Experimental and simulation study of neutron dosimetry at various neutron energies

Radiation Measurements

Volumn 28, Issue 1-6, 1997, Pages 495-498

  Jamil, K ^c   Ali, S ^c   Qureshi, I E ^c   Rehman, F ^c   Khan, H A ^c   Manzoor, S ^c   Waheed, A ^a,b   Cherubini, R ^a  

^a LABORATORI NAZIONALI DI LEGNARO   (Italy)

^b GOMAL UNIVERSITY   (Pakistan)

^c PAKISTAN INSTITUTE OF NUCLEAR SCIENCE AND TECHNOLOGY   (Pakistan)

Author keywords

CR 39; Detection efficiency; MCNP; Monte Carlo simulation; Neutron dosimetry; SSNTD; Track detector

Indexed keywords

CARBON; COMPUTER SIMULATION; MONTE CARLO METHODS; NEUTRON DETECTORS; OXYGEN; PARTICLE DETECTORS; PROTONS;

MONTE CARLO NEUTRON PHOTON (MCNP) TRANSPORT CODE; SOLID STATE NUCLEAR TRACK DETECTORS (SSNTD);

DOSIMETERS;

EID: 0031367286     PISSN: 13504487     EISSN: None     Source Type: Journal    
DOI: 10.1016/S1350-4487(97)00127-3     Document Type: Article

References (7)

1. Ing H. and Piesch E. (1985) Status of Neutron Dosimetry. Radiat. Prot. Dos. 10, 1-4, 5-15.
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3. Matiullah and Durrani S. A. (1987) Chemical and Electrochemical Registration of Protons in CR-39-Implication for Neutron Dosimetry. Nucl. Instru. Meth. B29, 508-514.
4. McLane V., Dunford C. L. and Rose P.F.(1988) Neutron Cross Sections, Volume 2. Academic Press, Inc, New York.
5. Piesch E. (1982) Albedo Neutron Dosimetry. Int. J. Radiat. Isot. 33, 1061-1076
6. Pretzsch G. (1982) Theoretical Determination of the Neutron Detection Efficiency of Plastic Track Detectors. Nucl. Instru. Meth. 195, 543-550.
7. Thomson W. L. (1990) MCNP-A General Monte Carlo Code for Neutron and Photon Transport, Revised Edition. Los Alamos Monte Carlo Group, Los Alamos National Laboratory, New Mexico, U.S.A.