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For the same burn-up as in the pure Pu recycling case, the computation has been repeated for the sum of plutonium and americium. The enrichments required in this Pu+Am case are still higher than the enrichments needed to recycle only the plutonium [56]. Recycling of americium together with plutonium in PWRs is limited due to the deterioration of the void reactivity coefficient Fuel fabrication aspects of plutonium and americium recycling in PWRs A study has been made on radiation dose protection and criticality safety in a MOX fuel refabrication plant. As a reference case, the MOX fabrication plant of Belgonucléaire at Dessel [57] has been taken. It was already known that the front-end stages are the most critical ones with respect to handling operations on the pure oxide powders. Currently, PuO 2 powders from the reprocessors are taken from storage. Subsequently, these powders are introduced into a glove box to be milled and blended together with UO 2 powders to produce the so-called primary blend. Such dose intensive operations are followed by secondary blending, pressing and sintering, before the sintered pellets are put into fuel pin cladding: dilution first and canning afterwards lower the dose rates. Therefore, primary blending is retained as the most typical source of the dose rate due to handling operations. Neutron and γdose rates at a distance of 30 cm from the external glove box wall have been calculated for a given configuration of blending devices. The dose rates at 30 cm from the glove box have been calculated for the reference PuO 2 powder (MOX 1; 7.3% Pu) and for the various powders, which correspond either to a second plutonium recycling step (MOX 2; 10.4% Pu) or to a first recycling of plutonium and americium (9.8%; MOX 1-Am). A value of 20 µSv/h (2 mrem/h) is taken as a guiding value for these comparisons, although it is no real limit in the plant as the staff will not stay longer than needed near the glove box of primary blending, according to the ALARA principle. The results [58] show that conditions for the MOX 2 fuel fabrication with 10.4% Pu are very similar to those of the MOX 1 with 7.3% Pu. There is indeed only a marginal increase in the main neutron source, which is induced by the α radiation from 238 Pu by (α,n) reactions. This favourable result is caused by the dilution of the plutonium from the MOX fuel with the plutonium from UO 2 fuel. Addition of americium to the plutonium powder for its recycling would induce a strong increase of the γdose (by a factor of 4.5). This increase could be mitigated by the addition of shielding with a layer of 25 mm of steel. It thus appears that dose rates could be controlled at the expense of extra shielding. Of course, the extra shielding would hinder the fabrication and would increase the cost of operations. Still the operations seem a feasible extension of the standard MOX fabrication conditions especially if one were to consider remote fabrication with automated processes. 2.2.1.3 Fabrication of MOX fuel containing MAs for fast reactor The first experience of fabrication of oxide fuels containing high contents of Am and Np (up to 20% of Am) has been done by the Institute of Transuranium Elements Karlsruhe (ITU) during 1984-1986. The fabrication process used was the SOL-GEL one (GSP, Gel-Supported process) followed by pressing and sintering of the spherical particles in order to obtain pellets. These fuels have been irradiated in the reactor Phénix (SUPERFACT Experiment) [59]. 142
In France, one complete subassembly has been fabricated successfully by Cogéma using industrial facilities in 1997, the fuel was (U, Pu)O 2 containing 5% of Np. Currently, laboratory studies on fabrication of innovative fuels for MA transmutation are performed in the ATALANTE facility at Marcoule (CEA). The CERCER manufacturing process qualification is in progress based on the mechanical mixing of MgO and AmO 2 and granulation. The preliminary tests have been done during 1998. The pellets to be fabricated will be irradiated in Phénix (ECRIX experiments) in 2000. A systematic programme has been planned in JNC for fabrication and investigation of irradiation behaviour of MOX containing MAs. Two fabrication methods, pellet-pressing and vibro-packing have been studied for neptunium-based fuel pins. The pellet type Np-based fuel will be fabricated at Tokai Works of JNC, and the fabrication of Np-based fuel by vibro-packing will be performed at PSI in collaboration with JNC. For Am-based fuels, the Alpha-Gamma Facility (AGF) at Oarai Engineering Center of JNC has already been adapted to fabricate MOX fuel pins containing Am at first and Am and Np afterwards. Remote assembling will be conducted in the Fuel Monitoring Facility (FMC). Both facilities will provide test beds for the post irradiation examination. Irradiation of Np- and Am-containing MOX fuel is planned in JOYO. In step with the JOYO MK-III schedule, the irradiation test will be initiated from around 2003. 2.2.1.4 Fabrication and irradiation of metal alloy fuel including MAs Since the 1960s, Argonne National Laboratory (ANL) has been engaged in developing metal alloy fuels. The initially-developed U-5% fissium alloy fuels with 85% smear density for commercial use were found to fail by swelling. In the 1970s, by lowering smear density to 75% and by increasing the plenum gas volume, over 10 at% burn-up was attained on U-5 wt% fissium and U-10 wt% Zr alloy fuels. As plutonium-containing alloy, U-Pu-10 wt% Zr alloy was selected as having a high melting point and compatibility with stainless steel cladding. Since 1984, U-Pu-Zr alloy fuels have been further investigated as part of the IFR Programme [43]. More than a thousand fuel pins were fabricated by injection casting and irradiated in EBR-II, some of them to a maximum of 18 at% burn-up with cladding temperature
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TABLE OF CONTENTS EXECUTIVE SUMMARY
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TABLE DES MATIÈRES NOTE DE SYNTHÈ
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PART II. TECHNICAL ANALYSIS AND SYS
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4. IMPACT OF P&T ON RISK ASSESSMENT
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Figure II.31 Evolution of the expec
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Part II: Technical analysis and sys
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There are several scenarios which c
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eactor concepts are still in the co
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intermediate storage management, th
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1. INTRODUCTION 1.1 Involvement of
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and natural decay play an important
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Figure I.2 A schematic diagram of b
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Instead of recycling, one could ado
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to address there is the separation
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improvement of the biological shiel
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Figure I.3 A schematic diagram of t
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Figure1.5 A notional materials flow
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A few specific regulatory and safet
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• irradiation of FR-fuel in Fast
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dispersion in the geosphere or bios
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In the meantime the burn-up of spen
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Any reprocessing campaign of spent
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4. CRITICAL EVALUATION • P&T may
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5. GENERAL CONCLUSIONS • Fundamen
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NOTE DE SYNTHÈSE ET PORTÉE DU RAP
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Présentation générale Cette part
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courts. L’application de cette te
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éalisable, à condition d’augmen
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PREMIÈRE PARTIE : PRÉSENTATION G
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La troisième réunion internationa
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1.5 Objectifs du rapport Dans l’e
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Figure I.1 Schéma de principe du c
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• l ’241 Am est le précurseur
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plutonium et environ 2 m 3 de déch
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2.3 Technologie de fabrication des
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cadre de la coopération EFTTRA ont
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On peut voir sur la Figure I.4 les
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Figure I.4 Flux de matières dans u
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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 138 and 139: 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
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• accounting for a number of safe
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Table II.17 Effective dose coeffici
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MOX fuel and recycling of recovered
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Figure II.22 Potential radioactivit
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Moreover, plutonium recycling and m
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Curium is assumed to be stored for
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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
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4.4.2 Radiological impact of waste
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0.191 man-Sv/TWhe and the RFC with
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• conventional reprocessing does
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The maximum inventory of reactor co
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pursued. Since 137 Cs and 90 Sr are
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Figure II.26 Surface facilities. Ge
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Doses are controlled by 36 Cl up to
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Figure II.30 Overview of the canist
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surface. The reference repository i
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Figure II.34 Cavern-convection scen
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Figure II.36 Normal evolution scena
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considerable energy release as a re
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• taking into account the potenti
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• on the subject of transmutation
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[12] Arnaud-Neu, F., et al., “Cal
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Fuel”, Proc. Int. Conf. on Future
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[64] Arai, T., Suzuki, Y., Handa, M
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[88] Murata, H., and Mukaiyama, T.,
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[117] Gudowski, W., “Accelerator-
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[146] D’angelo, A., Marimbeau, P.
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[172] OECD/NEA and IAEA, Uranium: R
Page 246:
[202] Schmidt, E., Zamorani, E., Ha
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