Effect of WIMSD4 libraries on Bushehr VVER-1000 Core Fuel Burn-up

Transcription

1 International Conference Nuccllearr Enerrgy fforr New Eurrope 2009 Bled / Slovenia / September ABSTRACT Effect of WIMSD4 libraries on Bushehr VVER-1000 Core Fuel Burn-up Ali Pazirandeh and Elham Younesian* Science and Research Branch, Islamic Azad University, Tehran Iran e_y1982@yahoo.com, Pzrud193y@srbiau.ac.ir The objective of this paper is to present the comparative results of fuel burn-up in VVER core using IAEA, WLIB and WIMKALL88 evaluated data files. The calculations performed using WIMSD4 and CITATION codes, for different type of assemblies, based on the generated 69-group cross-section library. In this study two actinides, Pu239 and Pu240, and two fission products, Xe135 and Sm149, were compared. There are significant differences using different libraries, also for different type of assemblies, further also neutron spectra at BOC and EOC are quite different. Keyword: WIMS Library, fuel burn-up, core parameters, evaluated data files. 1 INTRODUCTION The WIMSD4 [ 1,2]code is widely used for reactor calculations of for a variety of thermal reactors. It consists of a lattice transport code and the associated library. One of the most commonly uses is to generate a few group cross-section library in problem dependent form to interface with neutronic codes. The original WIMSD cross-section library being very old, by generating WIMSD library based on latest release of the evaluated data files, such as ENDF/B-VI, JENDL-3.2, and JEF-2.2 are now available from www-nds.iaea.org [3]. Because, in most developing countries WIMSD4 is used for reactor core physics calculations [4], we performed the calculations using WIMSD4. The code is now being used for calculation of Bushehr Nuclear Power Plant (BNPP) core parameters. These libraries were developed by processing data from different newly released basic evaluated cross-section data files. After release of different WIMSD libraries by IAEA based on recent cross-section data files, it is necessary to study the effect of new cross-sections on neutron spectra, core parameters in the fuel rods and the fuel burn-up of VVER-1000 core. 2 COMPUTATIONAL MODEL In this study, the calculations for cluster model performed to obtain parameters for 1.6%, 2.4% and 3.6% enriched uranium using WIMSD4 code, for two group neutron energy and IAEA, WLIB and WIMKALL88 data libraries. To model burnable poison (CrB2+Al) [5] effect on parameter values, three assemblies with 1.6%, 2.4% and 3.6% enriched uranium 116.1

2 116.2 respectively were chosen and the calculations performed as mentioned above. Since WIMSD4 input is defined in cylindrical geometry and VVER-1000 assemblies are hexagonal shape, and the assemblies approximated cylindrical shape using Wigner-Seitz model [6]: p 2n 3 r = 2 π where r and p are the cylinder radius and hexagonal VVER cell pitch, respectively. Fig. 1 Wigner-Seitz model of a hexagonal to cylinder for the WIMS-D4 cluster input [6]. In WIMSD4 input the ARRAYS radii of the equivalent n hexagonal cell assumed to be cylindrical as shown in Fig.1. The core parameters of some assemblies were calculated after 100 days at full operation at 3000MWt, using POWERC option of WIMS-D4 to obtain the products concentration and the depleted isotopes concentrations. The new isotopic concentrations used in WIMSD4 to evaluate new core parameters. The two group fluxes in each assembly were also computed using CITATION code. A triangular section of the VVER-1000 core, 1/6 th, with mirror image was used as a model for the flux and effective multiplication factor calculation. The computational assembly geometry model is shown in Fig.2. Reflector Empty Cannel Reflector Fuel Rod Burnable poison Fig.2. Assembly computational model and its reflector region

3 116.3 Table 1 Deviation of calculated parameters from the averaged values for different fuel enrichment; 1.6%, 2.4% and 3.6%.* Fuel Averaged Deviation% Dev.% Dev.% Dev.% Parameter ENDF/BVI JEF2.2a JENDL3.2 WLIB 1.6% D1= D2= E Σ 1a = E Σ 2a = E Σ 1R = E E νσ 1f = 4.733E νσ 2f = E k eff = E % D 1 = D 2 = E Σ 1a = E Σ 2a = E Σ 1R = E νσ 1f = E νσ 2f= E k eff = % D 1 = D 2 = E Σ 1a = E Σ 2a = E Σ 1R = E νσ 1f = E νσ 2f = E k eff = ENDF/BVI JEF2.2a JENDL3.2 WLIB * Please note: The last row does not follow exactly the top heading of each column. The eigenvalues were obtained from the CITATION outputs. 3 RESULTS AND CONCLUSIONS The core parameters, D g, Σ ag, Σ rg, νσ fg, k eff, were calculated for different type of assemblies and along with three libraries to be used for fuel burn-up calculations. The parameters are weighted parameters for group g as D g, diffusion coefficient, Σ ag, absorption cross section, Σ rg removal cross section, ν number of neutrons per fission, Σ fg fission cross section and k eff multiplication factor. The calculations performed for pin-cell and cluster model with fuel enrichment 1.6%, 2.4% and 3.6% respectively. Two group neutron fluxes and multiplication factor were obtained from CITATION-LDI2[7] code, using above core parameters for each library. In burn-up calculations variety of actinides and fission products are produced. In this work, only two actinides, Pu-239 and Pu-240 and two fission products, Xe-135 and Sm-149 were compared. There are significant differences using different libraries and different type

4 116.4 of assemblies. As Fig.(3) shows the production of Pu239 and Pu240 for FA with 2.4% U235 during first cycle increases for three data libraries. It is seen, for two libraries IAEA, WLIB results are the same, and for WIMKAL88 library, the increase is lower. This difference is due to greater capture cross sections in first two libraries. By looking at Fig.(4) and Figs(6,8,10) because of short half-lives of Xe135, 9.10h, and its predecessor I h and its high capture cross section ( b), it saturates after 7 days. In case of Sm149, capture cross section is much lower ( b) and it is a stable isotope. It is seem that enrichment and burnable poison do not have much effect on trend of the Xe135 and Sm149 variations. This indicates that the reactivity effects of these fission products during the first cycle have remained nearly constant. The neutron spectra for different FA types using different libraries were also calculated. Two spectra at the BOC and EOC are shown in Figs(11 to 14). The neutron spectra in core components; fuel, cladding, and moderator, burnable poison and structural materials are used as a weighting function to calculate effective cross sections of different reaction rates in each energy group of multi-energy group problems. To simplify the comparison, the spectra are divided into three groups, 1-fast, 2-resonance and 3- thermal. As Fig.11 shows in group 1, spectra have the same shape. In group 2, the spectrum of ENDFB6 is much lower than the other two. This means that the reactivity feedback due to Doppler Effect is much lower than the other two. In thermal group WIMKAL88 is lower, which means lower absorption. At EOC (300 days) calculated spectrum of ENDFB6 is much lower than the other two in groups 1 and 2. As Figs(14-15) show 20mg/cm 3 burnable poison has little effect on spectra as compared with BOC spectra. Since neutron spectrum varies with burn-up and power distribution in the reactor core, therefore different libraries gave different values for core parameters as shown in Table 1. In the table and figures, two important points capture reader s attention: deviations of constants from averaged values, which in some cases are rather large, (see Table 1). Second rather astonishing differences are the neutron spectra of different libraries. Thermal neutron fluxes at BOC and EOC for two libraries IAEA and WLIB are shown in Figs( ). The difference at BOC and EOC is mainly due to fuel burn-up, plutonium production and degradation of burnable poison, which leads to flux flattening.

5 116.5 Figs.3-4 Heavy metal isotopes and fission products for FA 1.6% U235 Fig.5-6 Heavy metal isotopes and fission products for FA 2.4% U235 with 20mg/cm 3 burnable poison

6 116.6 Fig.7-8 Heavy metal isotopes and fission products for FA 3.6% U235. Figs.9-10 Heavy metal isotopes and fission products for FA 3.6% U235, with 36mg/ cm 3 burnable poison.

7 116.7 Normalized Flux,n/cm 2.s ENDFB6 WIMKAL88 WLIB Neutron Flux,n/cm IAEA 24 WLIB 24 WKAL 0 5.E-02 3.E-01 1.E+00 2.E+00 4.E+02 7.E+04 1.E E-02 1.E-01 8.E-01 1.E+00 8.E+01 4.E+04 1.E+07 Neutron Energy,eV Neutron Energy,eV Fig.11. Neutron spectra for FA 2.4% U 235 at BOC Fig.12. Neutron spectra for FA 2.4% U 235 at U 235 at EOC 25 IAEA 25 24b20 IAEA Normalized Flux,n/cm 2.s WMKAL WLIB Neutron Flux,n/cm b20 WLIB 24b20 WKAL 0 5.E-02 3.E-01 1.E+00 2.E+00 4.E+02 7.E+04 1.E E-02 1.E-01 8.E-01 1.E+00 8.E+01 4.E+04 1.E+07 Neutron Energy,eV Neutron Energy,eV Fig.13. Neutron spectra for FA 2.4% U 235. Fig 14. Neutron spectra for FA2.4% U 235 with 20 mg/cm 3 burnable poison at BOC.. with 20 mg/ cm 3 burnable poison at EOC

8 116.8 Fig.15 Neutron Flux for Bushehr Nuclear Power Plant at BOC with WLIB library Fig.16. Neutron Flux for Bushehr Nuclear Power Plant at EOC with WLIB library Fig.17. Neutron Flux for Bushehr Nuclear Power Plant at BOC with IAEA library Fig.18. Neutron Flux for Bushehr Nuclear Power Plant at EOC with IAEA library

9 116.9 REFERENCES [1] Taubman C.J. The WIMS 69 group library tape , AEEW-M1324, Atomic Energy Establishment Winfrith, Dorchester,1975 [2] Joubert W.R., Review report for WIMSD44.1, RTE01-2/2-046, Atomic Energy Corporation of South Africa, 13 June [3]. Leszczynski F., Aldama D. López, Trkov A.;, International Atomic Energy Agency WIMS- D library update: Final report of a Coordinated Research Project ; Vienna: International Atomic Energy Agency, 2007.STI/PUB/1264; ISBN [4]. Ahmad Siraj-ul-Islam, Ahmad Nasir; Effect of updated WIMSD libraries on neutron energy spectrum at irradiation site of Pakistan Research Reactor-1 using 3D modeling ; Annals of Nuclear Energy 32(2005) [5] Final Safety Analysis Report, Chapter 4, Reactor, NPP Bushehr Unit 1, # , 10 October 2003 [6]. Hadad Kamal, Ayobian Navid, Piroozmand Ahmad; Quantitative accuracy analysis of burnup calculations for BNPP fuel assemblies using FFTBM method, Progress in Nuclear Energy 51 (2009) [7] Fowler T.B., Vondy D.R., Cunningham G.W., Nuclear reactor core analysis code : CITATION2, ORNL, July 1971