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

The sustainability and the public acceptance of nuclear energy production can be improved by the minimization and reduction of nuclear waste. The Minor Actinides (MA) have a long-term radio-toxicity and one of the possible future strategies for transmuting them is represented by the use of Accelerator Driven Systems (ADS), which allow a higher MA content in the fuel. On the other side, the cost/benefit ratio of such innovative systems has to be evaluated and challenging coordinated R&D is necessary.

Within the 6th Framework Program, the European Community has funded, besides other projects supporting partitioning and transmutation, a

conceptual design of an ADS (Domain DM1 of the Integrated Project EUROTRANS). This project is called EFIT (European Facility for Industrial Transmutation) and investigates the feasibility and the potentiality of such systems (Knebel, 2006). The design will be worked out to a level of detail which allows a cost study estimate. EFIT, of about 400 MWth, is loaded with MA and Pu in a CERCER U-free fuel. The core coolant, allowing a fast spectrum, is pure lead, as well as the windowless target for the 800 MeV proton beam. The reference sub-critical level has been postulated to be keff=0.97, figure that has to be confirmed by the full safety analysis (Rimpault, 2006).

This paper deals with the neutronic and thermal-hydraulic design of the EFIT core (Artioli et

* Corresponding author, carlo.artioli@bologna.enea.itTel: +39 051 6098436; Fax: +39 051 6098279.

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al., 2007a; Barbensi et al., 2007). The core has been conceived with the aims of: maximizing the fissions of MA, achieving a negligible keff swing during the cycle (to keep the proton current rather constant in order to avoid an oversizing of the target and of the accelerator), maximizing the average power density (i.e. the volume density of MA transmutation), while keeping low the coolant pressure drop.

For EFIT, as for any kind of reactors, the defence-in-depth concept has been applied. The demonstration of the adequacy of design with the safety objectives is structured along three kinds of basic conditions: The Design Basis Conditions (DBC–structured into 4 Categories), Design Extension Conditions (DEC–limiting events, complex sequences and severe accidents) and Residual Risk Situations. For the EFIT the safety principles and safety guidelines have been defined within EUROTRANS and a comprehensive and representative list of transients has been defined to test the safety behaviour of the reactor plant. For innovative reactors such as the EFIT ADT cliff-edge effects should be identified and excluded. For a safety classification fuel limits related to the different safety categories have been defined based on recent experimental evidence. Due to the existing uncertainties, fuel melting or disintegration should only be allowed in the DEC category. Important boundary conditions to be taken into account in the safety evaluation are the significant positive void worth, the missing of the Doppler prompt reactivity feedback, the very low delayed neutrons effective fraction (Artioli et al. 2007a) and the strong production of He via the transmutation process.

While coolant boiling processes can be excluded because of the high boiling point of lead coolant, pin failures could lead to a gas blow-down from the plena, to local voiding and reactivity addition. From the list of transients some representative ones, which are also traditionally investigated in fast reactor systems, have been chosen for the current paper, as the unprotected loss of flow (ULOF).

2. The inert matrix

In Europe a vast experience exists on oxide fuels, therefore the main emphasis of the ADT fuel development concentrated on the oxide route. In the EUROTRANS Domain AFTRA (DM3) various fuel forms as solid solution and/or composites as

CERCER and CERMET have been assessed and finally as matrices the materials ZrO2, MgO and Mo had been under closer investigation. The final recommendation on fuels gave a ranking of these fuels based on a number of criteria, ranging from fabrication, reprocessing via economics to safety. The composite CERMET fuel (Pu0.5,Am0.5)O2-x – 92Mo (93% enriched) has been recommended by AFTRA as the primary candidate for the EFIT (Maschek, 2008). This CERMET fuel fulfils adopted criteria for fabrication and reprocessing, and provides excellent safety margins. Disadvantages include the cost for enrichment of 92Mo and a lower specific transmutation rate of minor actinides, because of the higher neutron absorption cross-section. The composite CERCER fuel (Pu0.4,Am0.6)O2-x – MgO has therefore been recommended as a backup solution as it might offer a higher consumption rate of minor actinides, and can be manufactured for a lower unit cost. In the EFIT development the demonstration of an efficient transmutation performance is a key issue. Therefore the DM1 design concentrated on the CERCER core first, the more as preliminary analyses showed the compliance with normal operation and safety criteria.

The fabrication of composite pellets is considerably more difficult than solid solution oxide pellets. This is a result of the specific requirements of size and homogeneous distribution of the dispersed actinide phase. The fuel development for AFTRA is performed at CEA and at the Institute for Transuranium Elements (ITU). In the framework of the EFIT design the fissile phase volumetric content of these fuels is around 50%. The samples made for in-pile tests hold less than 30-40% of fissile particles because of nuclear facility constraints on authorised Minor Actinide contents. The fabrication route used at the laboratory scale for these highly radioactive materials firstly deals with fissile particles preparation by clean and necessarily, dust-free fabrication methods to minimize contamination in the gloveboxes. Two processes are used: an oxalic co-precipitation route (Brunon, 2004; Croixmarie, 2003) for CEA, and a combination of external gelation (Fernandez, 2006) and infiltration methods (Fernandez, 1999) for ITU. The following steps belong to the conventional powder metallurgy area: they consist in mixing and grinding the non-radioactive powders with the fissile powders. The blends are sieved and pressed. The green pellets are then sintered. Within the AFTRA framework,

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CERMET and CERCER pellets dedicated to irradiation tests, have been fabricated using the both procedures. Such produced pellets are irradiated within the framework of FUTURIX-FTA tests in the Phenix fast reactor (Donnet, 2005). The FUTURIX tests are of central importance for the development of these dedicated fuels.

3. Conceptual guidelines and rationales

Dealing with ADS, as with any complex system, a number of parameters either directly or indirectly interlinked ought to be kept simultaneously under control. Very often an “eel effect” occurs: paying attention and acting for optimization of some parameters other, not less important, are moved away and vice-versa. To help for getting a simultaneous vision at glance of the system, the A-BAQUS graph (fig. 1, reported numbers are those typical of the EFIT-Pb system) has been proposed (Artioli, 2007b). In the graph some key-parameters (namely burning efficiency, fuel enrichment, reactivity swing, active fuel volume, power and core size, accelerator proton current and its range along the cycle) are shown as well as their logical relationships by the mean of typical curves, each marked by the referred enrichment E (Pu/(Pu+MA)).

No matter the performances claimed about the MA burning efficiency, it has to be admitted that the fission rate is in any case 42 (rounded number) kg/TWhth, that is merely the 200 MeV/fission in changed units of measure, in any nuclear system (thermal, fast, low, high flux; soft, hard spectrum; small, huge size, with any coolant, etc.).

Fig. 1. A-BAQUS graph.

42 kg/TWhth is not a result of a design, but a physical constant. What can make the difference is either (the measure unit, i.e. kg/TWhth, is here omitted):

- the 42 fissioned can be differently split between MA and other heavy nuclides (Pu or U) or

- along the fissions (the universal 42), events on MA other than fission can occur; so the MA “disappearing” can be actually higher than 42, that in turn would simply mean the exceeding part has been transmuted in other heavy nuclides (i.e. Pu).

We can condensate all that in a pair of numbers: the first one indicates the overall MA disappearance (either fissioned or transmuted), the second one the new Pu production. Their difference must be in any case 42 (fig. 1, right double-marked axis of the up-right quarter). For instance “65;23” means that 65 MA disappear and 23 new Pu is produced, i.e. 42 out of 65 MA really fission and the remainder 23 transmute into new Pu. In this case EFIT acts as a converter from MA to Pu (red zone). The performance for MA depends on the Pu policy rather than on the MA one! It is easy to recognize how the value of the pair is directly ruled by the ratio between Pu and MA, i.e. by the enrichment. Yet this parameter also rules directly the reactivity swing in the cycle (left axis in the top-right quarter), that in turn drives the range of the accelerator current (bottom left quarter). With the MA and Pu vectors assumed in the EFIT design (Rimpault, 2006), the above mentioned case would mean: enrichment 27% with a Kswing about 0.019 (one year cycle).

In an ADS the unit of energy, one fission for 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.

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The right-bottom quarter allows to deal with the core size, keeping the selected performance “42;0”. Moving on the referred curve E=45.7% a core power size can be selected acting on the active fuel fraction, marked on the abscissa as the complement content of the inert matrix. Of course these right-bottom-quarter relationships are driven by the thermal-hydraulic setting of the core, namely the linear power rating, the enthalpy equation, the coolant velocity.

In the left-bottom quarter, current and its swing are reported for the selected enrichment, and therefore performance, according to the core size.

The burning capability is expressed in terms of kg/TWhth or/and in terms of “percentage of the inventored MA/year”. In the EFIT this rate is 4.5%/year. For a coherent comparison it has to be kept in mind that in EFIT the 4.5% are actually fissioned (and not partially transmuted in other heavy isotopes). Since this rate depends only on the MA cross sections and flux intensity, the only way to claim a better figure is to have a higher flux (and/or a more effective spectrum).

The rate of percentage has directly an economic implication: the shorter is the time the cheaper is the process. But as far as the efficiency is concerned, what is important is not the percentage/year (velocity of burning), but the percentage at the discharge. This last is ruled directly from the max allowed BU: if the flux is higher this maximum is reached earlier, but it does not change.

The rate 4.5%/year is ruled, via flux intensity, by the “external” constraints, as the available target cooling system (11 MWth) and the required subcriticality (here postulated to be 0.97).

In the EFIT a prudential max BU of some 100 MWd/kg (HM) has been assumed in first step. The final percentage at the discharge is then a satisfactory 13.9%. This figure means that, at every unit of MA fissioned, reprocessing losses of 7 units have to be associated.

Of course a complete characterization of the new fuel, either with the MgO or Mo inert matrix, could allow higher figure of BU and consequently higher figure of the percentage of fissioned MA at the discharge.

4. The EFIT equilibrium core

The actual “perfect MA burner” is the reactor where only new MA are used for refueling and only

fission products are unloaded. Preliminary analyses show that this is a possible scenario with EFIT. Of course for that purpose an equilibrium composition has to be reached, in which the equilibrium vector of the plutonium is quite different from the beginning one (i.e. richer in even isotopes and poorer in odd ones). Nevertheless, an equilibrium enrichment exists (about 60-70%) and, more important thing, such a mixture ought to have enough reactivity to sustain an EFIT core.

This paper deals with the EFIT start-up core. The first step in designing the core has been the definition of the unique enrichment that fits the “42;0” approach. Keeping constant this pair, a suitable optimization of the core can be pursued arranging the volumetric fractions and the geometry in order to reach the desired keff (0.97) (Barbensi et al. 2007) and to flat the radial distribution, both for economy and for respecting the technological constraints, mainly Tclad max 823 K, Tfuel max 1650 K (500 K below the disintegration temperature of the inert matrix; Maschek et al., 2008).

It is important to note that, being the Pu content rather constant in the cycle, the reactivity swing will not be large. This allows to keep a rather constant proton current, avoiding an oversizing of both the accelerator and the target module.

In the operating conditions, the mean outlet temperature of the coolant (pure lead) of 753 K is rather close to the maximum allowed temperature of the cladding of 823 K (USDOE, 2002). Therefore, the spread of the outlet temperatures of the subassemblies, belonging to the same zone of flow rate, must have a low peak factor (lower than 1.2 in first approximation). To meet this requirement the core is radially subdivided in three zones of flow rates, ruled by suitable orificing.

In order to flat the radial flux profile, the active fuel volume fraction is increased along the radius.

Since the “42;0” approach defines univocally the enrichment, to flatten the radial flux profile the active fuel VF has been increased along the radius. In detail:- from the inner zone to the intermediate one, the

fuel/matrix ratio has been changed from 43% up to 50%, by keeping the same pin diameter and and pitch;

- from the intermediate zone to the outer one (where the flux and the power density become quite lower anyway and less cooling is required) the pin diameter has been increased by keeping the same pitch and fuel/matrix ratio.

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4.1. Calculation tools for neutronic calculations

The core has been designed mainly by means of the deterministic code ERANOS (Rimpault, 1997), with both a 2D cylindrical and a 3D hexagonal schematization. The Monte Carlo code MCNPX (Hendricks, 2006) has also been used because it allows to transport particles at high energy. Moreover it can calculate a detailed power distribution with a heterogeneous description of the fuel assemblies. The whole system has been modelled for MCNP in a detailed 3D geometry (including thermal expansions and neutron libraries at different temperatures). The methodology followed (Burn, 1999) was thus to use MCNPX to calculate, starting from the 800 MeV proton beam, the neutron source for ERANOS. The neutron source is defined as the first neutrons appearing in the system with energy below 20 MeV. The spatial and energy distributions of these neutrons are used as input for ERANOS.

The neutron libraries used for the codes are: ERALIB1 (Jef2.2) for ERANOS; a combination of 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 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

spent UO2 but with the storage period of 15 years. With these vectors the enrichment fitting the pair goal “42;0” has been evaluated and found to be 45.7%.

Table 1MA and Pu weight compositions

To respect the maximum fuel temperature allowed, a limiting linear power rating has been evaluated. Since the pellet thermal conductivity depends on the inert matrix content, a linear power rating of 180 W/cm has been found for the pellet with 50% of matrix (minimum content, for the intermediate and outer zones) and a rating of 200 W/cm for the pellet with 57% of matrix (for the inner zone).

Fig. 2. The 3-zones EFIT core (180 fuel assemblies).

MA [w%] Pu [w%]237Np 3.884 238Pu 3.737241Am 75.510 239Pu 46.446

242mAm 0.254 240Pu 34.121243Am 16.054 241Pu 3.845243Cm 0.066 242Pu 11.850244Cm 3.001 244Pu 0.001245Cm 1.139246Cm 0.089247Cm 0.002

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Fig. 3. MA and Pu evolution during the fuel life.The dimension and the composition of the pin

and of the fuel assembly is reported in (Artioli, 2007a). While the pin diameter and the pitch derive from the thermal balance, the fuel assembly dimensions are driven by the size of the spallation module, which has to be inserted replacing the 19 central assemblies. The core is shown in fig. 2.

The residence time is stated in 3 years, life time that allows to reach the peak burn up of about 10% (Knebel, 2006), within the limit imposed by the corrosion and well below the dpa limit. This span of time is divided into three subcycles 1 year long. Due to the rather constant content in Pu during the irradiation the reactivity swing is very small, 200 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.

4.3. The source parameters

The main integral parameters (MCNPX results) at BOC are reported in Table 2. There is a discrepancy of 930 pcm in keff between ERANOS (with the ERALIB1-Jef 2.2 library) and MCNPX (with the Jeff 3.1 library). The MCNPX results using the ENDF/B-VI library for the fuel appear to be more similar to ERANOS (320 pcm of difference in keff). Note that the neutron source efficiency is *=0.52, while in the PDS-XADS design (Burn, 2003) was *=0.99 (kS and keff very similar). This effect is mainly due to the different fuel composition and to the larger radius of the target.

Table 2 MCNPX results at BOC (the error is the stand. deviation)

keff0.97403

0.00023Neutron source (S)(neutrons/proton) 23.02 0.08

M= all fission neutrons / S 19.45 0.25

kS = M / (M+1) 0.95111 0.00059

0.52

Proton current 13.2 mA

4.4. The power distribution

The flux radial flattening aims to reduce as much as possible the power radial form factor within each radial zone, and to reach the maximum power density peaks allowed (corresponding to 200 W/cm and 180 W/cm according to the different matrix content). Figure 4 shows the power density radial profile, on the peak plane (about midplane), obtained in a 2D RZ geometry. This flattening has been further improved by the 3D XYZ model (Artioli, 2007a).

Fig. 4. Radial profile of the homogeneous power density.

The overall power (beam excluded) of the core is 389 MWth, 5 MWth of which are dissipated in structural zones outside the active core. The obtained average homogeneous power density is 70.7 W/cm3.

The power deposition distribution has been calculated by means of both ERANOS and MCNPX. The results used as reference for the thermal hydraulic analysis are those from MCNPX: the power has been calculated in each assembly of the core, separated per ring. From these values, the 3 hottest assemblies in the 3 zones have been identified, together with the axial form factors and the value of the heat release in the hottest pin (16.4 kW). The maximum linear power in the fuel pins turns out to be 203 W/cm. The differences with ERANOS are within 5% for the hottest assemblies and within 3% for the axial form factors. As a result

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of this analysis, better zone contours can be defined, mainly for the Intermediate/Outer interface.

As far as the power in the target is concerned, MCNPX calculations show that 73% of the beam power is deposited in the target circuit. If, during the life of the system keff is always around 0.97, then the proton current is estimated to be at maximum 15.4 mA and the heat deposition in the target at maximum 9 MW.

5. Thermal-hydraulic and transient analysis

5.1. Nominal conditions

The thermal-hydraulic analyses of the core were performed with the static version of the SIM-ADS code (Schikorr, 2001) for each of the 3 core zones. Two core conditions are analyzed, namely Beginning-Of-Cycle (BOC) and End-Of-Cycle (EOC).

The thermal conductivity of the two different MA-fuel compositions, namely MgO volume fractions of (CZ1/CZ2/CZ3 = 57%,50%,50%), were calculated based on the known thermal conductivities of MgO and MOX-MA-fuels using the Bruggeman weighting scheme and applying an appropriate correction for burnup. More details of this procedure can be found in (Maschek, 2007).

Under BOC conditions, fresh fuel conditions are presumed. Under nominal conditions, the size of the gap between clad and fuel has closed down to about 110 µm for the average pin, or about 70% of the cold condition value, and the gas composition in the gap is dominated by He, namely (He/Xe/Kr = 0.976/0.023/0.001).

Under EOC conditions, a peak fuel burn-up of about 100 MWd/kg has been assumed for these calculations. The gap between clad and fuel is presumed to be essentially closed (min gap ~ 4 µm) and the fission gas composition in the gap is still dominated by He due to the higher helium fission gas production in MA fuel compared to conventional fuel (factor ~3.6 has been calculated), namely (He/Xe/Kr = 0.781/0.201/0.017). For the peak pins, pin pressures of (CZ1/CZ2/CZ3= 112/116/127) bars are calculated.

An additional parameter requiring closer attention in the thermal hydraulic analysis is the formation of an oxide layer on the cladding material. The formation of this oxide layer serves a protective

function against clad corrosion, on the one side; on the other side it will impede heat transfer from the clad surface to the coolant. A maximum layer thickness of 5-10% of the cladding thickness can be presumed as a guiding parameter for EOL analysis. Several oxide layer thicknesses, namely 100, 200, and 300 µm, have been used as a parameter in our analysis. The thermal conductivity of the oxide layer 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.

Table 3 Peak fuel, cladding temperatures (K) and cladding failure times (hours). Nominal conditions.

OxideLayer Avg Pin Peak Pin

Cladding failure times

Core zone

Thickness (m) Clad Fuel Clad Fuel

Avg Pin

(hrs)

Peak Pin

(hrs)

Inner (CZ1)

BOC 0 778 1493 803 1672 E11 E10

EOC 0 778 1097 812 1279 E9 E6

100 873 1399 7.0E4

200 950 1514 4.5E4

300 1031 1620 1.44

Intermediate

(CZ2)

BOC 0 776 1515 792 1638 E11 7.0E10

EOC 0 776 1115 796 1226 6.8E8 2.5E7

100 853 1331 1.9E5

200 923 1433 1.0E3

300 995 1531 9.8

Outer (CZ3)

BOC 0 770 1406 804 1667 E11 E10

EOC 0 770 1059 799 1206 6.9E8 5.2E6

100 844 1298 1.1E5

200 904 1390 1.2E3

300 968 1479 17.4

Table 3 summarizes the results of the calculations performed at nominal operations (100% load). Under nominal BOC conditions, peak fuel and peak clad temperatures are well within acceptable upper limits for all 3 core zones, namely ~ 1650 K for the fuel and about 823 K for the

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cladding. Under EOC conditions the acceptability depends on the actual thickness of the oxide layer.

Based on the above results, the current Pb-cooled EFIT design seems quite viable. Attention needs to be placed however on the operational control of the oxygen content in the Pb coolant in order to control chemical fowling and the buildup of the oxide layer.

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 used in the ULOF simulation.

Fig. 5. Transients of the pump head and the coolant mass flow rate.

With the above assumptions, the coolant mass flow rate will follow a transient process as shown in Fig. 6. It firstly decreases to about 20% of its initial value when the pump head arrives at zero and finally keeps a value of about 32% with some slight oscillation. Fig. 6 shows that, with the 32% remained coolant heat removing capacity, the fuel, clad, and coolant peak temperatures finally stabilized at around 1835 K, 1000 K, and 955 K, respectively. It also shows that during the ULOF transient, the highest temperatures that fuel, clad and coolant will experience are about 1860 K, 1030 K,

and 985 K, respectively. The clad and coolant temperatures are well below the failure limit and also the fuel peak temperature is well below the limits for melting and disintegration (2150 K) given by the ‘Fuel-Domain’ of EUROTRANS (Maschek et al., 2008). Fig. 7 shows that in the ULOF condition the increase of the reactivity and consequently the power in the core is low.

Fig. 6. Transients of the temperatures in the core.

Fig. 7. Transients of the power and reactivity.

6. Conclusions

The MA fission (120 kg/year) via an U-free lead cooled ADS as EFIT is proved to be viable. The “42;0” approach assures that every fission is devoted to an atom belonging to MA, while the Pu content is kept constant, acting as a catalyzer.

Normal condition and transient (ULOF) analyses show the respect of the technological limits, even if efforts have to be devoted for lowering the total pressure drop as well as for a better power distribution flattening.

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The current MA burning rate of 13.9% of the initially inventoried at the discharge, is strictly ruled by the max BU allowed. This figure in turn rules directly the total reprocessing losses. Therefore R&D effort has to be devoted to the qualification of the fuel, to the PCMI as well as to the qualification of the steel and its treatment in lead environment. The use of Mo-92 as more promising inert matrix has to be investigated.

7. Acknowledgement

The authors thank the partners of the IP-EUROTRANS project for their fruitful contribution to the project. Special thanks to the European Commission for the financial support through the FP5 and FP6 programs.

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