PWR AND WWER MOX BENCHMARK CALCULATION BY HELIOS Radoslav ZAJAC 1,2), Petr DARILEK 1), Vladimir NECAS 2) 1 VUJE, Inc., Okruzna 5, 918 64 Trnava, Slovakia; zajacr@vuje.sk, darilek@vuje.sk 2 Slovak University of Technology, Faculty of Electrical Engineering and Information Technology, Ilkovicova 3, Bratislava, Slovakia; vladimir.necas@stuba.sk ABSTRACT A depletion calculation benchmark devoted to MOX fuel cycles in an ongoing objective of OECD/NEA WPRS (Working Party on Scientific Issues of Reactor Systems) following the study of depletion calculation concerning UOX fuel [1]. Four corresponding fuel structures as super-cells were calculated: the MOX PWR, WWER-1000, WWER-440 assembly surrounded by UOX assembly and MOX assembly in an infinite lattice. These scenarios enable to show the importance of the UOX environment in the mixed UOX - MOX core. Studies have shown that a standard calculation based on an infinite medium pattern for MOX assembly can lead to a significant error in Pu-239 concentration at a high burn up. The model calculations were performed by a spectral code HELIOS 1.10. 1. INTRODUCTION The MOX is a means to re-use plutonium remaining in spent reactor fuel to provide electricity generation [2]. The MOX production demands plutonium and the remaining uranium separation (about 97 % of the spent fuel) from the fission products (FPs) (together about 3 %). Then plutonium needs to be separated from most or all of the uranium. All this is undertaken at a reprocessing plant. Plutonium as an oxide is mixed with depleted uranium left over from an enrichment plant to form fresh mixed oxide fuel (MOX = UO 2 + PuO 2 ). The MOX fuel, consisting of about 7-9 % Pu mixed with depleted U, is equivalent to UOX fuel enriched to about 4.5 % U-235 assuming that Pu has about two thirds of fissile isotopes [2]. When uranium prices were low the reprocessing to separate Pu for recycling at MOX was not itself economic. But with the rise of natural uranium prices coupled with reducing the spent fuel volume is becoming so [2] [3].
2. MODEL SPECIFICATIONS OF THE CALCULATIONS In this paper are presented six calculated models of the MOX assembly (fig 1-6). The first three models (fig 1-3) typify MOX assembly in the case of a mixed core. The mixed core is loaded by 30 % of the recycled Pu. The MOX assembly of three different rector types (PWR 900 MWe, WWER-1000 and WWER-440) on the figures 1, 2 and 3 is surrounded by UOX assemblies. The next three cases (fig 4-6) represent the MOX assembly of PWR 900, WWER-1000 and WWER-440 reactor in the infinite lattice. The standard MOX assembly used in this study includes three zones with different Pu contents to flatten the within assembly power distribution and to attenuate fission rate discontinuities at the MOX-UOX interface. The central zone is characterised by a high Pu content (5.5 %) and a peripheral zone by a low Pu content (2.85 %). The MOX PWR assembly profiling was used from the depletion calculation benchmark devoted to MOX fuel cycles in an ongoing objective of OECD/NEA WPRS (Working Party on Scientific Issues of Reactor Systems) [1]. This benchmark is directly connected with the French Post Irradiation Examination (PIE) program [4]. The profiling of MOX WWER-1000 and WWER-440 assembly was adopted from the MOX PWR assembly and the Pu content in each zone is the same as in the PWR MOX fuel assembly. All kinds of calculations were performed by the spectral computer code HELIOS 1.10. 2.1 SUPERCELL MODELS The requested calculations were performed to attain a constant target burn up of 42 GWd/tHM for all of the studied MOX fuel assemblies. The following figures (fig.1, 2 and 3) of MOX assembly surrounded by UOX assemblies - super cells is shown. The first super cell (fig. 1) consists of eight UOX PWR assemblies and the MOX PWR fuel assembly is situated in the centre of the super cell. The fig. 2 shows the super cell of WWER-1000. Six WWER-1000 UOX assemblies surround the central WWER-1000 MOX assembly. A design of the fig. 3 is very similar to fig.2. The MOX WWER-440 fuel assembly in the centre with six UOX WWER-440 fuel assemblies is demonstrated. Each one of all the super cell models is calculated by periodic boundary conditions. The three following operating periods of fuel cycle represent the MOX fuel irradiation history [1]: 1 st period: 285 full power days, burn up = 12 GWd/tHM Down time: 60 days 2 nd period: 300 full power days, burn up = 25 GWd/tHM Down time: 40 days 3 rd period: 280 full power days, burn up = 42 GWd/tHM
MOX PWR Assembly UOX PWR Assembly Fig. 1: MOX-UOX PWR geometry. Pu/(U+Pu+Am) = 4.42 % Burn up of UOX assembly: (24 GWd/tHM + 34 GWd/tHM) final 58 GWd/tHM Burn up of MOX assembly: app. 42 GWd/tHM Red - high (5.516 % Pu) Blue - medium (4.334 % Pu) Pink - low (2.856 % Pu) MOX WWER-1000 Assembly UOX WWER-1000 Assembly Fig. 2: Position of the MOX WWER-1000 in WWER-1000 supercell. Pu/(U+Pu+Am) = 4.42 % Burn up of UOX assembly: (24 GWd/tHM + 31 GWd/tHM) final 55 GWd/tHM Burn up of MOX assembly: app. 42 GWd/tHM Red - high (5.516 % Pu) Blue - medium (4.334 % Pu) Pink - low (2.856 % Pu)
MOX WWER-440 Assembly UOX WWER-440 Assembly Pu/(U+Pu+Am) = 4.42 % Geometry: Gd-2 assembly Burn up of UOX assembly: (24 GWd/tHM + 31 GWd/tHM) final 55 GWd/tHM Burn up of MOX assembly: app. 42 GWd/tHM Red - high (5.516 % Pu) Blue - medium (4.334 % Pu) Pink - low (2.856 % Pu) Fig. 3: MOX - UOX WWER-440 geometry. 2.2 MOX FUEL ASSEMBLIES, INFINITE LATTICE The profiled MOX fuel assemblies of the PWR 900, WWER-1000 and WWER-440 are shown on fig. 4, 5 and 6. Boundary conditions were set in HELIOS 1.10 for infinite lattice calculation for all of the studied assemblies. The final burn up was proposed to the same value as in super cell - 42 GWd/tHM. Fig. 4: MOX PWR assembly. Fig. 5: MOX WWER-1000 assembly. Fig. 6: MOX WWER-440 assembly.
2.3 MOX FUEL COMPOSITIONS The listed initial MOX composition is a representative of the realistic UOX fuel irradiated in a mixed MOX - UOX LWR core [1]. This MOX fuel consists of a typical plutonium vector for material derivated from reprocessing of thermal reactor UOX fuel. Plutonium isotopic composition, corresponding to the three Pu content zones, is given in Tab. 1 and uranium isotopic composition in Tab. 2. Tab.1: Plutonium isotopic composition in fresh MOX fuel [1]. Nuclide Isotopic Composition [Atom.%] Pu-238 0.8 Pu-239 66.7 Pu-240 20.6 Pu-241 7.5 Pu-242 2.9 Am-241 1.5 Tab.2: Uranium isotopic composition in fresh MOX fuel [1]. Nuclide Isotopic Composition [Atom.%] U-234 0.002 U-235 0.22 U-236 0.004 U-238 99.77 Tab.3: Initial MOX fuel content [1]. Pu content in the zone Plutonium Content, w/o Pu(total) + Am/[U+Pu+Am] High 5.6 Medium 4.4 Low 2.9 Plutonium from reprocessed fuel is usually fabricated into the MOX fuel as soon as possible to avoid problems with the decay of short-lived Pu isotopes [2]. In particular, Pu-241 (halflive 14.29 years [5]) decays to Am-241 which is a strong gamma emitter, giving rise to a potential occupational health hazard. It is so if separated Pu, over five years old, is used in a normal MOX plant. The Am-241 level in stored Pu increases about 0.5 % per year with corresponding decrease in fissile value of the Pu [2]. Content of Pu-238 (half-live 87.7 years [5]) is increased in high-burn up fuel. It is a strong alpha emitter and a source of spontaneous neutrons. Pu-239, Pu-240 and Pu-242 are long-lived and hence little changed with prolonged storage [2]. 2.4 UOX FUEL COMPOSITIONS For the calculations, the adjacent UOX fuel assemblies are modeled as already burned. They have an initial enrichment of 3.25 w/o and have reached a burn up of 24 GWd/tHM [1].
3. RESULTS AND CONCLUSIONS The Pu weight concentrations at MOX PWR assembly as a super cell part and as a part of infinite lattice are listed in tab. 4. The weight concentration final values in Pu high and Pu medium zones are similar but the values in the Pu low zone are different. These differences are caused by the fission rate discontinuities at the MOX-UOX interface. The Pu development in the zones of MOX PWR assembly as a part of structure depending on burn up is shown on the fig. 7. In all of the zones the Pu content decreases with burn up value. The comparison of average weight concentrations for MOX PWR, WWER-1000 and WWER-400 assembly in the infinite lattice and as a part of the super cell are presented in tab.5. The Pu-239, Pu-242, Pu and MA development of MOX PWR, WWER-1000 and WWER-440 assembly in the infinite lattice is presented on the last figure (fig. 8). The Pu content decreases in all of the Pu cases. MA dependance on the burn up slowly rises. On the base of results it is possible to see a Pu reduction in all of calculated cases. The reductions between initial and final burn up values for MOX assembly as a part of super cell: Pu-239 (PWR): -63 % Pu-239 (WWER-1000): -62 % Pu-239 (WWER-440): -54 % Pu (PWR): -33 % Pu (WWER-1000): -32 % Pu (WWER-440): -27 % Spent LWR UOX fuel containes of approximately 1 % of Pu. Then 4.5 LWR UOX assemblies are neccesary to reprocess for production of 1 MOX assembly with 4.42 % Pu content. If it is possible to spent 60 % of Pu-239 in MOX fuel then this Pu-239 depletion leads to Pu-239 spending from 2.7 LWR UOX assemblies. REFERENCES [1] ROQUE, B. MARIMBEAU, P. GROUILLER, J.P.: Specification for the Phase 2 of a Depletion Calculation Benchmark devoted to MOX Fuel Cycles. WPRS NEA/OECD. December 2007. [2] World Nuclear Association, Mixed Oxide (MOX) Fuel, http://www.worldnuclear.org/info/inf29.html, online, 2008. [3] SEBIAN, V.: Transmutation of radionuclides of spent fuel in WWER-440 reactors, (PhD thesis). Slovak University of Technology, March 2008. [4] ROQUE, B. MARIMBEAU, P. BIOUX, P.:The French Post Irradiation Examination Database for the validation of depletion calculation tools, Post Irradiation Examination (PIE) program, 2003. [5] TULI, K. J.: Nuclear Wallet Cards, National Nuclear Data Center, April 2005. [6] BAILLY, H. MENESSIER, D. PRUNIER, C.: The Nuclear Fuel of Pressurized Water Reactors and Fast Reactors. Lavoisier Publishing, 1999. [7] Manual: Helios 1.10, 2008. [8] Darilek, P.,Zajac,R.,Breza,J.,Necas,V.:Comparison of PWR-IMF and FR fuel cycles. In: 2007 GLOBAL 2007: Advanced Nuclear Fuel Cycles and Systems,pp.667-670
Tab. 4: Weight concenrations [%] in the zones of MOX PWR assembly as super cell part. PWR - 900 Nuclide Initial [%], MOX Assembly in Super Cell, 42 GWd/tHM MOX Assembly in Infinite lattice, 42 GWd/tHM 0 GWd/tHM High [%] Medium [%] Low [%] Average [%] High [%] Medium [%] Low [%] Average [%] U-235 0.21 0.09 0.08 0.07 0.08 0.09 0.08 0.08 0.08 Np-237 --- 0.006 0.006 0.006 0.005 0.006 0.006 0.006 0.006 Np-239 --- 0.01 0.01 0.01 0.01 0.009 0.009 0.009 0.009 Pu-238 0.04 0.07 0.06 0.04 0.06 0.08 0.06 0.04 0.06 Pu-239 3.00 1.42 1.07 0.87 1.15 1.42 1.11 0.96 1.19 Pu-240 0.92 1.18 0.91 0.63 0.94 1.17 0.91 0.64 0.94 Pu-241 0.34 0.68 0.54 0.39 0.56 0.68 0.54 0.41 0.56 Pu-242 0.13 0.35 0.31 0.24 0.31 0.35 0.31 0.23 0.30 Am-241 0.07 0.05 0.03 0.02 0.04 0.05 0.04 0.03 0.04 Pu Total 4.42 3.69 2.89 2.16 3.02 3.69 2.93 2.27 3.06 MA Total 0.07 0.24 0.21 0.16 0.21 0.25 0.21 0.16 0.21 Tab. 5: Average weight concentrations [%] in MOX assembly of PWR-900, WWER-1000 and WWER-440 reactor. Reactor Nuclide Initial [%], PWR-900, 42 GWd/tHM WWER-1000, 42 GWd/tHM WWER-440, 42 GWd/tHM 0 GWd/tHM Super Cell [%] Infinite L. [%] Super Cell [%] Infinite L. [%] Super Cell [%] Infinite L. [%] U-235 0.21 0.08 0.08 0.08 0.08 0.09 0.08 Np-237 --- 0.005 0.006 0.006 0.006 0.006 0.006 Np-239 --- 0.01 0.009 0.01 0.009 0.01 0.009 Pu-238 0.04 0.06 0.06 0.06 0.06 0.06 0.07 Pu-239 3.00 1.15 1.19 1.22 1.26 1.39 1.45 Pu-240 0.92 0.94 0.94 0.94 0.94 0.94 0.93 Pu-241 0.34 0.56 0.56 0.57 0.57 0.58 0.61 Pu-242 0.13 0.31 0.30 0.30 0.30 0.27 0.29 Am-241 0.07 0.04 0.04 0.04 0.04 0.04 0.05 Pu Total 4.42 3.02 3.06 3.08 3.13 3.24 3.34 MA Total 0.07 0.21 0.21 0.21 0.22 0.22 0.23
6.0 5.0 Pu-239 (HIGH) (Np+Am+Cm+Bk+Cf) HIGH (Np+Am+Cm+Bk+Cf) MEDIUM Pu(MOX PWR) LOW Pu-239 (MEDIUM) Pu(MOX PWR) HIGH Pu(MOX PWR) MEDIUM Pu-239 (LOW) (Np+Am+Cm+Bk+Cf) LOW 4.0 [%] 3.0 2.0 1.0 0.0 0 5 10 15 20 25 30 35 GWd/tHM (super cell) Fig. 7: Development of Pu in the zones of MOX PWR assembly as a super cell part. 5.0 4.0 Pu-239 (PWR) Pu(MOX PWR) Pu-239 (WWER1000) Pu(MOX WWER1000) Pu-239 (WWER440) Pu(MOX WWER440) Pu-242 (PWR) (Np+Am+Cm+Bk+Cf) PWR Pu-242 (WWER1000) (Np+Am+Cm+Bk+Cf) WWER1000 Pu-242 (WWER440) (Np+Am+Cm+Bk+Cf) WWER440 3.0 [%] 2.0 1.0 0.0 0 10 20 30 40 GWd/tHM (assembly) Fig. 8: Development of Pu in the MOX PWR, WWER-1000 and WWER-440 assembly in the infinite lattice.