Sensitivity Analysis of Neutron Cross Section for Graphite and Lead by Using MCNP Code and Experimental Data

Parvazian, A; Sadeghi, M

Sensitivity Analysis of Neutron Cross Section for Graphite and Lead by Using MCNP Code and Experimental Data

Document Type : Research Paper

Authors

A Parvazian

M Sadeghi

Abstract

Because of the importance of safety in a nuclear system, nuclear data which are used in neutronic reactors design must be in a good accuracy. So examination and enhancement of nuclear data are of great importance. The aim of this work is to investigate the neutron cross section with carbon and lead and consequently improving them. In a research program conducted at the Isfahan Production and Research Center, blocks of lead and graphite were used. In the core of lead and carbon blocks, we placed an Am-Be source and then measured the flux of neutron outside of the assembly. The flux distribution was also calculated theoretically by MCNP code and the results were compared with the experiment. Comparison of the computation and the experimental results showed that the carbon sensitivity coefficients in the range of energies lower than 1MeV are negligible, while for the higher energies, especially between 5 to 6 MeV, these coefficients are nearly 5 percent. Also, lead sensitivity coefficients was varied from 0.2 to 0.67 for different energies. Finally, with the help of defining sensitivity coefficients, the values of cross section were varied. By using more advanced equipment for fast neutron detection, we are able to find better results in the other range of energy groups.

Highlights

1. C.E. Burgart, “Capabilities of morse multigroup monte carlo code in solving reactor eigenvalue problems,” December (1971).

2. P.M. Song, M.Z. Youssef, M.A. Abdou, “A new approach and computational algorithm for sensitivity/ uncertainty analysis for SED and SAD with application to beryllium integral experiment,” Nucl. Sci. Eng. 113-339 (1993).

3. H.B. Choi, T.J. Downar, “Sensitivity theory for the closed nuclear fuel cycle,” Nucl. Sci, Eng. 111-205 (1992).

4. M.Z. Yossef, W. Conn, C.W. Maynard, “Impact of cross-section uncertainties on the nuclear design,” Nucl. Technol. Fusion, 2 (1982).

5. C.R. Weisbin, C.M. Oblow, J.H. Marable, R.W. Peele, J.L. Lucius, “Application of sensitivity and uncertainty methodology to fast reactor integral experiment analysis,” Nuc. Sci. Eng. 66-307 (1978).

6. D.G Cacusi, “Global optimization and sensitivity analysis,” Nucl. Sci. Eng. 104-78 (1990).

7. “MCNP4C monte carlo N-Particle transport code system,” Los Alamos National Laboratory, April (2000).

8. M.E. Dunn, P.B. Fox, N.M. Greene, L.M. Petrie, “ENDF/B-VI library generation and testing for the scale code system,” Oak Ridge Tennessee,USA (2005).

Keywords

Neutron Sources

Cross Sections

Neutron Flux

Nuclear Data Collections

Sensitivity Analysis

Monte Carlo Method

MeV Range

1. C.E. Burgart, “Capabilities of morse multigroup monte carlo code in solving reactor eigenvalue problems,” December (1971).

2. P.M. Song, M.Z. Youssef, M.A. Abdou, “A new approach and computational algorithm for sensitivity/ uncertainty analysis for SED and SAD with application to beryllium integral experiment,” Nucl. Sci. Eng. 113-339 (1993).

3. H.B. Choi, T.J. Downar, “Sensitivity theory for the closed nuclear fuel cycle,” Nucl. Sci, Eng. 111-205 (1992).

4. M.Z. Yossef, W. Conn, C.W. Maynard, “Impact of cross-section uncertainties on the nuclear design,” Nucl. Technol. Fusion, 2 (1982).

5. C.R. Weisbin, C.M. Oblow, J.H. Marable, R.W. Peele, J.L. Lucius, “Application of sensitivity and uncertainty methodology to fast reactor integral experiment analysis,” Nuc. Sci. Eng. 66-307 (1978).

6. D.G Cacusi, “Global optimization and sensitivity analysis,” Nucl. Sci. Eng. 104-78 (1990).

7. “MCNP4C monte carlo N-Particle transport code system,” Los Alamos National Laboratory, April (2000).

8. M.E. Dunn, P.B. Fox, N.M. Greene, L.M. Petrie, “ENDF/B-VI library generation and testing for the scale code system,” Oak Ridge Tennessee,USA (2005).