The Effect of Neutron Energy Spectrum on Actinide Management in High Temperature Reactors

Transcription

1 The Effect of Neutron Energy Spectrum on Actinide Management in High Temperature Reactors ABSTRACT Üner Çolak, Mehmet Türkmen Hacettepe University, Department of Nuclear Engineering Beytepe Campus, Ankara, Turkey Actinides play an important role in performance improvement and safety consideration of new type of nuclear fuels such as TRISO particle fuels in prismatic or pebble bed type of gas cooled high temperature reactors (HTR). One of the most important advantages of HTR with current core design capability is to limit the production of U, Pu, and other actinides, and the ability of burning weapon/reactor grade actinides by Deep Burn transmutation without permitting recovery. However, actinide depletion in fuel pebbles is directly related to neutron energy spectrum. In this respect, this study is aimed to find out the effect of neutron energy spectrum on actinide production/consumption and to calculate the composition of discharged fuel under various core configurations. The neutron transport calculations for group fluxes based on 44 group neutron energy cross section library is accomplished by a neutron transport code, SERPENT. The analysis indicates promising performance improvement on actinide management in HTR. 1 INTRODUCTION Energy-dependent flux is one of the important parameters in reactor physics calculations. Like thermal, fast, and epithermal reactor designs, the fuel utilization is optimized by the selection of proper neutron spectrum. As in CANDU designs, a hard spectrum can be important for actinide management since the power is produced mostly by induced fission of fissile plutonium. From the fuel utilization and actinide management points of view, it has already been questioned in HTR designs. In pebble-bed and prismatic core designs, neutron flux spectrum in the fuel region becomes more important for design purposes that is to burn actinides efficiently and to make a contribution to the non-proliferation of weapon-grade plutonium. In parallel with design purposes of HTR, this study is aimed to search the effect of neutron energy spectrum on actinide concentrations. Furthermore, this study provides analysis results for the infinite multiplication factor, isotopic content of actinides and activity depending on burnup level in fuel region when the neutron spectrum shift is performed. From the neutron moderation point of view, core effective multiplication factor and optimum burnup values are examined carefully to understand the performance of fuel under various neutron spectra. From the actinide management aspect, isotopes that are produced in large quantities under fast/thermal neutron spectra during fuel depletion are studied to determine accumulation rate of isotopes. From the final disposal/reprocessing option point of view, 312.1

2 312.2 spent fuel activity is calculated for various spectra. The analysis specifically targets spectral variations on actinides and their potential impacts on the HTR cores. Actinides play an important role in performance improvement and safety consideration of new type of nuclear fuels such as TRISOs in prismatic or pebble bed type of gas cooled high temperature reactors. For the transmutation of actinides, Jeong et al. [1] provided benchmark results including the accumulation of actinide, fission products, and tritium for HTR-10 reactor (10MWt prototype pebble bed reactor) constructed on BCC (Body Centred Cubic) unit cell. However, spectral effects were not included. For a specific design of VHTR (Very High Temperature Reactor), Tsvetkov et al. [2] worked on spectrum shifting to improve performance of the advanced actinide fuels and focused on TRU-impact on a singlebatch mode. SERPENT as a Monte Carlo reactor physics burnup calculation code system [3] is capable of dealing with HTR core analysis by using intrinsic geometry package in spherical coordinates, core definitions, and burnup calculation. Furthermore, depending on the considered model, the code gives extensive results about neutronic parameters, group constants, energy-dependent neutron flux distributions, and irradiated fuel inventory for any specified time. Point kinetic parameters, source normalization, B1 fundamental mode calculation, and diffusion parameters both for analog and P1 are the other important calculations. In addition to single machine use, the parallel capability is available. Although eigenvalue problems are solved very fast, excessive memory allocation during parallel calculation seems an important problem for low memory systems. 2 MODELING In this study, SERPENT developed at VTT Technical Research Centre of Finland since 2004 is used both for Monte Carlo analysis and fuel depletion calculations. Monte Carlo method results are verified by Leppänen et al. [4] by comparing with MCNP5, KENO-VI, and other code systems. Furthermore, depletion results are tested with the Monteburns2. The Monteburns2 code has been used many times in previous studies with sufficient accuracy in modelling of HTR core benchmarks. A comparative analysis for actinide inventory in a spent nuclear fuel for a pebble-bed HTR using three different MCNP-based depletion codes (Monteburns2, MCNPX2.6.0, and BGCore) is performed by Bomboni et al. [5]. Unit cell model -body centred cubic (BCC) with a packing fraction of is preferred instead of full core modelling in order to simplify the geometry. A visual presentation of considered geometry is given in Figure 1. Figure 1: Unit cell geometry modeling for calculations Spheres, called "fuel pebbles", in high-temperature gas-cooled reactors consist of numerous microscopic TRISO particles dispersed in a graphite matrix. For about 15,000 TRISO particles, the fuel pebbles with a diameter of 6.0 cm carry 9 g of slightly enriched UO 2 fuel on average. Material and geometry parameters are presented in Table 1.

3 312.3 Table 1: Material Properties and Geometry Descriptions Parameter Value Fuel Kernel Diameter 500 μm Particle Material Type UO 2 UO 2 Density 10.4 g/cm 3 Coating Material PyC/PyC/SiC/PyC Density of Materials in the Layer 1.05/1.9/3.2/1.9 g/cm 3 Average Pebble-Bed Packing Fraction BCC: 0.61 Fuel Enrichment 9.6% UO 2 Equilibrium Core Fuel Pebble Outer Radius 3.0 cm Thickness of Fuel Free Zone 0.5 cm Graphite Matrix and Fuel Free Zone Density 1.74 g/cm 3 Total HM Loading per Fuel Pebble 9 g (Equilibrium) Calculated Side Length of Cubical Lattice cm Calculated Lattice Cell Pitch BCC: 7.18 cm Coated Particle Radius 0.46 mm Packing Fraction of Coated Particles Number of Passes 6 Average Helium Temperature 771 o C Average Moderator Temperature 817 o C Average Fuel Temperature 830 o C The distribution of microscopic particles in a pebble is completely random in 3D dimensions; however, the number of TRISO particles per pebble is almost constant. Random distribution of the particles inside the entire volume is accomplished using stochastic package [6]. BCC unit cell as the regular arrangement of pebbles throughout the core is selected by agreeing to full core arrangement of pebbles. In this system, as well as core packing fraction is a function of core position [7], the average core packing fraction (or void coefficient) between the pebbles is taken as 0.61 instead of Such a system exactly consists of two pebbles: the single center pebble that is located at the center of volume within unit cell as a whole and the equivalent of one pebble from the eight corners. Lattice cell length is calculated based on the data for BCC. TRISO particles are placed at any location of a simple cubic unit cell to create a stochastic distribution [8]. Unit cell length is calculated by acting in accordance with packing fraction of [9]. The bilateral surfaces of BCC unit cell are connected by using periodic boundary condition to create an infinite system. Thus, neutrons leaving from any surface of the unit cell enter into the opposed surface without losing their trajectory. Neutron loss from the side surfaces of reactor core region including reflector region is defined as leakage reactivity due to the geometric buckling. The burnup calculation is performed by using non-linear leakage reactivity model defined by Driscoll et al. [9]. The fuel is exposed to neutron flux under constant power in multi-pass fuel-cycle followed by 6-passes of uranium fuel. Neutron flux spectrum is obtained for 44 neutron energy groups in fuel region that is enclosed by coating layers. The spectrum shifting is carried out by reducing the quantity of fuel pebbles in unit cell. Fuel pebbles are gradually replaced by moderator pebbles until there is no fuel pebble left in the unit cell. Carbon is excellent moderator for neutrons due to low absorption cross-section -hereby, they would provide extra moderation to the neutrons as they slow down-. Furthermore, all moderator pebbles made up of full graphite matrix without containing burnable poison and any fuel. Fuel pebbles (%) presented in this paper is the ratio of fuel pebble to the total pebble (fuel + moderator) in BCC unit cell. For instance, 100% indicates that all two pebbles are of the type of fuel pebbles.

4 RESULTS AND DISCUSSION Various weighted neutron flux spectra, which is obtained by dividing its total value, are illustrated in Figure 2 for various fuel pebbles (%) at zero-burnup. They are used in criticality, burnup, and actinide management analysis. The spectral differences come from the fast-tothermal flux ratio. It is effective on the reactivity. Figure 2: Energy-dependent neutron flux spectrum for various fuel pebbles (%) Using the spectra given above, criticality search is achieved. Figure 3 illustrates the variation of infinite multiplication factor with burnup for various fuel pebble fractions (%). According to the figure, the highest reactivity is obtained for fuel pebble of 25%; however, 100% with zero moderator pebble causes lowest reactivity. That is, moderator pebbles play important role in moderation. For the case of 12.5%, it is expected that U-235 is consumed rapidly, since the over-moderated systems with thermalized neutrons would be inherently absorbed by fissile uranium isotope because of a jump increase in absorption cross-section at low energy levels. Figure 3: k as a function burnup for various fuel pebbles (%)

5 312.5 Reactivity of the fuel through the core-life can be correlated both with fuel pebble fraction (%) and burnup. This relation is defined by Eq. (1) as a general form to obtain variation of reactivity for any burnup and fuel pebble fraction (%): l k i j ρ( B, f) = ρijb ( f) j= 0 i= 0 (1) where k and l are the degree of non-linear model to be used, i and j are the index number, B is the burnup in units of MWd/kgU, ρ ij are the unitless constants for any i and j values and f is the percent of fuel pebbles in unit cell. A high-order non-linear reactivity model with sufficient accuracy is used to obtain the results. The equation works well in all values of fuel pebble fraction and 0 and 95 MWd/kgU burnup. The twenty constants are computed by using least square fitting method and are listed in Table 2. Table 2: Fit Coefficients i/j E E E E E E E E E E E E E E E E E E E E-13 The effect of spectrum shift on discharge burnup is plotted in Figure 4. It is clear that discharge burnup as the indicator of fuel utilization is obtained about 110 MWd/kgU at the greatest. This value is optimized by fuel pebble of 65%. Furthermore, as the pass number increases in multipass fuel cycle, the fuel is utilized more effectively. Results show that there is a nonlinear relationship between discharge burnup and fuel pebble (%) in unit cell. Figure 4: Discharge burnup variation with fuel pebble (%) The other investigation is accomplished for the variation of isotopic constitute of actinides in spent fuel. Actinide content of spent fuel at discharge is plotted in Figure 5 and

6 312.6 Figure 6. From the results, quantities of certain isotopes ( 237 Np, 238 Pu, 242 Pu, 243 Am and Cm series) reach their maximum value for a certain spectra and then would decrease or become steady. However, as the spectrum becomes harder, some isotopes ( 239 Pu, 240 Pu, 241 Pu, 241 Am, and 242m Am) show a continuous increase with varying slope. All these behaviours can be attributed to the energy-dependent neutron cross-sections of isotopes and interaction types at that energy. It can be deduced that accumulation rate of isotopes as to its importance from the aspect of contamination or proliferation resistance can be simply adjusted. Figure 5: Isotope accumulation rate under various neutron flux spectra Figure 6: Isotope accumulation rate under various neutron flux spectrums Spent fuel activity variation with fuel pebbles (%) are showed in Figure 7 for some burnup values. Activity level increases as the fuel pebble (%) increases. High burnup results in high activity.

7 CONCLUSION Figure 7: Activity variation with fuel pebble (%) for various burnup The analysis indicates promising performance improvement on actinide management in HTR. As a result of analysis, the systems with higher fuel pebble (%) exhibit spectral shift effects towards harder neutron spectra with substantially reduced neutron populations at thermal energies. In the light of obtained results, following remarks are concluded. Softer spectrum yields higher positive reactivity; but, leads to lower fuel cycle length and rapid decrease in reactivity. A fuel cycle extension and low excess reactivity is obtained under hard spectrum. Fuel utilization is improved when the pass number of pebbles is increased. Using a certain spectrum, it is likely to limit to production of certain actinide isotopes regardless of fuel utilization. 237 Np is the principle source of long-term toxicity in high-level waste due to alpha decay with its half life of 2.14x10 6 years. It is also the source of principle alpha-emitter, Pu-238 for use as a heat source. 237 Np and 238 Pu increase as the fuel pebble fraction (%) increase. Fissile Pu fraction in total Pu becomes critical issue after fuel pebble fraction of 70%. Spectral effect appears itself in two ways: (i) fuel utilization improvement as a positive impact and (ii) increase of activity as a negative impact. Activity depends both on burnup level and fuel pebble fraction (%). The improved fuel depletion would allow for extended operation on fuel loading strategy up to lifetimes that are constrained by structural performance limits because of thermal loads, mechanical stresses and attained fast fluences. HTR core design concepts with pebble bed and prismatic block assembly give fuel management flexibility because of easy adjustment of neutron spectrum.

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