COMPLETE DOCUMENT (1862 kb) - OECD Nuclear Energy Agency

More documents

Recommendations

Info

planned for a burn-up range of 1.5 to 6 at% and up to 10 at% as soon as difficulties in restarting the Phénix reactor have been overcome. 2.2.1.5 Fabrication of nitride fuel including MAs Nitride has many advantageous properties as advanced fuel such as high-thermal conductivity, good FP retention, high heavy metal density and mutual solubility. In 1970s and 1980s, laboratory studies on a fabrication process for mixed nitride fuels has been performed at CEA. The fuel was seen as the best advanced fuel for FR application mainly due to its higher breeding gain and its absence of fuel/sodium reaction in the case of clad failure. The process developed was the carbothermic reduction of mixed oxide in an atmosphere of nitrogen, followed by decomposition of higher nitrides to mononitride. This dry preparation process was the most capable for industrial applications. Nitride pellet fuel pins have been fabricated for irradiation test (NIMPHE programme) in the Phénix reactor. The ITU fabricated, at the same time, nitrides by sol-gel process for irradiation in the HFR reactor (NILOC experiments). The fabrication technology for mixed nitride fuel based on the experience of carbide fuel has been developed, in both the Paul Scherrer Institute in Switzerland [62] and the Bhabha Atomic Research Center in India [63]. During the past 10 years, research on fabrication technology for mixed uranium-plutonium nitride fuel has been performed at JAERI with a view to an advanced fuel cycle system. Nitride pellets have been fabricated and supplied for measurement of their characteristics and for irradiation tests [64]. PNC has also fabricated some nitride fuel pellets and measured the fuel characteristics. Pellet fuel In 1960s and 1970s, actinide mononitride was mainly synthesised by a hydriding-nitriding route from the actinide metal, but recently the conversion of oxide to the nitride by carbothermic reduction has been improved [65,66]. Uranium-plutonium mixed nitride and neptunium-bearing nitride have been synthesised by reduction of the dioxide with graphite, usually in a nitrogen-hydrogen mixture. Pellet-type fuel is fabricated by milling, compacting and sintering in a similar way to MOX fuel. No important problems appear to remain except in the enrichment and recycling of 15 N. The use of 15 N would be preferable because of the massive formation of 14 C by the 14 N(n,p) 14 C reaction. Both 15 N and 14 C would be difficult to retain during the nitride dissolution in the PUREX reprocessing, while they could be easily recovered during the pyrochemical reprocessing with fused salt electrorefining. In view of the cost for the enrichment of 15 N, the content of which is only 0.365% in natural nitrogen, the recycling of 15 N would be one of major issues for feasibility of nitride fuel. Particle fuel Remote fabrication from particles has advantages with materials of high radioactivity, so techniques for the fabrication of nitride particles by a sol-gel process have been developed [62-67]. Minor actinides separated from the high-level waste (HLW) as nitrates are converted into solids. The feed solution is prepared by mixing actinide nitrate solution, carbon powder, hexamethylenetetramine (HMTA) and urea. HMTA decomposes to form ammonia on heating to about 80-100°C, so gel particles are formed as a mixture of the actinide oxide and carbon, which is converted to mononitride by carbothermic reduction in N 2 -H 2 mixture. The particles of pure UN with low oxygen and carbon contents have been successfully fabricated in Japan and Switzerland, and the method has also been 144
applied to uranium-plutonium mixed nitride. The technology for the fabrication of dense and pure nitride particles would be a future R&D subject. Targets with an inert matrix Nitride is also suitable as target compound for burning americium by diluting it with inert materials such as ZrN and TiN. The fabrication of the solid solution of ZrN and UN, which is a substitute of AmN, has been studied by a sol-gel route [62]. Such fuel is also considered to be a candidate for plutonium burning in the CAPRA reactor [68]. 2.2.2 Targets for heterogeneous recycling 2.2.2.1 Fabrication of AmO 2 target pins A selection of inert-matrix material, actinide support alone (Am based), and compound materials, heterogeneous Am targets have been studied for Am transmutation [69]. The selection of possible candidates was based on a number of criteria concerning their neutronic, physical and chemical properties in relation to fabrication, performance and reprocessing. The following classes of conceivable candidates were investigated: • pure Am compounds: Am 2 O 3 , AmO 2 , AmN and Am 2 C 3 ceramics. • ceramic solid solution materials: they are obtained by a chemical reaction between Am compound and the inert material. The final product forms a single phase which is physico-chemically different from that of the initial Am compound and from the inert material used. • two-phase materials: are compounds of two distinct phases in which the two components keep their chemical form. In this case, Am compound is dispersed in the inert material, either in the form of fine particles or in the form of macro-particles. Two categories are to be distinguished: • the CERCER material (dispersion of a Am CERamic in an inert CERamic) of oxide, nitride or carbide; • the CERMET material (dispersion of an Am CERamic in an inert METal). It was found that MgO, Y 2 O 3 , Al 2 O 3 , MgAl 2 O 4 , and Y 3 Al 5 O 12 are good potential candidates as inert matrices for fast reactor. CERCER composite materials with Am, MgO-AmO 2-x [70] have been fabricated in ITU, after sintering the pellets are a two-phases mixture. MgAl 2 O 4 -AmO 2-x pellets containing 10% 241 Am by weight have been fabricated by the impregnation method by ITU. A reaction between the Am oxide and MgAl 2 O 4 occurs to form a new compound. Taking into account the difficulties in preparing and controlling Am oxides of well-defined composition, it may be advantageous to use a dispersion of a (M, Am)O 2 , (M, Am) 2 O 3 type solid-solution in MgAl 2 O 4 [71]. 145
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 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
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
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
show all

COMPLETE DOCUMENT (1862 kb) - OECD Nuclear Energy Agency

You also want an ePaper? Increase the reach of your titles

Delete template?

Save as template?