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 maraczy@sunserv.kfki.hu
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/cm 3 ] for an
HPLWR fuel assembly. One can see the drastic change of k inf
vs. moderator density at low coolant density.
Kinf curves
1.22
1.20
1.18
1.16
1.14
1.12
1.10
1.08
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
Moderator density [g/cm3]
0.9
1.0
1.1
Figure 2
The effect of water densities on the multiplication factor.
SS clad and thin walls,
No rods, T fuel
=1000 [K], Bu=0., Equilibrium Xe, Sm, E=6%, 5%
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

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 boxes
MCNP k
∞
St. deviation
SS thick
SS-ZrO
2
-SS sandwich thick
SS thin
SS-ZrO
2
-SS sandwich thin
1.14670
1.19781
1.25567
1.28517
0.00027
0.00027
0.00027
0.00027
Table 1
Infinite multiplication factors with SS cladding
The basic enrichment of fuel: 7%
Coolant and moderator densities: 200 and 600 [kg/m
3
], T fuel
=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
1.05
1.00
0.95
0.90
0.85
0.80
0.75
0.70
0.65
1.25
1.20
1.15
1.10
0.60
0.55
0.0E+00
Thick, no absorber
Thick, absorber
Thin, no absorber
Thin, absorber
1.0E+04 2.0E+04 3.0E+04
Burnup [MWday/tU]
4.0E+04 5.0E+04 6.0E+04
Figure 3
Comparison of thinner box fuel assembly burnup to the thick one.
SS clad and walls. T fuel
=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 B
4
C 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 (k k
A
, k k
B
, k k
C
).
Case: k eff k k k k
A k k
B k k
C
Assembly, 0 Gd pin, no rod
Assembly, 4 Gd pins, no rod
1.221 1.053 -
1.056 1.125 -
Assembly, 4 Gd pins, inserted rod, natural boron 0.718 1.243 -
Assembly, 4 Gd pins, inserted rod, 40% B-10 0.665 1.245 -
-
-
-
-
-
-
-
-
Assembly, 4 Gd pins, inserted rod, 60% B-10 0.634 1.251 -
Assembly, 4 Gd pins, inserted rod, 80% B-10 0.609 1.255 -
-
-
-
-
Cluster, 4 Gd rods, inserted rod, natural boron 0.861 1.393 1.212 1.288 1.294
Cluster, 4 Gd rods, inserted rod, 40% B-10
Cluster, 4 Gd rods, inserted rod, 60% B-10
0.833 1.415 1.216 1.308 1.300
0.813 1.434 1.227 1.304 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],
Gd
2
O
3
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 % Gd
2
O
3
content. Figure 5 shows the infinite multiplication factors of the 6% base enrichment version.
Enrichment: 6%
0 Gd pin
8 Gd pins: 2%
4 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
Gd
2
O
3
content affects the time length of reactivity compensation. On the basis k inf
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,
235
U,
238
U,
239
Pu,
135
Xe and
149
Sm 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.
Coolant Temp.
Moderator Temp.
Gap Temp.
900
850
800
750
700
650
600
550
0 5 10 30 35 40 15 20
Axial Node
25
Figure 6
Water temperatures

Coolant Temp.
Moderator Temp.
Gap Temp.
900
850
800
750
700
650
600
550
0 5 10 15 20
Axial Node
25 30 35 40
Figure 7
Water temperatures
Coolant Temp.
Moderator Temp.
Gap Temp.
900
850
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 k eff
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: dρ/dT in
=-47 [pcm/K]
Fuel temperature reactivity coefficient: dρ/dT f
Effective delayed neutron fraction:
=-2.3 [pcm/K]
β eff
=7.12E-03[-]
Low rod worth was observed with natural boron with the B
4
C 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.
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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,
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3.
T. Schulenberg, S. Himmel, 4 July 2007 (personal communication)
4.
J. Starflinger et al.: "European Research Activities within the Project: High
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http://www.cheresources.com/iapwsif97.shtml
6.
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7.
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8.
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9.
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