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International Conference on the Physics of Reactors “Nuclear Power: A Sustainable Resource” Casino-Kursaal Conference Center, Interlaken, Switzerland, September 14-19, 2008

Minor actinide transmutation in ADS: the EFIT core design

C. Artiolia,*, X. Chenb , F. Gabriellib , G. Glinatsisa, P. Liub, W. Maschekb, C. Petrovicha, A. Rineiskib , M. Sarottoa, M. Schikorrb

a ENEA, Via Martiri di Montesole 4, IT-40129 Bologna, Italy b Forschungszentrum Karlsruhe (FZK), P.O. Box 3640, D-76021 Karlsruhe, Germany

Abstract

Accelerator-Driven-Systems represent one of the possible future strategies for transmuting minor actinides.

EFIT, the conceptual industrial burner designed in EUROTRANS IP, is an ADS of about 400 MW th, fuelled by MA and Pu in inert matrix, cooled by lead (673-753 K) and sustained by a 800 MeV proton of some 15 mA. It features the MA fission (42 kg/TWhth) while maintaining a zero net balance of Pu and a negligible k eff swing during the cycle. Three radial zones, differing in pin diameter or in inert matrix percentage have been defined in order to maximize the average power density together with the flattening of the assembly coolant outlet temperatures. Thermal-hydraulic analyses have been performed and show acceptable maximum temperatures: 1672 K peak fuel temperature (disintegration at 2150 K) and 812 K peak cladding temperature in nominal conditions (max 823 K). The behaviour of the core power, the temperature and the reactivity during the Unprotected Loss Of Flow transient (ULOF) has been studied as well by obtaining: a peak fuel temperature of 1860 K, a peak cladding temperature of 1030 K, a power increase of 2% removed by natural circulation.

1. Introduction conceptual design of an ADS (Domain DM1 of the Integrated Project EUROTRANS). This project is The sustainability and the public acceptance of called EFIT (European Facility for Industrial nuclear energy production can be improved by the Transmutation) and investigates the feasibility and minimization and reduction of nuclear waste. The the potentiality of such systems (Knebel, 2006). The Minor Actinides (MA) have a long-term radio- design will be worked out to a level of detail which toxicity and one of the possible future strategies for allows a cost study estimate. EFIT, of about 400 transmuting them is represented by the use of MWth, is loaded with MA and Pu in a CERCER U- Accelerator Driven Systems (ADS), which allow a free fuel. The core coolant, allowing a fast spectrum, higher MA content in the fuel. On the other side, the is pure lead, as well as the windowless target for the cost/benefit ratio of such innovative systems has to 800 MeV proton beam. The reference sub-critical be evaluated and challenging coordinated R&D is level has been postulated to be keff=0.97, figure that necessary. has to be confirmed by the full safety analysis Within the 6th Framework Program, the (Rimpault, 2006). European Community has funded, besides other This paper deals with the neutronic and projects supporting partitioning and transmutation, a thermal-hydraulic design of the EFIT core (Artioli et

* Corresponding author, [email protected] Tel: +39 051 6098436; Fax: +39 051 6098279.

1 al., 2007a; Barbensi et al., 2007). The core has been CERCER and CERMET have been assessed and conceived with the aims of: maximizing the fissions finally as matrices the materials ZrO2, MgO and Mo of MA, achieving a negligible keff swing during the had been under closer investigation. The final cycle (to keep the proton current rather constant in recommendation on fuels gave a ranking of these order to avoid an oversizing of the target and of the fuels based on a number of criteria, ranging from accelerator), maximizing the average power density fabrication, reprocessing via economics to safety. (i.e. the volume density of MA transmutation), The composite CERMET fuel (Pu0.5,Am0.5)O2-x – while keeping low the coolant pressure drop. 92Mo (93% enriched) has been recommended by For EFIT, as for any kind of reactors, the AFTRA as the primary candidate for the EFIT defence-in-depth concept has been applied. The (Maschek, 2008). This CERMET fuel fulfils demonstration of the adequacy of design with the adopted criteria for fabrication and reprocessing, safety objectives is structured along three kinds of and provides excellent safety margins. basic conditions: The Design Basis Conditions Disadvantages include the cost for enrichment of (DBC–structured into 4 Categories), Design 92Mo and a lower specific transmutation rate of Extension Conditions (DEC–limiting events, minor actinides, because of the higher neutron complex sequences and severe accidents) and absorption cross-section. The composite CERCER Residual Risk Situations. For the EFIT the safety fuel (Pu0.4,Am0.6)O2-x – MgO has therefore been principles and safety guidelines have been defined recommended as a backup solution as it might offer within EUROTRANS and a comprehensive and a higher consumption rate of minor actinides, and representative list of transients has been defined to can be manufactured for a lower unit cost. In the test the safety behaviour of the reactor plant. For EFIT development the demonstration of an efficient innovative reactors such as the EFIT ADT cliff-edge transmutation performance is a key issue. Therefore effects should be identified and excluded. For a the DM1 design concentrated on the CERCER core safety classification fuel limits related to the first, the more as preliminary analyses showed the different safety categories have been defined based compliance with normal operation and safety on recent experimental evidence. Due to the existing criteria. uncertainties, fuel melting or disintegration should The fabrication of composite pellets is only be allowed in the DEC category. Important considerably more difficult than solid solution oxide boundary conditions to be taken into account in the pellets. This is a result of the specific requirements safety evaluation are the significant positive void of size and homogeneous distribution of the worth, the missing of the Doppler prompt reactivity dispersed actinide phase. The fuel development for feedback, the very low delayed neutrons effective AFTRA is performed at CEA and at the Institute for fraction (Artioli et al. 2007a) and the strong Transuranium Elements (ITU). In the framework of production of He via the transmutation process. the EFIT design the fissile phase volumetric content While coolant boiling processes can be of these fuels is around 50%. The samples made for excluded because of the high boiling point of lead in-pile tests hold less than 30-40% of fissile coolant, pin failures could lead to a gas blow-down particles because of nuclear facility constraints on from the plena, to local voiding and reactivity authorised Minor Actinide contents. The fabrication addition. From the list of transients some route used at the laboratory scale for these highly representative ones, which are also traditionally radioactive materials firstly deals with fissile investigated in fast reactor systems, have been particles preparation by clean and necessarily, dust- chosen for the current paper, as the unprotected loss free fabrication methods to minimize contamination of flow (ULOF). in the gloveboxes. Two processes are used: an oxalic co-precipitation route (Brunon, 2004; Croixmarie, 2003) for CEA, and a combination of 2. The inert matrix external gelation (Fernandez, 2006) and infiltration methods (Fernandez, 1999) for ITU. The following In Europe a vast experience exists on oxide fuels, steps belong to the conventional powder metallurgy therefore the main emphasis of the ADT fuel area: they consist in mixing and grinding the non- development concentrated on the oxide route. In the radioactive powders with the fissile powders. The EUROTRANS Domain AFTRA (DM3) various fuel blends are sieved and pressed. The green pellets are forms as solid solution and/or composites as then sintered. Within the AFTRA framework,

2 CERMET and CERCER pellets dedicated to 42 kg/TWhth is not a result of a design, but a irradiation tests, have been fabricated using the both physical constant. What can make the difference is procedures. Such produced pellets are irradiated either (the measure unit, i.e. kg/TWhth, is here within the framework of FUTURIX-FTA tests in omitted): the Phenix fast reactor (Donnet, 2005). The - the 42 fissioned can be differently split between FUTURIX tests are of central importance for the MA and other heavy nuclides (Pu or U) or development of these dedicated fuels. - along the fissions (the universal 42), events on MA other than fission can occur; so the MA “disappearing” can be actually higher than 42, 3. Conceptual guidelines and rationales that in turn would simply mean the exceeding part has been transmuted in other heavy Dealing with ADS, as with any complex nuclides (i.e. Pu). system, a number of parameters either directly or We can condensate all that in a pair of numbers: the indirectly interlinked ought to be kept first one indicates the overall MA disappearance simultaneously under control. Very often an “eel (either fissioned or transmuted), the second one the effect” occurs: paying attention and acting for new Pu production. Their difference must be in any optimization of some parameters other, not less case 42 (fig. 1, right double-marked axis of the up- important, are moved away and vice-versa. To help right quarter). For instance “65;23” means that 65 for getting a simultaneous vision at glance of the MA disappear and 23 new Pu is produced, i.e. 42 system, the A-BAQUS graph (fig. 1, reported out of 65 MA really fission and the remainder 23 numbers are those typical of the EFIT-Pb system) transmute into new Pu. In this case EFIT acts as a has been proposed (Artioli, 2007b). In the graph converter from MA to Pu (red zone). The some key-parameters (namely burning efficiency, performance for MA depends on the Pu policy fuel enrichment, reactivity swing, active fuel rather than on the MA one! It is easy to recognize volume, power and core size, accelerator proton how the value of the pair is directly ruled by the current and its range along the cycle) are shown as ratio between Pu and MA, i.e. by the enrichment. well as their logical relationships by the mean of Yet this parameter also rules directly the reactivity typical curves, each marked by the referred swing in the cycle (left axis in the top-right quarter), enrichment E (Pu/(Pu+MA)). that in turn drives the range of the accelerator No matter the performances claimed about the current (bottom left quarter). With the MA and Pu MA burning efficiency, it has to be admitted that the vectors assumed in the EFIT design (Rimpault, fission rate is in any case 42 (rounded number) 2006), the above mentioned case would mean: kg/TWhth, that is merely the 200 MeV/fission in enrichment 27% with a Kswing about 0.019 (one year changed units of measure, in any nuclear system cycle). (thermal, fast, low, high flux; soft, hard spectrum; In an ADS the unit of energy, one fission for small, huge size, with any coolant, etc.). instance, is largely more costly than in any nuclear power plant: then fissioning Pu in ADS would prove to be an uneconomic use of the fuel. On the other hand the EFIT fundamental choice of the inert matrix implies that new Pu production has to be avoided. Therefore the Pu balance should be 0, that leads to the “42;0” pair. Of course it does not mean that every MA atom belonging to the “disappeared 42” is directly fissioned: a good part is transmuted in Pu and in the meantime a same amount of Pu is fissioned. Should a different Pu policy be chosen, either Pu burning (<42 for MA) or Pu producing (>42 for MA), it would be easily reached in EFIT. In the graph is shown as the selected pair “42;0” implies a 45.7% enrichment, whose expected reactivity swing is some 200 pcm/year. Fig. 1. A-BAQUS graph.

3 The right-bottom quarter allows to deal with the fission products are unloaded. Preliminary analyses core size, keeping the selected performance “42;0”. show that this is a possible scenario with EFIT. Of Moving on the referred curve E=45.7% a core course for that purpose an equilibrium composition power size can be selected acting on the active fuel has to be reached, in which the equilibrium vector of fraction, marked on the abscissa as the complement the plutonium is quite different from the beginning content of the inert matrix. Of course these right- one (i.e. richer in even isotopes and poorer in odd bottom-quarter relationships are driven by the ones). Nevertheless, an equilibrium enrichment thermal-hydraulic setting of the core, namely the exists (about 60-70%) and, more important thing, linear power rating, the enthalpy equation, the such a mixture ought to have enough reactivity to coolant velocity. sustain an EFIT core. In the left-bottom quarter, current and its swing This paper deals with the EFIT start-up core. are reported for the selected enrichment, and The first step in designing the core has been the therefore performance, according to the core size. definition of the unique enrichment that fits the The burning capability is expressed in terms of “42;0” approach. Keeping constant this pair, a kg/TWhth or/and in terms of “percentage of the suitable optimization of the core can be pursued inventored MA/year”. In the EFIT this rate is arranging the volumetric fractions and the geometry 4.5%/year. For a coherent comparison it has to be in order to reach the desired keff (0.97) (Barbensi et kept in mind that in EFIT the 4.5% are actually al. 2007) and to flat the radial distribution, both for fissioned (and not partially transmuted in other economy and for respecting the technological heavy isotopes). Since this rate depends only on the constraints, mainly Tclad max 823 K, Tfuel max 1650 MA cross sections and flux intensity, the only way K (500 K below the disintegration temperature of to claim a better figure is to have a higher flux the inert matrix; Maschek et al., 2008). (and/or a more effective spectrum). It is important to note that, being the Pu content The rate of percentage has directly an economic rather constant in the cycle, the reactivity swing will implication: the shorter is the time the cheaper is the not be large. This allows to keep a rather constant process. But as far as the efficiency is concerned, proton current, avoiding an oversizing of both the what is important is not the percentage/year accelerator and the target module. (velocity of burning), but the percentage at the In the operating conditions, the mean outlet discharge. This last is ruled directly from the max temperature of the coolant (pure lead) of 753 K is allowed BU: if the flux is higher this maximum is rather close to the maximum allowed temperature of reached earlier, but it does not change. the cladding of 823 K (USDOE, 2002). Therefore, The rate 4.5%/year is ruled, via flux intensity, the spread of the outlet temperatures of the by the “external” constraints, as the available target subassemblies, belonging to the same zone of flow cooling system (11 MWth) and the required rate, must have a low peak factor (lower than 1.2 in subcriticality (here postulated to be 0.97). first approximation). To meet this requirement the In the EFIT a prudential max BU of some 100 core is radially subdivided in three zones of flow MWd/kg (HM) has been assumed in first step. The rates, ruled by suitable orificing. final percentage at the discharge is then a In order to flat the radial flux profile, the active satisfactory 13.9%. This figure means that, at every fuel volume fraction is increased along the radius. unit of MA fissioned, reprocessing losses of 7 units Since the “42;0” approach defines univocally have to be associated. the enrichment, to flatten the radial flux profile the Of course a complete characterization of the active fuel VF has been increased along the radius. new fuel, either with the MgO or Mo inert matrix, In detail: could allow higher figure of BU and consequently - from the inner zone to the intermediate one, the higher figure of the percentage of fissioned MA at fuel/matrix ratio has been changed from 43% up the discharge. to 50%, by keeping the same pin diameter and and pitch; - from the intermediate zone to the outer one 4. The EFIT equilibrium core (where the flux and the power density become quite lower anyway and less cooling is required) The actual “perfect MA burner” is the reactor the pin diameter has been increased by keeping where only new MA are used for refueling and only the same pitch and fuel/matrix ratio.

4 spent UO2 but with the storage period of 15 years. With these vectors the enrichment fitting the pair 4.1. Calculation tools for neutronic calculations goal “42;0” has been evaluated and found to be 45.7%. The core has been designed mainly by means of the deterministic code ERANOS (Rimpault, 1997), Table 1 with both a 2D cylindrical and a 3D hexagonal MA and Pu weight compositions MA [w%] Pu [w%] schematization. The Monte Carlo code MCNPX 237Np 3.884 238Pu 3.737 (Hendricks, 2006) has also been used because it 241Am 75.510 239Pu 46.446 allows to transport particles at high energy. 242mAm 0.254 240Pu 34.121 Moreover it can calculate a detailed power 243Am 16.054 241Pu 3.845 distribution with a heterogeneous description of the 243Cm 0.066 242Pu 11.850 fuel assemblies. The whole system has been 244Cm 3.001 244Pu 0.001 modelled for MCNP in a detailed 3D geometry 245Cm 1.139 (including thermal expansions and neutron libraries 246Cm 0.089 at different temperatures). The methodology 247Cm 0.002 followed (Burn, 1999) was thus to use MCNPX to To respect the maximum fuel temperature calculate, starting from the 800 MeV proton beam, allowed, a limiting linear power rating has been the neutron source for ERANOS. The neutron evaluated. Since the pellet thermal conductivity source is defined as the first neutrons appearing in depends on the inert matrix content, a linear power the system with energy below 20 MeV. The spatial rating of 180 W/cm has been found for the pellet and energy distributions of these neutrons are used with 50% of matrix (minimum content, for the as input for ERANOS. intermediate and outer zones) and a rating of 200 The neutron libraries used for the codes are: W/cm for the pellet with 57% of matrix (for the ERALIB1 (Jef2.2) for ERANOS; a combination of inner zone). Jeff 3.1 (NEA, 2006), ENDF/B-VI, LA150 (Chadwick, 1999) for MCNPX. For high energy interactions, the CEM03 physics model (Mashnik, 2006) has been used.

4.2. The core-layout

The chosen structural material is Ferritic- martensitic steel T91, for which a maximum temperature allowed for the clad, taken into account a suitable treatement, is 823 K. At present a residence time of 3 years is considered for the fuel. To limit the corrosion effect and meantime to have a Fig. 2. The 3-zones EFIT core (180 fuel assemblies). low pressure drop through the core, the coolant speed is not higher than 1 m/s. The fuel is a U-free one, with MgO as inert matrix. To assure the thermal conductivity in the pellet, a minimum matrix content of 50% must be used. The isotopic compositions of the used Pu and MA are reported in Table 1. These vectors have been obtained as a result of a mixing of MA coming from the spent UO2 fuel (90%) and the spent MOX (10%) of a typical PWR unloaded at the burnup of 45 MWd/kgHM, then cooled down for a period of 30 years. Plutonium is extracted from the same

5 kS = M / (M+1) 0.95111  0.00059 (1 k ) / k  *  eff eff 0.52 Fig. 3. MA and Pu evolution during the fuel life. (1 k S ) / kS The dimension and the composition of the pin Proton current 13.2 mA and of the fuel assembly is reported in (Artioli, 2007a). While the pin diameter and the pitch derive 4.4. The power distribution from the thermal balance, the fuel assembly dimensions are driven by the size of the spallation The flux radial flattening aims to reduce as module, which has to be inserted replacing the 19 much as possible the power radial form factor central assemblies. The core is shown in fig. 2. within each radial zone, and to reach the maximum The residence time is stated in 3 years, life time power density peaks allowed (corresponding to 200 that allows to reach the peak burn up of about 10% W/cm and 180 W/cm according to the different (Knebel, 2006), within the limit imposed by the matrix content). Figure 4 shows the power density corrosion and well below the dpa limit. This span of radial profile, on the peak plane (about midplane), time is divided into three subcycles 1 year long. Due obtained in a 2D RZ geometry. This flattening has to the rather constant content in Pu during the been further improved by the 3D XYZ model irradiation the reactivity swing is very small, 200 (Artioli, 2007a). pcm/year, i.e. some 6% of the subcriticality (3000 pcm), that accounts for a little spread of the proton current required. Figure 3 shows the mass evolution of the MA and Pu during 3 years of nominal power irradiation. As a consequence of the selected enrichment, 45.7% (“42;0” approach), the mass of the Pu remains rather constant, while only the MA are fissioned. It has to be noted that the reactivity is almost completely sustained by the Pu (some 2450 kg) while the remainder some 2900 kg of MA is actually the target to be fissioned.

Fig. 4. Radial profile of the homogeneous power density. 4.3. The source parameters The overall power (beam excluded) of the core

The main integral parameters (MCNPX results) is 389 MWth, 5 MWth of which are dissipated in at BOC are reported in Table 2. There is a structural zones outside the active core. The discrepancy of 930 pcm in keff between ERANOS obtained average homogeneous power density is (with the ERALIB1-Jef 2.2 library) and MCNPX 70.7 W/cm3. (with the Jeff 3.1 library). The MCNPX results The power deposition distribution has been using the ENDF/B-VI library for the fuel appear to calculated by means of both ERANOS and be more similar to ERANOS (320 pcm of difference MCNPX. The results used as reference for the in keff). Note that the neutron source efficiency is thermal hydraulic analysis are those from MCNPX: *=0.52, while in the PDS-XADS design (Burn, the power has been calculated in each assembly of 2003) was *=0.99 (kS and keff very similar). This the core, separated per ring. From these values, the 3 effect is mainly due to the different fuel composition hottest assemblies in the 3 zones have been and to the larger radius of the target. identified, together with the axial form factors and the value of the heat release in the hottest pin (16.4 Table 2 kW). The maximum linear power in the fuel pins MCNPX results at BOC (the error is the stand. deviation) turns out to be 203 W/cm. The differences with keff 0.97403  0.00023 ERANOS are within 5% for the hottest assemblies Neutron source (S) 23.02  0.08 and within 3% for the axial form factors. As a result (neutrons/proton) of this analysis, better zone contours can be defined, M= all fission neutrons / S 19.45  0.25 mainly for the Intermediate/Outer interface.

6 As far as the power in the target is concerned, clad surface to the coolant. A maximum layer MCNPX calculations show that 73% of the beam thickness of 5-10% of the cladding thickness can be power is deposited in the target circuit. If, during the presumed as a guiding parameter for EOL analysis. life of the system keff is always around 0.97, then the Several oxide layer thicknesses, namely 100, 200, proton current is estimated to be at maximum 15.4 and 300 µm, have been used as a parameter in our mA and the heat deposition in the target at analysis. The thermal conductivity of the oxide layer maximum 9 MW. is assumed to be ~ 1 W/m/K. To assure a uniform pressure drop across the entire core, orificing of core zones 1 and 2 are required. 5. Thermal-hydraulic and transient analysis Table 3 Peak fuel, cladding temperatures (K) and cladding failure 5.1. Nominal conditions times (hours). Nominal conditions. Cladding Oxide Avg Pin Peak Pin failure The thermal-hydraulic analyses of the core were Layer performed with the static version of the SIM-ADS times code (Schikorr, 2001) for each of the 3 core zones. Avg Core Thicknes Peak Pin Two core conditions are analyzed, namely Clad Fuel Clad Fuel Pin zone s (m) (hrs) Beginning-Of-Cycle (BOC) and End-Of-Cycle (hrs) (EOC). The thermal conductivity of the two different BOC 0 778 1493 803 1672 E11 E10 MA-fuel compositions, namely MgO volume Inner EOC 0 778 1097 812 1279 E9 E6 fractions of (CZ1/CZ2/CZ3 = 57%,50%,50%), were (CZ1) calculated based on the known thermal 100 873 1399 7.0E4 conductivities of MgO and MOX-MA-fuels using 200 950 1514 4.5E4 the Bruggeman weighting scheme and applying an 300 1031 1620 1.44 appropriate correction for burnup. More details of this procedure can be found in (Maschek, 2007). BOC 0 776 1515 792 1638 E11 7.0E10 Under BOC conditions, fresh fuel conditions IntermeEOC 0 776 1115 796 1226 6.8E8 2.5E7 are presumed. Under nominal conditions, the size of diate the gap between clad and fuel has closed down to (CZ2) 100 853 1331 1.9E5 about 110 µm for the average pin, or about 70% of 200 923 1433 1.0E3 the cold condition value, and the gas composition in the gap is dominated by He, namely (He/Xe/Kr = 300 995 1531 9.8 0.976/0.023/0.001). BOC 0 770 1406 804 1667 E11 E10 Under EOC conditions, a peak fuel burn-up of about 100 MWd/kg has been assumed for these Outer EOC 0 770 1059 799 1206 6.9E8 5.2E6 calculations. The gap between clad and fuel is (CZ3) 100 844 1298 1.1E5 presumed to be essentially closed (min gap ~ 4 µm) 200 904 1390 1.2E3 and the fission gas composition in the gap is still dominated by He due to the higher helium fission 300 968 1479 17.4 gas production in MA fuel compared to conventional fuel (factor ~3.6 has been calculated), Table 3 summarizes the results of the namely (He/Xe/Kr = 0.781/0.201/0.017). For the calculations performed at nominal operations (100% peak pins, pin pressures of (CZ1/CZ2/CZ3= load). Under nominal BOC conditions, peak fuel 112/116/127) bars are calculated. and peak clad temperatures are well within An additional parameter requiring closer acceptable upper limits for all 3 core zones, namely attention in the thermal hydraulic analysis is the ~ 1650 K for the fuel and about 823 K for the formation of an oxide layer on the cladding material. cladding. Under EOC conditions the acceptability The formation of this oxide layer serves a protective depends on the actual thickness of the oxide layer. function against clad corrosion, on the one side; on Based on the above results, the current Pb- the other side it will impede heat transfer from the cooled EFIT design seems quite viable. Attention

7 needs to be placed however on the operational by the ‘Fuel-Domain’ of EUROTRANS (Maschek et control of the oxygen content in the Pb coolant in al., 2008). Fig. 7 shows that in the ULOF condition order to control chemical fowling and the buildup of the increase of the reactivity and consequently the the oxide layer. power in the core is low.

5.2. ULOF analysis

ULOF, as one of the key accident scenario, has been analyzed by means of the code SIMMER-III (Kondo et al., 1992, Maschek et al., 2005). The “unprotected” means that no beam shut down takes place during the transient. The total pressure loss in the primary system has not been finally decided in the EFIT design group, while currently a total pressure drop of 1.1 bar has been assumed. Meanwhile, the final pump head transient data after the pump coast down are not available yet, a pressure transient curve shown in Fig. 5 has been Fig. 6. Transients of the temperatures in the core. used in the ULOF simulation.

Fig. 7. Transients of the power and reactivity. Fig. 5. Transients of the pump head and the coolant mass flow rate. With the above assumptions, the coolant mass 6. Conclusions flow rate will follow a transient process as shown in Fig. 6. It firstly decreases to about 20% of its initial The MA fission (120 kg/year) via an U-free value when the pump head arrives at zero and lead cooled ADS as EFIT is proved to be viable. finally keeps a value of about 32% with some slight The “42;0” approach assures that every fission is oscillation. Fig. 6 shows that, with the 32% devoted to an atom belonging to MA, while the Pu remained coolant heat removing capacity, the fuel, content is kept constant, acting as a catalyzer. clad, and coolant peak temperatures finally Normal condition and transient (ULOF) stabilized at around 1835 K, 1000 K, and 955 K, analyses show the respect of the technological respectively. It also shows that during the ULOF limits, even if efforts have to be devoted for transient, the highest temperatures that fuel, clad and lowering the total pressure drop as well as for a coolant will experience are about 1860 K, 1030 K, better power distribution flattening. and 985 K, respectively. The clad and coolant The current MA burning rate of 13.9% of the temperatures are well below the failure limit and initially inventoried at the discharge, is strictly ruled also the fuel peak temperature is well below the by the max BU allowed. This figure in turn rules limits for melting and disintegration (2150 K) given directly the total reprocessing losses. Therefore

8 R&D effort has to be devoted to the qualification of Fernández, A., et al., 1999. Preparation of spinel the fuel, to the PCMI as well as to the qualification (MgAl2O4) spheres by y hybrid sol-gel of the steel and its treatment in lead environment. technique. Advances in Science and The use of Mo-92 as more promising inert matrix Technology 15, 167-174. has to be investigated. Fernandez, A., et al., 2006. Overview of ITU work on inert matrix fuels, 9th-IEMPT conference, September 25-29, Nîmes (France). 7. Acknowledgement Hendricks, J.S., et al., 2006. MCNPX Version 26C, LANL report LA-UR-06-7991. The authors thank the partners of the IP- Knebel, J., et al., 2006. EUROTRANS: European EUROTRANS project for their fruitful contribution Research Programme for the Transmutation of to the project. Special thanks to the European High-level Nuclear Waste in an Accelerator- Commission for the financial support through the driven System, Ninth Information Exchange FP5 and FP6 programs. Meeting, Nimes, France. Kondo Sa., Morita, K., Tobita, Y., Shirakawa, N., 1992. SIMMER-III: An Advanced Computer References Program for LMFBR Severe Accident Analysis, ANP'92, Tokyo, Japan, Oct. 25-29, No. 40-5. Artioli, C. et al., 2007a. Optimization of the minor Maschek, W., et al., 2005. SIMMER-III and actinides transmutation in ADS: The European SIMMER-IV Safety Code Development for facility for industrial transmutation - EFIT-Pb Reactors with Transmutation Capability, M&C concept, AccApp’07, Pocatello, U.S.A. 2005, Avignon, France, Sept. 12-15. Artioli, C., 2007b. A-BAQUS: a multi-entry graph Maschek, W., et al., 2007. A Comparative assisting the neutronic design of an ADS, 5th Assessment of Safety Parameters and Core International Workshop on the Utilization and Behavior for the CERCER and CERMET Reliability of High Power proton Accelerator, Oxide Fuels Proposed as EFIT (intermediate Mol, Belgium. Report), WP3.2, DM3 AFTRA, D3.1. Barbensi, A., et al., 2007. EFIT: the European Maschek, W., et al., 2008. Accelerator driven Facility for Industrial Transmutation of Minor systems for transmutation: Fuel development, Actinides, AccApp’07, Pocatello, U.S.A. design and safety, Progress in Nuclear Energy Brunon, E., L. Donnet et al., 2004. 8 IEMPT 50, 333-340. conference, Nov 09-11, La Vegas : The Mashnik, S.G., et al., 2006. LANL report LA-UR- FUTURIX–FTA experiment in Phénix. 06-1764, http://mcnpx.lanl.gov. Burn, K.W., 1999. A Decoupled Approach to the NEA, 2006. The Jeff-3.1 Nuclear Data Library, Jeff Neutronics of Sub-Critical Configurations: Report 21, OECD, ISBN 92-64-02314-3. Evaluating the Neutron Source, RT/ERG/99/2, Rimpault, G. et al., 1997. Schema de Calcul de ENEA, Bologna, Italy. Reference du Formulaire Eranos et Orientations Burn, K.W., et al., 2003. Preliminary Nuclear pour le Schema de Calcul de Projet, CEA XT- Design Calculations for the PDS-xADS LBE- SBD-0001. Cooled Core, Proc. of the Intern. Workshop on Rimpault, G., 2006. Definition of the detailed P&T and ADS Development, Mol, Belgium. missions of both th Pb-Bi cooled XT-ADS and Chadwick, M.B., et al., 1999. Nuclear Science and Pb cooled EFIT and its gas back-up option, Engineering 131, 3, 293. CEA NT/DEN/DER/SPRC/LEDC/05-420, Croixmarie, Y., et al., 2003. Fabrication of D1.1 EFIT. transmutation fuels and targets: the ECRIX and Schikorr, W.M.., 2001. Assessments of the kinetic CAMIX-COCHIX experience. J.Nucl. Mater. and dynamic transient behaviour of sub-critical 320 (1-2), 11-17. systems (ADS) in comparison to critical reactor Donnet, L., et al., 2005. The FUTURIX- systems, NED, Vol. 210, pp. 95-123. Transmutation Experiment in PHENIX: Status USDOE Nuclear Energy Research Advisory of Fuel Fabrication, GLOBAL’2005, Tsukuba, Committee and the Generation IV International Japan, 9-13 October. Forum, 2002. A Technology Roadmap for

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