International Conference on the Management of Spent Fuel from Nuclear Power Reactors 2019: Learning from the Past, Enabling the Future

Core Modeling and Source Term Calculations Using MCNPX Code for Fukushima Daiichi Unit-1 Nuclear Power Plant Accident

26 Jun 2019, 10:30

30m

Poster Track 4: Recycling as a spent fuel management option Track 4 Poster Session

Speaker

Prof. MOHAMED SHAAT (Consultant for Nuclear Affairs, Egyptian Atomic Energy Authority )

Description

Management of severely damaged spent fuel and corium plays an important role in the nuclear safety issues for the nuclear power plants. The calculation of burned fuel inventory is required for determining the composition, activity of core melt and in the estimation of the radiological source term in the environment. Isotopic inventory of the burned fuel at the time of the accident of Fukushima Daiichi Unit 1 (FD-U1) was calculated using Monte Carlo analysis MCNPX 2.7 code linked to depletion calculation code CINDER'90 and ENDF/B-VII.0 cross section data library. The reactor core model results were validated with experimental measurements which was carried out by Japan Nuclear Energy Safety Organization (JNES) and verified with published results using ORIGEN-Code by Japan Atomic Energy Agency (JAEA). The verification comparison was in good agreement for all the radionuclides, and more radionuclides were obtained using MCNPX-Code. The total activity of the burned fuel at the time of the accident was 9.86E+19 Bq and after 50 Yrs. was 1.89E+17 Bq and the higher inventory concentration in the fuel was dominated by the trans-uranic elements. Also, the specific activity was calculated for the inventory at the time of the accident and after 50 Yrs. And found to be 1.84E+15 Bq/gm and 5.86E+12 respectively.

Peer reviewing

Paper

Accepted on 27 May 2019 by Amparo Gonzalez-Espartero

Paper files:

• SFM19_Conf._Paper_V32-final_after_Ref_Correction.docx