efficient. In the recent past, consideration was given to using the separated plutonium as a feedstock for FR-fuel where the “incineration” of even as well as uneven numbered isotopes resulted in an increased fission rate compared to that in LWRs. The reduction of the radiotoxic inventory of nuclear materials can only be accomplished by using transmutation in fast spectrum devices (FRs or ADS). Recycling of reprocessed uranium in LWR-fuel is done industrially but entails an increase of the fissile material enrichment in the fresh fuel. From pure radiotoxic point of view, the stockpiling of depleted uranium and of reprocessed uranium has a greater impact than e.g. neptunium. 3.2.3.6 Reprocessing of LWR-MOX Up to now the reprocessing of LWR-MOX has mainly been done by diluting the LWR-MOX fuel with LWR-UO 2 fuel according to the ratio in which it occurs in the reactor-core (UO 2 /MOX = 2). Reprocessing of spent LWR-MOX without dilution in UO 2 fuel has been demonstrated at Cogéma La Hague UP2 plant in 1992 as a special campaign (~5 t) and can be performed industrially if the reprocessing plant has been licensed for the treatment of increased plutonium concentrations and a much higher total plutonium inventory. The radiotoxic inventory of spent LWR-MOX fuel is about 8 times higher than that of spent LWR-UO 2 . Conventional reprocessing will remove U+Pu which accounts for about 30% of the total α- activity and the residual 70% enters into the HLLW made up of Np, Am and Cm. However, the Cm and Am isotopes constitute the overwhelming majority of this α-activity. In a perspective of P&T, it would be indispensable to remove the TRUs from the HLLW before vitrification. The techniques to be used are in principle the same as for the LWR-UO 2 fuel (see 3.2.3.2 Separation of MAs from HLLW resulting from spent LWR fuel), but the higher α-activity level will interfere with the extraction because of increased radiation damage. Another option is to store the spent LWR-MOX fuel for example during 50 or more years and to let 244 Cm decay (18 year half-life) to 240 Pu before carrying out the reprocessing. The chemical extraction processes are much easier to perform after the extended “cooling” period as the α-decay heat is reduced by a factor of 7 or more, depending on the isotopic composition. Multiple recycling of LWR-MOX is possible if sufficient fresh plutonium from reprocessing of LWR-UO 2 with moderate burn-up is available. 3.2.3.7 FR-MOX fuel fabrication with limited MA content The largest industrial experience has been gained in the FR-MOX fuel fabrication since for several decades FR programmes were undertaken in many nuclear countries [12]. The fabrication of FR-MOX fuel with 15 to 25% Pu has been realised routinely and on a commercial basis. But the plutonium quality used for these purposes was derived from low burn-up UO 2 fuel with low 238 Pu and 241 Pu contents. 49
In the meantime the burn-up of spent LWR-UO 2 and LWR-MOX has reached 50 GWd/tHM. The isotopic composition of plutonium resulting from the reprocessing of such fuels is seriously degraded, with high 238 Pu and 242 Pu levels, and low 239 Pu and 241 Pu concentrations. A thorough study of the economic issues involved was published in 1989 by <strong>OECD</strong>/NEA and is still valid as general reference [13]. In a perspective of the use of advanced FBuRs (CAPRA) still higher plutonium concentrations are envisaged (up to 45%). The recycling of fuels containing high 238 Pu levels and limited amounts of MAs is still more difficult and requires the design and construction of remotely operated fuel fabrication plants. For homogeneous recycling of MAs in FR-MOX, admixtures of 2.5% 237 Np and/or 241 Am are currently studied. The specific activity of 237 Np, an α-emitter, is low and there is no major handling problem involved but the admixture of 241 Am at the 2.5% level will induce a γ-field around the glove-boxes or hot cells. However, the major interfering nuclide is 238 Pu at the 3% level which is a heat and neutron source (5 kWth/tHM; 5.10×10 8 neutrons/s tHM). The FR-MOX fuel fabrication with limited MA admixture will also be influenced by the degree of separation of the REs (strong γemitters) and last but not least by 244 Cm which will accompany 241 Am and 243 Am when separated from HLLW. The presence of small amounts of 244 Cm will increase the neutron emission of the resulting fresh FR-MOX fuel. The separation coefficients from REs and 244 Cm required in order to permit industrial fuel fabrication operations will greatly depend on the permissible RE concentration acceptable in fresh FR- MOX fuel and on the permissible 244 Cm concentration during the fuel fabrication process. Heterogeneous recycling of MAs is a means to avoid the dilution of troublesome nuclides, e.g. 244 Cm, throughout the fuel fabrication step and carry out this operation in small, but dedicated and well shielded facilities. 3.2.3.8 Metal fuel fabrication for ALMRs and advanced fuels for burner reactors In the framework of the Integral Fast Reactor (IFR) project a specific fuel fabrication technology has been developed and tested on cold (and hot) pilot scale. At the EBR-II facility metal fuel was recycled by casting a U-Pu-Zr alloy on laboratory and hot pilot scale. It is obvious that these processes are still in the exploratory stage and cannot be considered as proven technology but their potential should be investigated since metal fuel permits very high burn-ups and has good material and neutronic characteristics for transmutation of TRUs. Very recently, attention was drawn on the potential of nitride and carbide fuels for FBuRs. Nitride TRU fuel containing macroscopic quantities of MAs can be produced by a combination of an internal gelation method and a carbothermic synthesis. These nitride fuels can be reprocessed by electrorefining methods similar to the technology developed for metal fuel. 50
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 and 30: to address there is the separation
Page 31 and 32: improvement of the biological shiel
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: dispersion in the geosphere or bios
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
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This is confirmed by the radiotoxic
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For the same burn-up as in the pure
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planned for a burn-up range of 1.5
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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
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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
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decrease is obtained in the radioto
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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
Page 216 and 217:
Doses are controlled by 36 Cl up to
Page 218 and 219:
Figure II.30 Overview of the canist
Page 220 and 221:
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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