1 Copyright © 20xx by ASME Proceedings of the 2014 22nd International Conference on Nuclear Engineering ICONE22 July 7-11, 2014, Prague, Czech Republic ICONE22-30057 233 U EVALUATION COMPARISON STUDY Mohammad Alrwashdeh Tsinghua University, Department of Engineering Physics Beijing, China Aniseh Abdalla Tsinghua University, Department of Engineering Physics Beijing, China Kan Wang Tsinghua University, Department of Engineering Physics Beijing, China ABSTRACT The aim of this study is to investigate the availability and accuracy of the cross section data for 233 U to perform the calculations of the critical system. Two evaluated data libraries are available, U.S. data bank (ENDF) and the Japanese data bank (JENDL), by using BAYES method for resonance parameters available in SAMMY code and weighted least square method with nonlinear regression by using FITWR computer code. Evaluation of the 233 U has been investigated by using of SAMMY code, in order to generate a useful data base for critical calculations, the computer code FITWR for experimental - experimental data fitting show same results obtained from Bayes method included within SAMMY code, with a slight deference in the results at the evaluated cross sections due to different mathematical methods have different results INTRODUCTION The FITWR computer program fits the experimental nuclear data extracted from EXFOR data base, in the absence of the theoretical background data to evaluate the nuclear cross sections for a specific energy range of an incident particle for a certain nuclei, were the fitting procedure is also adapted when there is a need to evaluate the model parameters in order to achieve the completeness of the physical model. The 233 U evaluations in the ENDF/B-IV and JENDL-4.0 are almost identical, where ENDF/B-IV data for 233 U are based on 1978 evaluation which employs the Alder-Alder approximation (5) . Most of the evaluations for 233 U in the resolved energy range were done a few years ago (6) by using the reduced R-matrix Reich-Moore formalism (7) in the computer code SAMMY (1) . This formalism is appropriate for fissile isotopes like 233 U. The evaluations were extended up to 150 eV for better representation of cross section energy self- shielding effects, these kinds of evaluations were adapted in JENDL-4.0. Table 1. Summarizes overall status of ENDF/B-IV and JENDL-4.0 evaluations. Table 1- 233 U ENDF and JENDL evaluation features (10) Library Feature ENDF JENDL Energy Range 0.79 eV – 60 eV 10 -5 eV - 150 eV Formalism Reich - Moore Reich - Moore Background cross section Yes (not suitable for temperature effect) No ( no problem for temperature effects) ANALYSIS An important quantity called Maxwellian average cross section at thermal energies were used to interpretation of the thermal benchmarks, by using the calculation of the Wescott gw factors and K1 parameters, defined as following: 1/ 2 1 0 2 ( ) w x x g (1) Where gw stands for fission (gf), absorption (ga). And 0 1 0 g g f f a a k (2) Where σ a and σ 0a are the Maxwellian average cross sections at 0.0253 eV respectively. Fission and capture resonance integrals, and the ratio between them (α= Ic/If) are important parameters in the evaluation procedure. Table 2 shows the value of this quantities. Where the resonance integral defined as: 20 1 0.5 MeV x eV I x dx (3)