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HIGH PERFORMANCE LIGHT WATER REACTOR CORE CALCULATIONS
Cs. Maráczy, Gy. Hegyi, G. Hordósy, E. Temesvári, Cs. Hegedűs, A. Molnár
KFKI Atomic Energy Research Institute, H-1525 Budapest 114, POB 49, Hungary
[email protected]
ABSTRACT
The High Performance Light Water Reactor (HPLWR) belongs to the six reactor types
currently being investigated within the framework of the Generation IV International Forum:
the Supercritical Water Cooled Reactor [1]. The paper summarizes the activity of KFKI
Atomic Energy Research Institute in the field of core design carried out within the framework
of the "HPLWR Phase 2" project.
INTRODUCTION
The objective of the HPLWR project is to assess the feasibility of a high efficiency light water
reactor operating at thermodynamically supercritical region. In a once-through concept (see
figure 1.), the water enters the reactor as water and exits as high pressure steam without
change of phase.
Fig.1
The Supercritical Water Cooled Reactor
With low investment costs an efficiency of about 44% is expected. Seven research institutes,
two universities and an industrial partner works on the concept. The main task of KFKI-AEKI
is the core design. Besides burnable absorbers cluster type control rods are applied to
compensate the excess reactivity at the beginning of cycle as no boric acid reactivity control is
allowed. As the coolant density along the axial direction shows remarkable change, coupled
neutronic-thermohydraulic calculations are essential which take into account the heating of
moderator in the special water rods of the assemblies. The most visible advantages of the
HPLWR are the low construction costs because of size reduction of components and
buildings compared to current LWR and the low electricity production costs due to high
efficiency.
NEUTRONIC MODELING OF THE HPLWR FUEL ASSEMBLIES
The coolant density at the top of the HPLWR assemblies is very low, so to reach sufficient
moderation in the tight-lattice assemblies the application of water rods is necessary. To
minimize the relative amount of structural materials, small fuel assemblies were proposed
which are arranged in fuel clusters containing 9 assemblies [2]. In the series of the HPLWR
fuel assembly proposals, the fuel assembly with thick, solid stainless steel (SS) moderator and
FA boxes resulted in impermissibly low multiplication factor. The parasitic absorption of
structural materials could allow only extremely short cycle length. To resolve the problem the
application of thinner steel box walls with stiffeners and with zirconium-oxide heat insulation
has been emerged. The heat insulation helps to avoid the extreme heat up of moderator, so it
provides better moderation for neutrons. In figure 2 the infinite multiplication factor curves
vs. moderator density can be seen for fixed coolant densities from 0.065 to 1.00 [g/cm3] for an
HPLWR fuel assembly. One can see the drastic change of kinf vs. moderator density at low
coolant density.
Kinf curves
1.22
1.20
1.18
1.16
1.14
1.00
0.90
0.80
0.70
0.60
0.50
0.40
0.30
0.25
0.20
0.15
0.10
0.065
1.12
1.10
1.08
Kinf
1.06
1.04
1.02
1.00
0.98
0.96
0.94
0.92
0.90
0.88
0.86
0.0
0.1
0.2
0.3
0.4
0.5
0.6
0.7
0.8
0.9
1.0
1.1
Moderator density [g/cm3]
Figure 2
The effect of water densities on the multiplication factor.
SS clad and thin walls,
No rods, Tfuel=1000 [K], Bu=0., Equilibrium Xe, Sm, E=6%, 5%
The results of the MCNP Monte Carlo code parametric studies on different wall types can be
seen in table 1. The application of the thin stainless steel wall with stiffeners and insulation is
preferable. In the HPLWR Phase 2 project at last the application of a special fiber insulation
was accepted [3].
Type of box walls
Moderator and fuel assembly
MCNP k∞
St. deviation
boxes
SS
thick
1.14670
0.00027
SS-ZrO2-SS sandwich
thick
1.19781
0.00027
SS
thin
1.25567
0.00027
SS-ZrO2-SS sandwich
thin
1.28517
0.00027
Table 1
Infinite multiplication factors with SS cladding
The basic enrichment of fuel: 7%
Coolant and moderator densities: 200 and 600 [kg/m3], Tfuel=1200 [K]
In figure 3 the neutronic advantage of the of thinner box fuel assembly over the thicker one
can be seen as a function of burnup.
Comparison of thick and thin FA designs
1.25
1.20
1.15
1.10
1.05
1.00
Kinf
0.95
Thick, no absorber
Thick, absorber
Thin, no absorber
Thin, absorber
0.90
0.85
0.80
0.75
0.70
0.65
0.60
0.55
0.0E+00
1.0E+04
2.0E+04
3.0E+04
4.0E+04
5.0E+04
6.0E+04
Burnup [MWday/tU]
Figure 3
Comparison of thinner box fuel assembly burnup to the thick one.
SS clad and walls. Tfuel=1000 K. E=6%, 5%, no Gd absorber
Assessing the effect of control rods some parametric studies were carried out. We had a
geometric constraint: The control rods can be inserted only into 5 assemblies of the 3x3 fuel
cluster. Similarly to the VVER-1000 reactor, stainless steel cladded B4C control rods were
assumed (See figure 4).
A
B
C
Figure 4
The 3x3 fuel assembly cluster with absorber rods.
A series of calculations were done to calculate the rod worth and power peaking factors,
where the enrichment of boron in B-10 was increased. In table 2 on the basis of the cluster
calculations the power peaking factors normalized for assemblies A, B and C are also
indicated (kkA , kkB , kkC).
Case:
keff
kk
kkA
kkB
kkC
Assembly, 0 Gd pin, no rod
Assembly, 4 Gd pins, no rod
Assembly, 4 Gd pins, inserted rod, natural boron
Assembly, 4 Gd pins, inserted rod, 40% B-10
Assembly, 4 Gd pins, inserted rod, 60% B-10
Assembly, 4 Gd pins, inserted rod, 80% B-10
Cluster, 4 Gd rods, inserted rod, natural boron
Cluster, 4 Gd rods, inserted rod, 40% B-10
Cluster, 4 Gd rods, inserted rod, 60% B-10
1.221
1.056
0.718
0.665
0.634
0.609
0.861
0.833
0.813
1.053
1.125
1.243
1.245
1.251
1.255
1.393
1.415
1.434
1.212
1.216
1.227
1.288
1.308
1.304
1.294
1.300
1.264
Cluster, 4 Gd rods, inserted rod, 80% B-10
0.794 1.421 1.211 1.312 1.259
Table 2
Infinite multiplication factors and power peakings.
SS clad and insulated SS walls. Basic enrichment: 7%,
Coolant and moderator densities: 200 and 600 [kg/m3], Tfuel=1200 [K],
Gd2O3 content of Gd rods: 4 [w/o].
As it can be seen from the results, raising the boron enrichment up to 80%, the rod worth can
be raised roughly by 50%, which gives potential possibility to shut down the core. The pinwise power peaking factors are not affected considerably by raising the B-10 enrichment.
GD BURNABLE ABSORBER CALCULATIONS
The preliminary optimization of Gd burnable poisons is possible with applying 2D transport
calculations. Considerable reactivity must be compensated because of the lack of boron
control. For assessing the cycle length and the Gd reactivity compensation 2-D transport
calculations vs. burnup at nominal reactor conditions were made. The basic enrichment of
assemblies varied from 3 to 7%. The number of Gd containing rods was 0, 4 and 8 per
assembly with 2 and 4 % Gd2O3 content. Figure 5 shows the infinite multiplication factors of
the 6% base enrichment version.
k∞
Enrichment: 6%
0 Gd pin
4 Gd pins: 2%
8 Gd pins: 2%
8 Gd pins: 4%
4 Gd pins: 4%
1.25
1.20
1.15
1.10
1.05
1.00
0.95
0.90
0.85
0.80
0
10000
20000
30000
40000
50000
60000
Burnup [MWd/tU]
Figure 5
Burnup calculations of the 6% basic enrichment versions at nominal parameters.
The number of Gd rods affects the value of initial reactivity compensation, whereas the
Gd2O3 content affects the time length of reactivity compensation. On the basis kinf curves 30
[MWday/kgU] discharge burnup was foreseen as a rough estimate, detailed 3D calculations
will be carried out later.
3-D CORE CALCULATIONS
As the target average outlet temperature (500 C) and the maximum foreseen cladding
temperature (620 C) are very close to each other, the 3-pass core concept was proposed [4].
The hot spots in this proposal can be eliminated by multiple flow of coolant through the active
core with mixing after each passing.
The KFKI Atomic Energy Research Institute uses the KARATE-SPROD coupled code for
core design. In the SPROD code the IAPWS-IF97 water property functions [5], and the Watt
[6] and Jackson-Hall [7] heat transfer correlations are used presently.
The KARATE-SPROD code system [8, 9] uses the parametrized 2-group cross sections as a
function of burnup, absorber cluster insertion factor, 235U, 238U, 239Pu, 135Xe and 149Sm
concentration, coolant, moderator rod and gap water density and fuel temperature. The
parameter range covers the cold zero power and hot full power states.
As an example, figures 6-8 will be presented below, which illustrate how the water
temperatures vary in 3 representative fuel assemblies without thermal insulation in the 3-pass
core.
Evaporator, FA 8
Coolant Temp.
Moderator Temp.
Gap Temp.
900
850
T [K]
800
750
700
650
600
550
0
5
10
15
20
Axial Node
Figure 6
Water temperatures
25
30
35
40
Superheater 1, FA 14
Coolant Temp.
Moderator Temp.
Gap Temp.
900
850
T [K]
800
750
700
650
600
550
0
5
10
15
20
25
30
35
40
Axial Node
Figure 7
Water temperatures
Superheater 2, FA 17
Coolant Temp.
Moderator Temp.
Gap Temp.
900
850
T [K]
800
750
700
650
600
550
0
5
10
15
20
25
30
35
40
Axial Node
Figure 8
Water temperatures
In the first and second superheater the temperature of moderator and gap water reaches or
exceeds the pseudocritical temperature resulting in poor moderation. Raising the mass flow
fractions, the keff is practically the same because the heat transfer coefficient from box wall to
moderator increases significantly with higher mass flux, so the density increase in these
regions is rather limited. Applying heat insulators helps to increase the local multiplication in
the evaporators which favorably affects the power distribution and k eff of the core. A
preliminary first loading of the HPLWR core was assessed, which contains insulated
assemblies with and without Gd. The initial reactivity excess is compensated by control rods.
The core arrangement contains 3x52 assembly clusters, the core outlet temperature is 500 C,
the burnup is zero. At full power conditions equilibrium Xe was applied. The radial power
distribution can be seen in figure 9.
Figure 9
The power distribution of a possible first loading of the
HPLWR 3 pass core.
XX: Coolant flow paths: 0: up, 1: down, 2: up
YY: Radial power distribution
The main result of the calculation from the point of view of assessing the linear pin powers:
The maximum linear pin power including the assessed power peaking inside the assembly is
283.0 [W/cm], which contains 1.378 reserve from the 390 [W/cm] limit. It should cover the
calculational, technological uncertainties and maneuverability. The relation to the maximum
cladding temperature will be checked by hot channel calculations in the project. Equilibrium
cycle calculations will be done in the future to establish frame parameters usable in safety
analysis. The reinforced, heat insulated fuel assembly design gives chance to solve the
reactivity problem by less parasitic absorption and improves the power at the peripheral
assemblies by higher moderation.
Some neutronic parameters of the potential first loading at full power:
Inlet temperature reactivity coefficient:
Fuel temperature reactivity coefficient:
Effective delayed neutron fraction:
dρ/dTin =-47 [pcm/K]
dρ/dTf =-2.3 [pcm/K]
βeff
=7.12E-03[-]
Low rod worth was observed with natural boron with the B4C rods depicted in figure 4, but
the reactor can be shut down, provided all the clusters have control rod drives [10]. The
square absorber rod design of FZK will provide higher rod worths, so that geometry will be
preferred in later studies. The reactivity coefficients of HPLWR with respect to temperatures
show substantial negative feedback at normal operating conditions, and together with the
kinetic parameters are close to that of the PWR’s at corresponding conditions.
FUTURE PROSPECTIVES
Our future goal is the analysis of HPLWR assemblies and core to improve core performance
and to calculate basic safety features with the KARATE code system.
ACKNOWLEDGEMENTS
The authors would like to thank the European Commission and the Hungarian National Office
for Research and Technology for their financial support.
REFERENCES:
1. A Technology Roadmap for Generation IV Nuclear Energy Systems, U.S. DOE
Nuclear Energy Research Advisory Committee and the Generation IV International
Forum, December 2002.
2. J. Hofmeister, T. Schulenberg, J. Starflinger: "Optimization of a fuel assembly for a
HPLWR" International Congress on Advances in Nuclear Power Plants, Seoul,
KOREA, May 15-19, 2005, Proceedings of ICAPP ’05, Paper 5077
3. T. Schulenberg, S. Himmel, 4 July 2007 (personal communication)
4. J. Starflinger et al.: "European Research Activities within the Project: High
Performance Light Water Reactor Phase 2 (HPLWR Phase 2)" International Congress
on Advances in Nuclear Power Plants, Nice, FRANCE, May 13-18, 2007, Proceedings
of ICAPP ’07, Paper 7146
5. http://www.cheresources.com/iapwsif97.shtml
6. Yamaji, T. Tanabe, Y. Oka, J. Yang, J. LIU, Y. Ishiwatari, S. Koshizuka, Evaluation
of the Nominal Peak Cladding Surface Temperature of the Super LWR with
Subchannel Analyses, Proceeding of GLOBAL 2005, Tsukuba, 2005.
7. J. D. Jackson and W. B. Hall, "Forced convection heat transfer to fluids at
supercritical pressure," Turbulent Forced Convection in Channels and Bundles edited
by S. K a k a c and D. B. Spalding, Vo1.2, Hemisphere Pub., 563 (1979).
8. Cs. Hegedűs, Gy. Hegyi, G. Hordósy, A. Keresztúri, M. Makai, Cs. Maráczy, F.
Telbisz, E. Temesvári, P. Vértes: The KARATE Program System, PHYSOR 2002,
Seoul, Korea, October 7-10, 2002
9. Yamaji, Y. Oka, S. Koshizuka: Three-dimensional Core Design of SCLWR-H with
Neutronic and Thermal-hydraulic Coupling, GLOBAL2003, New Orleans, USA, Nov.
2003
10. Maráczy Csaba, Hegyi György, Hordósy Gábor, Temesvári Emese, Hegedűs Csaba,
Molnár Attila: HPLWR Core Design Calculations, 2007, AEKI Report (In Hungarian)
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