Experimental Validation of Ex-Vessel Neutron Spectrum by Means of Dosimeter Materials Activation Method

Abstract

Neutron spectrum information in reactor core and around of ex-vessel reactor needs to be known with a certain degree of accuracy to support the development of fuels, materials, and other components. The most common method to determine neutron spectra is by utilizing the radioactivation of dosimeter materials. This report presents the evaluation of neutron flux incident on M3dosimeter sets which were irradiated outside the reactor vessel,as well as the validation of neutron spectrum calculation. Al capsules containing both dosimeter set covered withCd and dosimeter set without Cd cover have been irradiated during the 35th operational cycle in the M3 ex-vessel irradiation hole position207 cmfrom core centerline at the space between the reactor vessel and the safety vessel. The capsules were positioned at Z=0.0 cm of core midplane. Each dosimeter set consists of Co-Al, Sc, Fe, Np, Nb, Ni, B, and Ta. The gamma-ray spectra of irradiated dosimeter materials were measured by 63 cc HPGe solid-state detector and photo-peak spectra were analyzed using BOB75 code. The reaction rates of each dosimeter materials and its uncertainty were analyzed based on 59Co (n,g) 60Co, 237Np (n,f) 95Zr-103Ru, 45Sc (n,g) 46Sc, 58Fe (n,g) 59Fe, 181Ta (n,g) 182Ta, and 58Ni (n,p)58Co reactions. The measured Cd ratios indicate that neutron spectrum at the irradiated dosimeter sets was dominated by low energy neutron. The experimental result shows that the calculated neutron spectra by DORT code at the ex-vessel positions need correction, especially in the fast neutron energy region, so as to obtain reasonable unfolding result consistent with the reaction rate measurement without any exception. Using biased DORT initial spectrum, the neutron spectrum and its integral quantity were unfolded by NEUPAC code. The result shows that total neutron flux, flux above 1.0 MeV, flux above 0.1 MeV, and the displacement rate of the dosimeter set not covered with Cd were 1.75× 1012 n cm2 s-1, 1.83× 108 n cm2 s-1, 2.94× 1010 n cm2 s-1, and 2.39× 10-11 dpa s-1, respectively. The uncertainty of neutron flux by NEUPAC was mainly due to the error of the initial spectrum.

Received: 10 December 2015; Revised: 14 July 2016; Accepted: 25 September 2016

Keywords

Irradiation; Dosimeter material; Gamma spectrometry; Theory transport; Unfolding method