Comparative of blanket reactor design in assuring of tritium self-sufficiency

The comparison of four blanket modules in DEMO made up the optimization material having a reasonable requirement as blanket material in this study. Either neutron flux distribution in blanket material or material endurance under neutron irradiation, from four modules, the WCLL has a high tolerance neutron distribution and the best neutron irradiation endurance. Furthermore, many suggestions closed to the statement to use the benefits of water coolant and lithium lead (compose Li-6) as a material component in the blanket.

Determination of thermal neutron flux distribution at the rotary rack served for elemental concentration analysis using the k0-INAA method

The accuracy of elements concentration determination using the k0-standardization method directly depends on irradiation and measurement parameters including Non-1/E epithermal neutron flux distribution shape α (ϕ epi ≈1/E1+α ) , thermal-to-epithermal neutron flux ratio f, efficiency ε, peak area… In the case of the irradiation position at the rotary rack of the Dalat Nuclear Research Reactor (DNRR), the difference of thermal neutron flux between the bottom (3.54x1012 n.cm-2.s-1) and the top (1.93x1012 n.cm-2.s-1) of the 15 cm aluminum container is up to 45%. Therefore, it is necessary to accurately determine above-mentioned parameters in the sample irradiation position. The present paper deals with the determination of the distribution of thermal neutron flux along the sample irradiation container by using 0.1% Au–Al wire activation technique. The thermal neutron flux was then used to calculate the concentration of elements in the Standard Reference Material 2711a and SMELS type III using k0-INAA method at different positions in the container. The obtained results with the neutron flux correction were found to be in good agreement with the certified values. In conclusion, the proposed technique can be applied for activation analyses without sandwiching flux monitors between samples during irradiations.

Reactivity Effect Evaluation of Fast Reactor Based on Angular-Dependent Few-Group Cross Sections Generation

Among all the possible occurring reactivity effects of a fast reactor, the situations whereby the control rod was inserted, or the coolant was voided could lead to strong anisotropy of neutron flux distribution, therefore the angular dependence on neutron flux should be considered during the few-group cross-sections generation. Therefore, the purpose of this paper is to compare the influence whether the angular dependence on neutron flux is considered in the calculation of few-group cross sections for the reactivity effect calculation. In the study, the 1-D SN finite difference neutron transport equation solver was implemented in the TULIP of SARAX code system so that the high-order neutron flux could be obtained. Meanwhile, the improved Tone’s method was also applied. The numerical results were obtained based on three experimental FR cores, the JOYO MK-I core, ZPPR-9 core, and ZPPR-10B core. Both control rod worth and sodium void reactivity were calculated and compared with the measurement data. By summarizing and comparing the results of 46 cases, significant differences were found between different consideration of the neutronic analysis. The consideration of angular dependence on neutron flux distribution in the few-group cross-sections generation was beneficial to the neutronic design analysis of FR, especially for the reactivity effect calculation.

APPLICATION OF A CUSTOM DEPLETION FRAMEWORK TO THE PREDICTION OF NEUTRON FLUX DISTRIBUTION THROUGH DEPLETION

Previous works by the authors have introduced the spatial flux variation method (SFV) for predicting the changes in neutron flux due to a change in material compositions. In order to remove a full transport solution at the end-of-step, this work presents a framework responsible for computing macroscopic cross sections after a depletion event. These end of-step cross sections are estimators of changes in neutron loss and production, and enable the prediction of neutron flux using only information obtained from a single beginning of-step transport solution. The framework reads in all relevant data needed to model the depletion system, including one-group cross sections and effective fission yields to reproduce the problem using an external solver. The framework also supports extrapolating microscopic cross sections in order to rebuild the end-of-step macroscopic cross sections needed for the flux prediction. Results indicate that the SFV method is not adversely effected by the external depletion solution, and can be implemented alongside an existing transport-depletion framework.