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
b
ENEA, Via Martiri di Montesole 4, IT-40129 Bologna, Italy
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 MWth, 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.
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 Ufree 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).
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 radiotoxicity 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
*
Corresponding author, carlo.artioli@bologna.enea.it
Tel: +39 051 6098436; Fax: +39 051 6098279.
1
This paper deals with the neutronic and
thermal-hydraulic design of the EFIT core (Artioli et
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.
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 –
92
Mo (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
92
Mo 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, dustfree 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 nonradioactive powders with the fissile powders. The
blends are sieved and pressed. The green pellets are
then sintered. Within the AFTRA framework,
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
2
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.
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 upright 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.
The right-bottom quarter allows to deal with the
core size, keeping the selected performance “42;0”.
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.
3
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 rightbottom-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.
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.
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
4
4.1.
goal “42;0” has been evaluated and found to be
45.7%.
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.
Table 1
MA and Pu weight compositions
MA
[w%]
Pu
237Np
238Pu
3.884
241Am
239Pu
75.510
242mAm
240Pu
0.254
243Am
241Pu
16.054
243Cm
242Pu
0.066
244Cm
244Pu
3.001
245
Cm
1.139
246Cm
0.089
247Cm
0.002
[w%]
3.737
46.446
34.121
3.845
11.850
0.001
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).
The core-layout
The chosen structural material is Ferriticmartensitic 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
Fig. 2. The 3-zones EFIT core (180 fuel assemblies).
Fig. 3. MA and Pu evolution during the fuel life.
5
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.
Proton current
4.4.
13.2 mA
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.
4.3.
The source parameters
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
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
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)
keff
0.97403  0.00023
Neutron source (S)
23.02  0.08
(neutrons/proton)
M= all fission neutrons / S
19.45  0.25
kS = M / (M+1)
0.95111  0.00059
(1  k eff ) / k eff
* 
0.52
(1  k S ) / k S
6
proton current is estimated to be at maximum 15.4
mA and the heat deposition in the target at
maximum 9 MW.
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.
5.
Table 3
Peak fuel, cladding temperatures (K) and cladding failure
times (hours). Nominal conditions.
5.1.
Thermal-hydraulic and transient analysis
Nominal conditions
Oxide
Layer
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
Avg Pin
Peak Pin
Cladding
failure
times
Avg
Peak Pin
Thicknes
Clad Fuel Clad Fuel Pin
s (m)
(hrs)
(hrs)
Core
zone
BOC
0
778 1493 803 1672 E11
E10
Inner EOC
(CZ1)
0
778 1097 812 1279 E9
E6
100
873 1399
7.0E4
200
950 1514
4.5E4
300
1031 1620
1.44
BOC
0
776 1515 792 1638 E11 7.0E10
Interme EOC
diate
(CZ2)
0
776 1115 796 1226 6.8E8 2.5E7
100
853 1331
1.9E5
200
923 1433
1.0E3
300
995 1531
9.8
BOC
0
770 1406 804 1667 E11
E10
Outer EOC
(CZ3)
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
cladding. Under EOC conditions the acceptability
depends on the actual thickness of the oxide layer.
Based on the above results, the current Pbcooled 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.
7
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. 6. Transients of the temperatures in the core.
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
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.
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.
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.
8
7.
Knebel, J., et al., 2006. EUROTRANS: European
Research Programme for the Transmutation of
High-level Nuclear Waste in an Acceleratordriven System, Ninth Information Exchange
Meeting, Nimes, France.
Kondo Sa., Morita, K., Tobita, Y., Shirakawa, N.,
1992. SIMMER-III: An Advanced Computer
Program for LMFBR Severe Accident Analysis,
ANP'92, Tokyo, Japan, Oct. 25-29, No. 40-5.
Maschek, W., et al., 2005. SIMMER-III and
SIMMER-IV Safety Code Development for
Reactors with Transmutation Capability, M&C
2005, Avignon, France, Sept. 12-15.
Maschek, W., et al., 2007. A Comparative
Assessment of Safety Parameters and Core
Behavior for the CERCER and CERMET
Oxide Fuels Proposed as EFIT (intermediate
Report), WP3.2, DM3 AFTRA, D3.1.
Maschek, W., et al., 2008. Accelerator driven
systems for transmutation: Fuel development,
design and safety, Progress in Nuclear Energy
50, 333-340.
Mashnik, S.G., et al., 2006. LANL report LA-UR06-1764, http://mcnpx.lanl.gov.
NEA, 2006. The Jeff-3.1 Nuclear Data Library, Jeff
Report 21, OECD, ISBN 92-64-02314-3.
Rimpault, G. et al., 1997. Schema de Calcul de
Reference du Formulaire Eranos et Orientations
pour le Schema de Calcul de Projet, CEA XTSBD-0001.
Rimpault, G., 2006. Definition of the detailed
missions of both th Pb-Bi cooled XT-ADS and
Pb cooled EFIT and its gas back-up option,
CEA
NT/DEN/DER/SPRC/LEDC/05-420,
D1.1 EFIT.
Schikorr, W.M.., 2001. Assessments of the kinetic
and dynamic transient behaviour of sub-critical
systems (ADS) in comparison to critical reactor
systems, NED, Vol. 210, pp. 95-123.
USDOE Nuclear Energy Research Advisory
Committee and the Generation IV International
Forum, 2002. A Technology Roadmap for
Generation IV Nuclear Energy Systems, GIF002-00.
Acknowledgement
The authors thank the partners of the IPEUROTRANS 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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