Annals of Nuclear Energy 29 (2002) 1871–1889
www.elsevier.com/locate/anucene
Neutronic analysis of PROMETHEUS reactor
fueled with various compounds of thorium and
uranium
Hüseyin Yapıcıa,*, Mustafa Übeylib, Şenay Yalçınc
a
Erciyes Üniversitesi, Mühendislik Fakültesi, 38039 Kayseri, Turkey
b
Gazi Üniversitesi, Teknik Eǧitim Fakültesi, Ankara,
Turkey
.
c
Bahçeşehir Üniversitesi, Fen-Edebiyat Fakültesi, Istanbul, Turkey
Received 28 December 2001; accepted 18 January 2002
Abstract
In this study, neutronic performance of the DT driven blanket in the PROMETHEUS-H
(heavy ion) fueled with different fuels, namely, ThO2, ThC, UO2, UC, U3Si2 and UN is
investigated. Helium is used as coolant, and SiC is used as cladding material to prevent fission
products from contaminating coolant and direct contact fuel with coolant in the blanket.
Calculations of neutronic data per DT fusion neutron are performed by using SCALE 4.3
Code. M (energy multiplication factor) changes from 1.480 to 2.097 depending on the fuel
types in the blanket under resonance-effect. M reaches the highest value in the blanket fueled
with UN. Therefore, the investigated reactor can produce substantial electricity in situ. UN
has the highest value of 239Pu breeding capability among the uranium fuels whereas UO2 has
the lowest one. 239Pu production ratio changes from 0.119 to 0.169 according to the uranium
fuel types, and 233U production values are 0.125 and 0.140 in the blanket fueled with ThO2
and ThC under resonance-effect, respectively. Heat production per MW (D,T) fusion neutron
load varies from 1.30 to 7.89 W/cm3 in the first row of fissile fuel breeding zone depending on
the fuel types. Heat production attains the maximum value in the blanket fueled with UN.
Values of TBR (tritium breeding ratio) being one of the most important parameters in a
fusion reactor are greater than 1.05 for all type of fuels so that tritium self-sufficiency is
maintained for DT fusion driver. Values of peak-to-average fission power density ratio, , are
in the range of 1.390 and 1.476 depending on the fuel types in the blanket. Values of neutron leakage out of the blanket for all fuels are quite low due to SiC reflector. The maximum
neutron leakage is only 0.025. Consequently, for all cases, the investigated reactor has high
neutronic performance and can produce substantial electricity in situ, fissile fuel and tritium
required for (D,T) fusion reaction. # 2002 Elsevier Science Ltd. All rights reserved.
* Corresponding author. Tel.: +90-352-437-49-01; fax: +90-352-437-57-84.
E-mail address: [email protected] (Hüseyin Yapıcı).
0306-4549/02/$ - see front matter # 2002 Elsevier Science Ltd. All rights reserved.
PII: S0306-4549(02)00016-6
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1. Introduction
At present, the most widely used conventional nuclear reactors in the world are
light water reactors (LWR) and the Canada deuterium uranium reactor (CANDU)
that convert nuclear energy to electric energy. LWR uses enriched uranium (3–4%
235
U) as fuel while CANDU uses natural uranium which is the main source for fissile
fuel consists of 0.7% 235U available as the main fissile fuel presently. For the great
number of LWRs, fast breeder reactors (FBR) are not sufficient to supply nuclear
fuel because of the very long doubling time (10–30 years) of a FBR. Some 239Pu can
be produced in CANDU and LWR with a very low conversion ratio (0.6–0.7).
Therefore, LWRs can expand only 1% of the natural uranium, and still, 99% of
the fissionable natural resources are not being used for energy production. In addition, thorium reserves are estimated to be about three times more abundant than the
natural uranium reserves in the world so that it has an important potential for use in
nuclear reactors.
1.1. Concept of fusion energy breeder
Two different systems have been studied to produce nuclear fuel required for
production of energy in the conventional nuclear reactors and transmutation of
nuclear wastes. There are two main rich neutron sources as fissile fuel producers.
These are:
(1) Fusion reactors based on (D,T) or (D,D) reactions,
(2) Electro-nuclear breeders based on principle of spallation of heavy nucleus
under bombardment of relatively high energetic protons ( 1 GeV).
The idea of production of abundant fissile fuel by using fusion breeders or electronuclear breeders is quite old (Berwald et al., 1982; Greenspan, 1984; Lee et al., 1982;
Moir, 1979; Mynat, 1977; Raghep et al., 1979; Teller, 1981). The studies indicate
that a fusion breeder can produce up to 30 times more fissile fuel than in a FB (fast
breeder) per unit of energy (Teller, 1981). A hybrid reactor is a combination of
fusion and fission processes that surrounds the fusion plasma with a blanket containing fertile materials (232Th and/or 238U) and lithium for producing fissile fuel
and tritium, respectively. In addition, the fertile materials may also undergo a substantial amount of fission and waste actinides can be burnt effectively due to highly
energetic fusion neutrons.
Therefore, fusion energy breeder reactors are alternative by using thorium
reserves, natural uranium and spent fuel containing a substantial amount of 238U for
producing fissile fuel and electricity in situ, and burn all nuclear waste actinide’s,
such as, 237Np, 238Pu, 240Pu, 241Am, 243Am and 244Cm. However, fissile fuel produced in fusion breeders can become highly prolific and must be denatured. Hence,
considerations for the denaturation of these new nuclear fuel sources become very
important (Şahin et al., 1998, 2001a,b; Şahin and Yapıcı, 1999).
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There are many research and development studies that have been done on ICF
(inertial confinement fusion) power reactors. Most of the studies have been focused
on glass lasers. The studies about laser-driven ICF power reactors with blankets
containing fertile or fissile fuel have also been an important research field due to the
fact that these systems could reduce the power requirements for the laser beam,
increase the net driver efficiency, and provide the high safety of a hybrid fusion-fission reactor. The blanket concept should be adaptable to most nuclear applications,
including the use of natural uranium, LWR or CANDU spent fuel or thorium as
fertile materials. The capability of the production of fissile fuel in a fusion energy
breeder reactor has been studied in earlier works (Berwald et al., 1982,1987; Erickson et al., 1981; Greenspan, 1984; Greenspan et al., 1983; Jassby et al., 1986; Kumar
and Şahin, 1983; Lee et al., 1982; Maniscalco et al., 1984; Moir, 1979, 1982, 1985;
Moir et al., 1984a,b, 1985a,b; Moir and Lee, 1986; Mynat, 1977; Raghep et al., 1979;
Şahin, 1989; Şahin and Al-Kusayer, 1985, 1986; Şahin and Yapıcı, 1989a,b, 1998,
1999; Şahin et al., 1985, 1986, 1991a,b, 1994, 1998, 1999, 2001a,b; Yapıcı et al.,
1999a,b, 2000a,b, 2001; Yapıcı and Özceyhan, in press).
Inertial fusion energy (IFE) is a potentially attractive approach to fusion energy.
IFE has an affordable development path resulting from:
the separability and modularity of components and systems,
the breadth of possible attractive options in targets, drivers, and chambers,
the leverage from other applications of ICF technology such as those in
national security, high-energy physics, and industry,
the fact that engineering development of power plant chambers can be done
in scaled, lower-cost reaction chambers (Campbell et al., 1999).
1.2. Overview of PROMETHEUS reactor
The PROMETHEUS-L (laser) and PROMETHEUS-H (heavy Ion) IFE power
plant designs were completed by DOE in USA in early 1992. The PROMETHEUSL design would use a direct-drive target, driven by a KrF laser, whereas the PROMETHEUS-H design would use an indirect-drive target, driven by 4 GeV lead ions.
The two designs are nearly identical from the consideration of target chambers. The
major differences between the two designs include the need for high-Z material in
the indirect-drive target and the dimensions of the target chamber. The heavy-iondriven chamber would be only about nine-tenths the size of the laser-driven design.
Therefore the smaller size in the PROMETHEUS-H means that less material would
be required for the first wall and blanket system, it also would result in a higher
neutron flux at the first wall. However, the higher first wall neutron loading and the
need for high-Z material suggest that the hazards of the heavy-ion-driven design
may be greater than those of the laser-driven design. The PROMETHEUS-H design
would use a single-beam linear accelerator to accelerate Pb+2 ions to energies of 4
GeV. A total of 7.0 MJ would be delivered to the target, and 719 MJ of fusion
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energy would be released for a target gain of 103. The repetition rate would be 3.54
Hz. Since the heavy-ion design would deliver 999 MWe to the grid, the repetition
rate does not need to be adjusted for comparison with the other designs (Waganer et
al., 1992).
The PROMETHEUS-L/-H IFE design was strongly based on considerations of
safety, reliability, simplicity, and flexibility. Safety considerations led to the choice of
a helium-cooled, solid breeder blanket with low activation material-SiC structure
and neutron reflector and Li2O breeder. The cavity consists of the components
directly surrounding the exploding targets, including the first wall system, blanket,
coolant manifolding, vacuum vessel, and shield. These components contain the
energy of the blast, absorb the neutrons produced, convert energy into usable heat,
breed tritium sustain the DT fuel cycle, and shield components and personnel from
the high radiation environment. Thus, it has an important role in determining the
major attributes of the reactor, such as cost, safety and environmental features,
engineering attractiveness, and technical feasibility. Several fundamental principles
were established at the beginning of the PROMETHEUS cavity design process, and
guided the major design decisions throughout the study. A top priority was the
desire for inherent safety and minimum activation. This desire influenced the material choices for the first wall, blanket, and shield. The first wall employs low-activation SiC composite. Both long-term and short-term activation is small, thus
minimizing waste disposal problems and providing negligible decay heat (Tillack et
al., 1993). Low activation property is a key parameter in developing fusion reactor
materials and important for reducing the risk related to accidents in a fusion reactor.
In general, the three low activation material systems can be listed as increasingly
favorable in the following order: SiCf/SiC composites—vanadium alloy—ferritic/
martensitic steels (Dyomina et al., 1998; Jones et al., 1999; Ehrlich et al., 1998). SiC
fiber-reinforced SiC matrix composites (SiCf/SiC) are being considered as a candidate structural material for fusion reactors because of their low-induced reactivity
by 14-MeV neutron irradiation, their excellent high-temperature strength, and corrosion resistance. The high thermal efficiency is one of the advantages of an SiCf/SiC
material system (Aiello et al., 2000; Hasegawa et al., 2000; Jones and Henager, 1995;
Kohyama et al., 2000; Riccardi et al., 2000; Seki et al., 1998).
The blanket also uses SiC structure and reflector, with low-activation Li2O breeder and He coolant. The tritium inventory in the breeder was minimized. Use of He
at relatively low pressure, together with multiple containment barriers, makes blanket failures unlikely. To reduce activation, the PROMETHEUS uses an innovative,
highly-effective shield consisting of Al structure, water coolant, and B4C, Pb, and
SiC absorbers as the shield material instead of concrete (Perlado and Sanz, 1998;
Tillack et al., 1993).
Wetted walls have many potential engineering advantages, including good beamline accommodation, relaxed repetition rate limitations (as compared with thick
films), flexible engineering features, and low inventory and flow rate of the liquid
film according to the studies done. That is why in the PROMETHEUS design, a
wetted-wall design was adopted with separate first wall and blanket. It was concluded that the wetted-wall concept could be adapted to both driver designs. The
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overall configuration of PROMETHEUS is a low aspect ratio cylinder with hemispherical end caps. This configuration was chosen for several reasons.
(1) Maintenance of a cylinder is easier than a sphere and maintenance paths are all
straight vertical lines and the configuration allows independent removal of FW
panels and blanket modules. (2) A cylinder provides better control of film flow.
Problems protecting the upper hemisphere can be reduced with higher aspect ratio,
in which the distance from the blast to the upper end cap can be maximized. (3) A
cylindrical configuration is more consistent with conventional plant layouts.
The main disadvantage of this concept is nonuniform power distribution and
higher peak loads. The higher peak-to-average loading leads to larger size and
higher cost for a given total reactor power (Tillack et al., 1993).
2. Blanket geometry
2.1. First wall of the PROMETHEUS-H reactor
The PROMETHEUS IFE uses a thin liquid Pb film supplied from Pb coolant
tubes through a porous structure of SiC composite material to protect the first wall.
A film thickness nominally is 0.5 mm that is allowed to form on the surface facing
the pellet explosions. The porosity of the bulk SiC is tailored to allow Pb from the
cooling channels to slowly seep onto the surface. Wetting of SiC by Pb is assisted by
coating the SiC with a metal as part of the chemical vapor deposition (CVD) process. Protection of the upper end cap is a particularly difficult problem. In order to
maintain a fully-wetted surface without Pb falling off into the cavity, a thin Pb jet is
injected at the top of the upper end cap. The jet injector structure (SiC) is exposed to
the blast effects and is made easily accessible for repair or replacement, as required.
The fluid leaves the injector and flows along the surface. The surface film evaporates
in response to the intense heat flux from the target explosions and then recondenses
prior to the next shot. The first wall coolant must have acceptable neutronic properties (either breed, multiply neutrons, or be transparent). Therefore, the choices are
limited to Li-bearing materials and neutron multipliers. Pb was selected due to its
good neutron multiplication, and chemical compatability with SiC. And also, its
thermophysical properties provide good operating temperature ranges and its relatively high saturation temperature leads to good conduction heat transfer into the
coolant. Moreover, its boiling point is not too high for materials temperature limits
and compatibility, and the relatively high bulk coolant temperature gives good
thermal conversion efficiency. Bi and BiPb were considered as alternate multipliers,
but they have much higher radioactivity. The main disadvantages of Pb are its high
density and activation (Tillack et al., 1993).
The first wall system consists of a series of panels 2-m wide which are lowered into
the cavity vertically and locked into a support system attached to the blanket. The
ability to provide removable panels which lock into the blanket is essential to allow
more frequent maintenance of the first wall panels and still decrease the mechanical
effects of the blast by absorbing the loads into the blanket and support system. The
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cooling channels are 5-cm thick to provide neutron multiplication needed for tritium
breeding. The first wall is kept thin (5 mm) to provide good heat conduction into the
coolant. The film thickness is also minimized for good heat transfer as well as to
reduce the problem of liquid entering the cavity. The SiC structure must be flexible
enough to withstand cyclic loading from the blast, but strong enough to support
itself and the internal pressure of the film. A supply region behind the porous
structure serves to slowly feed the liquid and also to remove the heat from the first
wall (40% of the total fusion power). Blast energy is removed from the cavity initially by evaporation. During the recondensation phase of each pulse, heat is conducted through the relatively thin film and into the first wall coolant (see Fig. 1). In
the PROMETHEUS design, cavity clearing requires good conduction heat transfer
into the cooling channels (Tillack et al., 1993).
2.2. Blanket of the PROMETHEUS-H reactor
The blanket of which material compositions and dimensions are given in Table 1
contains several rings through the cylindrical and hemispherical sections. Blanket
modules are pre-assembled into the rings, which stack vertically on top of one
another. SiC is used to make blanket modules and contains a number of U-bend
woven SiC tube sheets inside which the pressurized He coolant flows. The Li2O that
is placed in packed bed form between the tube sheets and is purged by He flowing
along the axis of the module, in conjunction with the first wall Pb coolant supplies
the potential for adequate tritium breeding without the need for Be as a multiplier.
Li2O having high tritium breeding potential is an attractive candidate for the solid
breeder material of a D–T fusion power plant, because of its high lithium atom
Fig. 1. Schematic of a blanket module.
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density (compared to other lithium ceramics or metallic lithium) and its relatively
high thermal conductivity. A schematic of a module which is shown in Fig. 1 consists of a simple layered configuration (Waganer et al., 1992).
SiC has been selected as the structural material in the first wall and blanket/
reflector/plena regions due to the high neutron reflectivity property of carbon. SiC
has lower absorption cross-section for neutrons than stainless steels. A lower
Table 1
Material composition and dimension of the zones of the blanket
Zone
Material
Cavity
Pb Film
Vacuum
Pb
SiC
Pb
SiC
Pb
SiC
Vacuum
SiC
He
Li2O
SiC
He
Fuelb
SiC
He
Li2O
SiC
He
SiC
He
SiC
He
SiC
First Wall
Gap
Blanket Wall
Tritium
Breeding
Fissile Fuel
Breeding
Tritium
Breeding
Reflector
Plena
Blanket Wall
a
b
c
Dimension (cm)
Fraction (%)
450.00
0.05
0.50
5.00
0.50
3.95
2.50
15.00
12.00
33.00
20.00
17.50
4.00
100
90
10
10
90
100
84
16
44a
22
34
44c
22
34
44a
22
34
90
10
10
90
100
Material
Density (g/cm3)
He
Li2O
SiC
Pb
ThO2
ThC
UO2
UC
U3Si2
UN
0.00715
2.013
3.2
11.34
9.86
10.65
10.55
13.63
12.2
14.31
80% TD.
ThO2, ThC, UO2, UC, U3Si2, UN.
76% Fuel, 3% clad (SiC) and 21% coolant (He).
Fig. 2. Zones and dimensions in one-dimensional cavity model (FW=first wall, BW=blanket wall,
TB=tritium breeding, FFB=fissile fuel breeding), (dimensions are given in cm).
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absorption rate in the structural material causes to a higher absorption rate in the
breeding material so that, use of SiC is advantageous as far as tritium breeding is
concerned.
2.3. Fissile fuel breeding zone
In this study, the tritium breeding zone of the PROMETHEUS-H IFE reactor
that has thickness of 50 cm, is divided into three parts as 15, 12 and 27 cm. Then, in
order to breed fissile fuel from fertile fuel, FFB (fissile fuel breeding) zone containing
fuel spheres filled with compounds of thorium or uranium and cladded with SiC, is
located instead of the second part of the tritium breeding zone having a thickness of
12 cm (see Fig. 2). Inner and outer diameters of the fuel spheres are 1.1 and 1.12 cm,
respectively, and they are arranged as hexagonal in ten rows having pitch length=1.2
cm in radial direction. Different type of fuels, ThO2, ThC, UO2, UC, U3Si2 and UN,
are considered in the FFB to investigate the neutronic performance of the blanket.
Helium is used as a coolant in the FFB to supply nuclear heat transfer.
3. Numerical results
3.1. Calculation procedure
The FFB zone is divided into ten equal sub zones, and each sub zone is again
divided into three intervals to get more precise results. The neutronic calculations
have been performed by solving the Boltzmann transport equation with the help of
neutron transport code XSDRNPM/SCALE4.3 (Anon., 1995), and 238 energy
groups neutron transport and activity cross-section data library (Greene et al., 1994).
This library is also known as the Library to Analyze Radioactive Waste (LAW) Library.
The library has 148 fast groups and 90 thermal groups (below 3 eV), and all nuclides in
238-group LAW Library use the same weighting spectrum, consisting of Maxwellian
spectrum (peak at 300 K) from 104 to 0.125 eV, a 1/E spectrum from 0.125 eV to
67.4 keV, a fission spectrum (effective temperature at 1.273 MeV) from 67.4 keV to
10 MeV, and a 1/E spectrum from 10 to 20 MeV. The integration of the angular
neutron flux has been applied in S8-P3 approximation. We have used SCALE4.3
module, CSAS1X and codes, namely, NITAWL-II, BONAMI and XSDRNPM.
CSAS1X is a control sequence that creates the binary input data files for use by the
codes BONAMI, NITAWL, and, optionally, XSDRNPM and/or ICE to provide a
problem dependent cross-section library. BONAMI performs resonance self-shielding calculations in the unresolved resonance range (when Bondarenko type parameters are available) while NITAWL-II performs resonance self-shielding
calculations in the resolved resonance range. XSDRNPM is a transport code that
calculates cell-weighted cross-sections and keff for a one dimensional system.The
numerical output of the XSDRN calculations have further been processed by using
code XSCALC (Yapıcı, 2001) to determine neutronic parameters of blanket fuled
with different fuels mentioned above.
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3.2. Neutronic performance of the overall blanket
Integral neutronic data per DT fusion neutron in the overall investigated blanket
for different fuel types are given in Table 2. The energy multiplication factor, M, is
Table 2
Neutronic data per DT fusion neutron in the overall blanket
Type of fuel
ThO2
a
M
232
Thg
232
Thf
238
Ug
235
Uf
238
Uf
f
nsf
T6
T7
TBR
L
a
1.466
1.480b
0.165
0.125
0.0072
0.0072
–
–
–
–
–
–
0.0072
0.0072
0.0232
0.0232
1.104
1.144
0.060
0.060
1.164
1.204
1.476
1.476
0.0217
0.0218
ThC
UO2
UC
U3Si2
UN
1.481
1.492
0.184
0.140
0.00844
0.00844
–
–
–
–
–
–
0.00844
0.00844
0.0274
0.0274
1.097
1.141
0.061
0.061
1.158
1.202
1.459
1.459
0.0224
0.0225
1.870
1.899
–
–
–
–
0.163
0.119
0.00594
0.00682
0.0284
0.0285
0.0344
0.0354
0.114
0.117
1.157
1.202
0.060
0.060
1.217
1.262
1.412
1.393
0.0233
0.0235
2.046
2.088
–
–
–
–
0.213
0.160
0.00797
0.00943
0.0390
0.0392
0.0470
0.0486
0.156
0.160
1.150
1.204
0.060
0.060
1.210
1.264
1.445
1.420
0.0230
0.0235
1.972
2.003
–
–
–
–
0.178
0.130
0.00683
0.00780
0.0346
0.0347
0.0414
0.0425
0.138
0.141
1.163
1.211
0.060
0.060
1.223
1.271
1.409
1.390
0.0253
0.0256
2.053
2.097
–
–
–
–
0.220
0.169
0.00820
0.00986
0.0396
0.0398
0.0478
0.0497
0.159
0.163
1.133
1.185
0.059
0.059
1.192
1.244
1.472
1.442
0.0222
0.0225
Under resonance-free effect.
Under resonance-effect.
M=Energy multiplication ratio [total energy release (MeV)/14.1 MeV+1].
232
Thg=233U Breeding ratio.
238
Ug=239Pu Breeding ratio.
232
Thf=232Th(n,f).
235
Uf=235U(n,f).
238
Uf=235U(n,f).
f=total fission reaction.
nsf=fission neutron breeding ratio.
T6=6Li(n,)T.
T7=7Li(n,,n0 )T.
TBR=tritium breeding ratio.
(=peak-to-average fission power density ratio.
L=neutron leakage.
b
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defined as the ratio of the total energy release in the blanket to the incident fusion
neutron energy. Total energy release in blanket can be calculated as;
Total energy release in blanket ¼ 200f þ 4:784T6 2:467T7
ð1Þ
Whereas the lowest value of M (energy multiplication factor) is 1.480 in the blanket fueled with ThO2, the highest one is obtained as 2.097 in the blanket fueled with
UN by taking into account of resonance effect. M values of the blankets fueled with
thorium (as ThO2 and ThC) are smaller than that of with uranium (as UO2, UC,
Fig. 3. Fissile fuel breeding ratio per cm3 per (D,T) fusion neutron in the FFB zone for different fuels.
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U3Si2 and UN) since M mainly depends on the fission rate of the blanket and fission
cross section of 238U is higher than that of 232Th. And also, as it can be seen from
Table 2, total fission values of uranium fuels are much higher than that of thorium
fuels.
All values of the fissile breeding ratio per (D,T) neutron in the FFB zone of the
blanket are given in Table 2. Values of the 233U production ratio are 0.165 and 0.125
for the blanket fueled with ThO2 whereas 0.184 and 0.140 for the blanket fueled with
ThC under resonance-free and resonance-effect, respectively. UN has the highest
value of 239Pu breeding capability among the uranium fuels while UO2 has the lowest one. 239Pu production ratio changes from 0.119 to 0.169 according to the
Fig. 4. Fission rate per cm3 per (D,T) fusion neutron in the FFB zone for different fuels.
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uranium fuel types under the resonance-effect that lowers the (n,) reaction cross
section much as in the thorium fuels mentioned above. 239Pu breeding ratio values
are 0.163 and 0.119 for UO2, 0.213 and 0.160 for UC 0.178 and 0.130 for U3Si2,
0.220 and 0.169 for UN in the blanket under resonance-free and resonance-effect,
respectively.
Fig. 3 shows FFBR (fissile fuel breeding ratio) per cm3 per DT fusion neutron in
the FFB zone of the blanket fueled with six different fuels mentioned above. The
233
U and the 239Pu fissile fuel can be produced from 232Th(n,) and 238U(n,)
Fig. 5. Heat production per cm3 per MW/m2 first wall load in the FFB zone for different fuels.
H. Yapıcı et al. / Annals of Nuclear Energy 29 (2002) 1871–1889
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reactions, respectively. 233U breeding capability of ThC fuel is slightly higher than
that of ThO2 in the blanket. Resonance-effect in the calculation which causes the
decrease of 233U breeding capability of fuels is seen clearly although it causes a little
change in the fission rate of thorium fuels. Both 233U and 239Pu breeding ratios for all
fuel types under resonance-free and resonance-effect are the highest at a radius of 477.5
and sharply decline in radial direction towards to end of the blanket as shown in Fig. 3.
Fig. 6. Neutron spectrum per (D,T) fusion neutron in the FFB zone fueled with ThC under resonanceeffect. (Energy=(Ei+1-Ei)/lethargy, lethargy=ln(Ei+1/ Ei), i=1,2,.238.).
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All fission reaction values per (D,T) neutron in the FFB zone of the blanket are
also given in Table 2. Fission reactions of 232Th are 0.007 and 0.008 for the blanket
with ThO2 and ThC under resonance-effect, respectively. Fissions of 235U in the
blanket fueled with UO2, UC, U3Si2 and UN are 0.007, 0.009, 0.008 and 0.010,
respectively while fissions of 238U are 0.029, 0.039, 0.035 and 0.040 in the same order
Fig. 7. Neutron spectrum per (D,T) fusion neutron in the FFB zone fueled with UN under resonance-
H. Yapıcı et al. / Annals of Nuclear Energy 29 (2002) 1871–1889
1885
by considering resonance-effect. Although 235U is fissile, value of its fission rate is
lower than 238U because of highly energetic fusion neutron and low fraction of 235U
in uranium. The maximum total fission rate value is 0.048 reached in the blanket
fueled with UN when the lowest one is 0.007 reached in the blanket with ThO2
under resonance-effect. The total fission reaction values of the blankets fueled with
ThC and ThO2 are much smaller than that of the blankets with UO2, UC, U3Si2 and
UN, as expected.
Fig. 4 depicts fission rate per cm3 per DT fusion neutron in the FFB zone of the
blanket fueled with six different fuels under condition of with resonance-effect.
These fission rates are also maximum at a radius of 477.5 cm in the blanket and
decrease in radial direction towards to right side of the blanket as illustrated in
Fig. 4. It is shown that all FFBRs are greater than fission rates. Hence, the main aim
of the blanket is fissile fuel breeding rather than energy production.
TBR (tritium breeding ratio) values in the blanket for all type of fuels which are
greater than 1.05 means that tritium self-sufficiency is maintained for DT fusion
driver in all cases. TBR value changes from 1.158 to 1.223 under resonance-free
and from 1.202 to 1.271 under resonance-effect depending on the fuel types. The
maximum reached TBR value is 1.271 in the blanket fueled with U3Si2 under
resonance-effect.
Peak-to-average fission power density ratio, is a measure of spatial non-uniformly fission energy density that must be reduced to 1.0 for obtaining a flat fission power density. values are in the ranges of 1.409–1.476 and 1.393–1.476
depending on the type of fuels under resonance-free and resonance-effect, respectively. It is shown in Table 2 that under resonance-effect, its minimum value is
obtained as 1.393 in the blanket fueled with U3Si2 whereas its maximum value is
1.476 in the blanket fueled with ThO2.
SiC reflector reduces the neutron leakage out of the blanket drastically due to high
reflectivity of carbon so that neutron leakage out of the blanket, L, is quite low in all
cases. This value changes from 0.022 to 0.025 according to the fuel types.
3.3. Nuclear heat production
Nuclear heat released in the intervals of the FFB zone and calculated in MeV/cm3
per DT fusion neutron by the XSDRNPM code can be also calculated in W/cm3
according to first wall load and the blanket dimensions as follows:
qv ¼ Cf Pw qi Fw
ð2Þ
where,
qv=heat release in each interval of fuel zone in W/cm3
Cf=1.60219.1013 conversion factor in J/MeV,
Pw=first wall load in MW/m2,
=4.42657.1013 neutron flux in n/cm2s for 1 MW/m2 fusion neutron load,
qi=heat release in each interval of fuel zone in MeV/cm3 per DT fusion neutron,
Fw=2..R.L area of FW in cm2,
1886
H. Yapıcı et al. / Annals of Nuclear Energy 29 (2002) 1871–1889
R=radius of fusion chamber in cm,
L=height of fuel zone in cm.
In this work, heat production calculations are performed under a first wall load of
1 MW/m2. Fig. 5 depicts the heat production per cm3 per MW/m2 (D,T) fusion
neutron load in the FFB zone of the blanket for all fuels. The heat production has
the highest values at radius of 477.5 cm and then decreases sharply in radial direction to end of the FFB zone for all type of fuels. Its value mainly depends on the
fission rate in the FFB zone. Therefore it is not surprised that the heat values of
uranium fuels are much higher than thorium fuels. Its variation in radial direction is
due to non-uniformity of fission density caused by neutron spectrum softening
towards the outer region of the FFB zone. In other words, in the inner regions, the
neutron spectrum is harder and causes more fission reactions. And then it softens
due to collisions of the neutron with fuel materials, and causes the decrease in fission
rate. The maximum heat productions occur in the first row of the FFB zone for all
fuel types, and their values are 1.30, 1.28, 5.42, 7.60, 6.50 and 7.89 W/cm3 for ThO2,
ThC, UO2, UC, U3Si2 and UN under resonance-effect, respectively.
3.4. Neutron fluxes in the fissile fuel zone
Figs. 6 and 7 illustrate the spatial variation of neutron spectrums of ThC and UN
under only resonance-effect, respectively. Neutron spectrums of the other fuel compounds are not necessary to plot due to the fact that neutron spectrum of ThO2 is
similar to ThC and that of UN similar to UO2, UC and U3Si2. Generally, it is
observed that the fusion neutron energy decreases by deeper penetration in the
blanket. The fast neutron fluxes decrease in the radial direction while the lower
energy group fluxes increase because they are generated mainly beyond the fuel zone
and reflected back. In other words, the neutron flux curves show a variation towards
the outer boundary from the harder neutron spectrum shapes to the softer ones.
However, the neutron spectrum in the blanket fueled with ThC softens less than the
other fuels. It is also shown that the flux values increase in the range of 104–106 eV,
and reach 0.0005 n/cm2s due to collisions of neutron with other materials in the FFB
zone and release of fission neutrons having an average energy of 2 MeV.
4. Conclusions and recommendations
In this study, the neutronic performance of the PROMETHEUS-H fusion reactor
by using different type of fuels that are ThO2, ThC, UO2, UC, U3Si2 and UN has
been investigated. The main conclusions for this study are as follows:
1. The highest M value of the blanket reaches 2.097 by using fertile fuel in the
blanket. Higher M values are observed in the blanket fueled with uranium
fuels.
H. Yapıcı et al. / Annals of Nuclear Energy 29 (2002) 1871–1889
2.
3.
4.
5.
6.
1887
233
U breeding values for ThO2 and ThC are nearly the same. However, 239Pu
breeding values change depending on the type of fuels much. The blanket
fueled with UN has the highest 239Pu breeding value while UO2 has the lowest one.
Total fission values for uranium fuels are much higher than that for thorium
fuels in the blanket. The highest total fission value is found in the blanket
fueled with UN whereas the lowest one is obtained in the blanket ThO2.
TBR values are greater than 1.05 for all cases so that tritium production in
the blanket is sufficient for DT fusion driver.
( values change in the ranges of 1.409–1.476 and 1.393–1.476 depending on
the fuel types under resonance-free and resonance-effect, respectively due to
non-uniform fission density in the FFB zone.
The maximum heat production per cm3 per MW/m2 is obtained as 7.89 in the
FFB zone with UN fuel.
In conclusion, it is recommended that by using different type of fuels, the fissile
fuel breeding in the PROMETHEUS-H IFE reactor is possible. At the same time,
M can be increased to 2.097 means that substantial increment in energy production
is possible with respect to the original PROMETHEUS-H IFE Reactor.
For further studies, the blanket fueled with LWR and FBR spent fuels in the
PROMETHEUS IFE can be investigated to get better neutronic results. And also,
power flattening can be considered as a further research subject for this study.
Acknowledgements
This study is dedicated to our beloved colleague Dr. Nurettin Akçakoyun
(1967–2001) Engineering Faculty, Niǧde University, Turkey.
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