Abstracts and Available Papers Presented at the
1995 International RERTR Meeting
Comparison of the FRM-II HEU Design With an Alternative LEU Design*
S. C. Mo, N. A. Hanan and J. E. Matos
Argonne National Laboratory
Argonne, Illinois 60439-4841 USA
Presented at the1995 International Meeting on
Reduced Enrichment for Research and Test Reactors
September 18-21, 1995
*Work supported by the US Department of Energy
Office of Nonproliferation and National Security
under Contract No. W-31-109-38-ENG
The FRM-II reactor design of the Technical University of Munich has a compact core that utilizes fuel plates containing highly-enriched uranium (HEU, 93%). This paper presents an alternative core design utilizing low-enriched uranium (LEU, <20%) silicide fuel with 4.8 g/cm3 that provides nearly the same neutron flux for experiments as the HEU design, but has a less favorable fuel cycle economy. If an LEU fuel with a uranium density of 6.0 6.5 g/cm3 were developed, the alternative design would provide the same neutron flux and use the same number of cores per year as the HEU design.
The results of this study show that there are attractive possibilities for using LEU fuel instead of HEU fuel in the FRM-II. Further optimization of the LEU design and near-term availability of LEU fuel with a uranium density greater than 4.8 g/cm3 would enhance the performance of the LEU core. The RERTR Program is ready to exchange information with the Technical University of Munich to resolve any differences that may exist and to identify design modifications that would optimize reactor performance utilizing LEU fuel.
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The FRM-II reactor design of the Technical University of Munich
is designed for the production of high intensity thermal neutrons
for use in a wide variety of applications in structural research
and spectroscopy. The HEU design is characterized by a compact
core and a moderate power level of 20 MW, which results in a high
flux to power ratio. The general concepts of compact reactor design
can be found in References 1 and 2. In a previous study, a successive
linear programing technique was used to optimize a core design
(Ref. 3) using LEU silicide fuel.
In this study, the design objectives for the LEU core were to match both the peak thermal flux (8 x 10(14) n/cm2/s) and the cycle length (50 days) of the FRM-II HEU design using a two-stage approach. In the first stage, LEU silicide fuel with a uranium density of 4.8 g/cm3 was used to obtain the same technical performance and an acceptable economic performance in a core with a higher power level than the HEU design. In the second stage, LEU fuel with a higher uranium density was substituted into the same core geometry and the reactor power level was increased slightly so that both the peak neutron flux and the cycle length matched those of the HEU design. This approach assumes that LEU fuel with a uranium density greater than 4.8 g/cm3 will be successfully developed.
REACTOR DESIGNS AND MODELS
Schematic diagrams of the FRM-II HEU core design and of the alternative
LEU core design are shown in Figures 1 and 2. Key design and performance
parameters are listed in Table 2. The FRM-II HEU core design consists
of 113 involute fuel plates containing 7.5 Kg of U-235 in 93%
enriched uranium. The core is cooled by light water and is surrounded
by a heavy water reflector. The reactor is controlled at the center
of the core using a hafnium control rod with a beryllium reflector
follower. Power peaking is reduced by grading the fuel meat in
each plate into two regions with uranium densities of 3.0 and
1.5 g U/cm3. Additional power flattening is achieved by placing
a boron ring containing 6 grams of natural boron near the bottom
of the core. This ring has a relatively small reactivity worth
of about 0.5% dk/k in the fresh core.
The LEU design follows the same concept as the HEU design, but has a larger diameter and higher core that contains 153 involute plates. Since the average and peak power densities in the larger LEU core are considerably lower than those in the FRM-II HEU core, fuel grading has not been incorporated into the LEU design. However, fuel grading could be introduced if it is needed.
Diffusion theory calculations were performed for each reactor design using the DIF3D code and 15 energy-group cross sections generated using the WIMS-D4M code and ENDF/B-V data (Ref. 4). Monte Carlo calculations were performed using the MCNP code (Ref. 5) and ENDF/B-V data to validate the results of the diffusion theory calculations and to calculate the control rod worth. The MCNP core models were represented by concentric fuel rings that preserved the total uranium loading, the meat, clad and coolant channel thicknesses.
A comparison of eigenvalues and peak thermal fluxes in the reflector that were obtained from the diffusion theory and Monte Carlo calculations are shown in Table 1. Peak thermal fluxes are expressed in the form of Keff x (thermal flux) to account for the movement of the control rod. The diffusion theory calculation underpredicted the reactivity of the HEU design by about 0.7% dk/k. Much better agreement was obtained in the LEU case. The peak thermal fluxes obtained from the Monte Carlo and diffusion theory calculations are in reasonably good agreement.
Table 1. Comparison of MCNP and DIF3D Eigenvalues and Peak Thermal Fluxes in the Reflector for the FRM-II HEU Design and the Alternative LEU Design with 4.8 g/cm3 Silicide Fuel
HEU (20 MW)
HEU (20 MW)
LEU (30 MW)
LEU (30 MW)
|Keff (no B10)|
1.2000 ± 0.0008
1.2079 ± 0.0014
|Keff (with B10)|
1.1937 ± 0.0006
|Keff x (th flux) (n/cm2/s)|
8.0 x 10(14)
7.6 x 10(14) ± 0.3%
7.8 x 10(14)
7.5 x 10(14) ± 0.6%
Depletion calculations were performed for both the HEU and LEU cores using the REBUS-3 code (Ref. 6) assuming an end-of-cycle reactivity of 7% dk/k. A detailed 19 fission-product-chains model was used in the depletion calculations to describe the buildup of fission products in the reactor (Ref. 7). The depletion calculations were performed with the control rod at its fully withdrawn position.
Table 2. Key Parameters in FRM-II HEU Design and Alternative LEU Design
|Number of Fuel Plates|
|Core Height (cm)|
|Core Inner - Outer Radius (cm)|
6.75 - 11.2
9.78 - 16.04
9.78 - 16.04
|Core Volume (liters)|
|Length of Involute Plate (cm)|
|Fuel Meat/Clad Thickness (mm)|
|Coolant Channel Thickness (mm)|
|Fuel Meat Uranium Density (g/cm3)|
|Core Power (MW)|
|Core Loading (Kg U-235)|
|Keff at BOC|
|Cycle Length (Full Power Days) (a)|
|Average Number of Cores/Year (b)|
|Average Burnup (% U-235 burned)|
|Average Fission Rate in Fuel Meat
Peak Rate in Fuel Meat (fissions/cm3/s)
2.1 x 10(14)
1.8 x 10(14)
2.0 x 10(14)
|Average Fission Rate: Fuel Particles |
Peak Rate in Fuel Particles
7.9 x 10(14) (c)
4.2 x 10(14)
3.5 x 10(14)
|Average Fission Density in Fuel Meat|
1.0 x 10(21)
0.45 x 10(21)
0.78 x 10(21)
|Average Power Density
Peak Power Density - rod out (W/cm3)
|Peak Thermal Flux, |
Keff x (max flux) (n/cm2/s)
8.0 x 10(14)
7.8 x 10(14)
8.2 x 10(14)
|Reflector Volume (liters) with
Keff x flux > 7 x 10(14) n/cm2/s
(a) EOC excess reactivity = 7% dk/k; (b) Based on 250 days operation per year; (c) In 3.0 g/cm3 fuel of the HEU design.
COMPARISON OF REACTOR PERFORMANCE
Key performance parameters of the FRM-II HEU and the alternative LEU design are shown in Table 2 and are summarized in Table 3. Thermal flux distributions at the core midplane are compared in Figure 3.
Table 3. Summary Comparison of Performance for the FRM-II HEU Design and the Alternative LEU Design
FRM-II HEU Design
|Uranium Density, g/cm3|
|Power Level, MW|
|Peak Neutron Flux, n/cm2-s|
8.0 x 10(14)
7.8 x 10(14)
8.2 x 10(14)
|Cycle Length (Full Power Days)|
|Number of Cores per Year|
The LEU design with both 4.8 and 6.4 g/cm3 fuels can be further
optimized to improve reactor performance. For example, the LEU
fuel meat thickness can be increased from 0.51 mm to the 0.60
mm thickness of the HEU design. With 4.8 g/cm3, this change would
result in a cycle length that is estimated to be 33-35 days requiring
7 - 8 cores per year. The LEU density needed to match the neutron
flux and cycle length performance of the HEU core would change
from 6.4 g/cm3 to about 6.0 g/cm3.
The LEU design is capable of producing nearly the same intensity of thermal fluxes in the outer reflector region as the HEU design. A comparison of effective volume in the high flux region (locations with keff x (thermal flux) > 7x10(14) n/cm2-s) in the heavy water reflector shows that the LEU design with an advanced fuel offers considerably more usable volume for the installation of experimental facilities.
Although thermal-hydraulic studies have not been performed for the LEU design, the lower power densities and larger coolant channel suggest that the heat transfer requirement of the LEU core are likely to be less stringent than in the HEU design.
The excess reactivity during the reactor operation is controlled
by the movement of a central control rod with a beryllium follower.
The control rod in the HEU design consists of a cylindrical column
of aluminum covered with a 0.25 cm thick layer of hafnium absorber.
The HEU core has a excess reactivity of 16.2% dk/k at the beginning
of cycle. Assuming the combined reactivity worth from the experimental
facilities, temperature coefficients and reactivity reserve is
about -7% dk/k, a minimum control rod worth of about -10% dk/k
will be needed to control the reactor. The worth of the control
rod at fully inserted position was calculated to be about -15
In the LEU cores, the interior surface of the core is much larger than in the HEU design. This large surface affords many possible designs for the control rod.
The results of this study show that there are attractive possibilities for using LEU fuel instead of HEU fuel in the FRM-II. A two-stage approach was used to identify a core design that would allow the use of LEU fuel and still have the same peak thermal flux available for experiments and the same cycle length as in the HEU design. In the first stage, LEU silicide fuel with a uranium density of 4.8 g/cm3 was used to obtain the same technical performance and an acceptable economic performance in a core with a higher power level than the HEU design. In the second stage, LEU fuel with a higher uranium density was substituted into the same core geometry and the reactor power level was increased slightly so that both the peak neutron flux and the cycle length matched those of the HEU design. The LEU design can be further optimized to improve its performance.
This approach assumes that LEU fuel with a uranium density in the range of 6.0 - 6.5 g/cm3 will be successfully developed. There are good indications (Ref. 8) that LEU silicide fuel with 6.0 g/cm3 is feasible, although the testing is not complete to our knowledge. Other fuels (Ref. 9) may also be successfully developed. If the RERTR advanced fuel development effort begins as scheduled in October 1995, we are optimistic that a fuel with 6.0 - 6.5 g U/cm3 will be successfully developed and licensed.
Only by changing the current HEU core design is it possible to use LEU fuel in the FRM-II. An LEU fuel that could be substituted for the HEU fuel in the current FRM-II HEU core geometry and have comparable flux performance and fuel cycle economics would require a uranium density greater than 16 g/cm3, which is not feasible. However, as shown in this study, alternative FRM-II core designs can be developed in which feasible LEU fuels can be used.
The RERTR Program is ready to exchange information with the Technical University of Munich to resolve any differences that may exist and to identify design modifications that would optimize reactor performance using LEU fuel.
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Munich Compact Core Design", Proc. 1988 Int. Reactor Physics
Conferences, ANS Vol II 1988.
2. J. M. Ryskamp, D. S. Selby and R. T. Ptimms III, "Reactor Design of theAdvanced Neutron Source", Nucl. Tech., Vol 93 Mar. 1991.
3. S. C. Mo, "Application of the Successive Linear Programming Technique to the Optimum Design of a High Flux Reactor Using LEU Fuel", 14th International RERTR Meeting, Jakarta, Indonesia, 1991.
4. J. R. Deen, W. L. Woodruff and C. I. Costescu, "WIMS-D4M User Manual Rev. 0", ANL/RERTR/TM-23, Argonne National Laboratory, July 1995.
5. J. F. Briesmeister, "MCNP-A General Monte Carlo N-Particle Transport Code", LA-12625-M, Los Alamos National Laboratory, 1993.
6. B. J. Toppel, "A User's Guide for the REBUS-3 Fuel Cycle Analysis Capability", ANL-83-2, March 1983.
7. S. C. Mo, "Methodology and Application of the WIMS-D4M Fission Product Data", Proceedings of the 1994 International Meeting on Reduced Enrichment for Research and Test Reactors, September 17-23, 1994, Williamsburg, Virginia, USA, to be published.
8. Durand J. P. and Fanjas Y., "LEU Fuel Development at CERCA: Status as of October 1993", Proceedings of the 16th International Meeting on Reduced Enrichment for Research and Test Reactors, October 4-7, 1993, Oarai, Japan, JAERI-M 94-042, March 1994.
9. J.P. Durand - P. Laudamy (CERCA) and K. Richter (European Commission, Joint Research Center, ITU), "Preliminary Development of MTR Plates with Uranium Nitride", Proceedings of the 1994 International Meeting on Reduced Enrichment for Research and Test Reactors, September 17-23, 1994, Williamsburg, Virginia, USA, to be published.