Prediction of neutron energy spectrum in a typical MTR type research reactor using Monte Carlo simulations

Malkawi, Salaheddin; Khuwaileh, Bassam A.; Al-Momani, Malik

doi:10.1016/j.anucene.2012.12.025

Cited by 5 publications

(3 citation statements)

References 12 publications

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“…The gamma ray spectra of all the irradiated Co dosimeter show sharp photopeaks of 1173. 24 Ce (T 1/2 = 284.89 d) did not appear clearly in the gamma ray spectra since the dosimeter set was irradiated for only a short period.…”

Section: Gamma Ray Spectrummentioning

confidence: 99%

“…Precise knowledge of neutron energy spectrum is imperative to various fundamental and experimental studies in many field of nuclear physics, nuclear engineering and technology, medical sciences, and health physics research [8][9][10][11][12][13][14][15][16]. An accurate description of neutron spectrum is needed for understanding a reactor's characteristics and evaluating the quantity of neutron irradiation effect on reactor components such as fuel and materials [1,[17][18][19][20].…”

mentioning

confidence: 99%

“…The dosimeter material activation technique is a basic method for determination of neutron doses internal to the reactor core and around reactor vessel [9,21,22], especially for the determination of neutron flux, the characteristics of neutrons in the core, and the characterizations of fuel, materials, and other components, as well as for reactor safety study [2,[23][24][25]. The accuracy enhancement of neutron spectrum measurement is still a major issue in the reactor dosimetry research, in order to improve the accuracy of determination of the neutron spectrum, especially in reflector and reactor vessel area, and to verify the results of reactor surveilance program [26][27][28][29][30][31].…”

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confidence: 99%

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

Santa

2017

Atom Indo.

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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 M3 dosimeter sets which were irradiated outside the reactor vessel, as well as the validation of neutron spectrum calculation. Al capsules containing both dosimeter set covered with Cd and dosimeter set without Cd cover have been irradiated during the 35 th operational cycle in the M3 ex-vessel irradiation hole position 207 cm from 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 59 Co (n,) 60 Ta, and 58 Ni (n,p) 58 Co 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

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Section: Gamma Ray Spectrummentioning

confidence: 99%