Yasushi Nauchi

Proposal of Direct Calculation of Kinetic Parameters βeff and Based on Continuous Energy Monte Carlo Method

Direct calculation methods of kinetic parameters are proposed based on the continuous energy Monte Carlo method. In the proposed methods, the effective delayed neutron fraction βeff and the neutron generation time ∧ are estimated using eigenvalue calculations. The expected number of fission neutrons in the next generation is newly applied to the proposed methods instead of the adjoint flux that has been conventionally used. The algorithms to estimate the kinetic parameters are established and incorporated into the continuous energy Monte Carlo transport calculation code MCNP-4C, which is versatile for eigenvalue calculations of nuclear reactor cores with various types of neutron energy spectrum and geometry. The proposed methods were validated since the calculated values agreed with the experimental data of βeff and βeff/∧ for the critical cores within accuracies of 4.5% and 10%, respectively.

Development of Calculation Technique for Iterated Fission Probability and Reactor Kinetic Parameters Using Continuous-Energy Monte Carlo Method

New algorithms and techniques are developed for calculating the iterated fission probability (IFP ) using a Monte Carlo eigenvalue calculation scheme. The proportionality of the calculated IFP to the adjoint flux is confirmed numerically. This IFP is then used with the MCNP code and its point-wise cross section data libraries to calculate kinetic parameters weighted by the adjoint flux. Experimental data for critical cores validate these theoretical estimates.

Early construction and operation of the highly contaminated water treatment system in Fukushima Daiichi Nuclear Power Station (IV) - Assessment of hydrogen behavior in stored Cs adsorption vessel

Hydrogen diffusion behavior in a cesium adsorption vessel is assessed. The vessel is used to remove radioactive substance from contaminated water, which is proceeded from Fukushima accident. Experiment and numerical calculation are conducted to clarify the characteristics of natural circulation in the vessel. The natural circulation arising from the temperature difference between inside and outside the vessel is confirmed. We develop an evaluation model to predict the natural circulation and its prediction agrees well with the results obtained by the experiment and the calculation. Using the model, we predict steady and transient behavior of hydrogen concentration. Results indicate that hydrogen concentration is kept lower than the flammability limit when the short vent pipe is open.

Concept of capture credit based on neutron-induced gamma ray spectroscopy

Capture credit (CapC) based on neutron-induced gamma ray spectroscopy (NIGS) is proposed to confirm the subcriticality of fuel debris in which nuclear fuel and structural materials are co-melted or mixed. By NIGS, rates of some capture reactions can be measured in relation to fission reactions. By the ratio, we can credit the negative reactivity inserted by the capture reactions. The theory of CapC is described. In order to demonstrate the benefit to take CapC for storage of the fuel debris, numerical simulations are performed for a hypothetical array of canisters in which the fuel debris is stored. A procedure of CapC based on NIGS is also proposed, which consists of several technologies: (1) NIGS, (2) simulations of a response and an efficiency of the γ ray detection, and (3) unfolding of the γ ray pulse height spectrum to obtain reaction rates. Experimental studies of NIGS have been launched in Kyoto university critical assembly facility. NIGS is firstly studied for simulated fuel debris of a few kinds of mixture of stainless steel and uranium in subcritical systems. The measured γ ray pulse height spectra and preliminary analyses indicate that CapC based on NIGS is worth to be investigated further for the efficient storage of fuel debris.