Tsukasa Kikuchi

Kinetic Parameter βeff/ℓ Measurement on Low Enriched Uranyl Nitrate Solution with Single Unit Cores (600𝛗;, 280T, 800𝛗;)of STACY

The kinetic parameter βeff/ℓ of low enriched (10% EU) uranyl nitrate solution was measured by the pulsed neutron source (PNS) method in STACY This measurement was repeated systematically over several uranium concentrations from about 195 gU/l to about 430 gU/l. Used core tanks were two cylindrical tanks whose diameters are 600 mm and 800 mm and one slab tank which has 280 mm thickness and 700 mm width. In this report, experimental data such as solution conditions, critical solution level for each solution condition, subcritical solution levels where measurements were conducted, measured decay time constants of prompt neutron and extrapolated βeff/ℓ values are described as well as basic principle of the PNS method. The kinetic parameter βeff/ℓ values were evaluated also by computation with the diffusion code CITATION in SRAC and the nuclear data library JENDL 3.2. Strong linear correlation has been found between kinetic parameter βeff/ℓ and uranium concentration regardless of differences of reflecting conditions or core tank conditions. Experimentally or computationally evaluated βeff/ℓ value is about 90 s-1 at 195 gU/l and about 170 s-1 at 430 gU/l and both experimental and computational values show good agreement within an error of 3% which is comparable to about 2% uncertainty of measurement.

Intra-pellet neutron flux distribution measurements in LWR critical lattices

We have developed inexpensive and easy-handling measurement methods on intra-pellet neutron flux. A foil activation method with metallic foils, which were fabricated by punching out technique and etching technique to reduce fabrication error and positioning error, was used for the intra-pellet neutron flux distribution measurement. The developed method was applied to measure intra-pellet neutron flux distributions in a reduced–moderation light water reactor (LWR) lattices, and uncertainty of the distributions was estimated to be 1% to 2%. Measured values were analyzed with a continuous energy Monte Carlo code. Comparison of measurements and analyses revealed that the developed method is useful for the validation of an advanced fuel design method considering neutron behavior in fuel pellets.

Void reactivity evaluation by modified conversion ratio measurements in LWR critical experiments

We have developed a void reactivity evaluation method by using modified conversion ratio measurements in a light water reactor (LWR) critical lattice. Assembly-wise void reactivity is evaluated from the “finite neutron multiplication factor”, k*, deduced from the modified conversion ratio of each fuel rod. The distributions of modified conversion ratio and k* on a reduced-moderation LWR lattice, for which the improvement of negative void reactivity is a serious issue, were measured. Measured values were analyzed with a continuous-energy Monte Carlo method. The measurements and analyses agreed within the measurement uncertainty. The developed method is useful for validating the nuclear design methodology concerning void reactivity.

Benchmark experiment for radio-activation of ex-core structural materials using NCA and analysis with three-dimensional transport codes

In this study, a radio-activation experiment was conducted using stainless steel outside the active fuel region (active core) in the Toshiba Nuclear Critical Assembly (NCA) in order to verify the homogenization method by simulating the NCA experimental reactor system and understand the effects of this method on the analysis accuracy. In order to validate homogenization method, we simulated the system using the continuous energy Monte Carlo code MCNP, which allows heterogeneously modeling, and examined application of the homogenization method used for modeling commercial boiling water reactors (BWRs) with the TORT code. The calculation results of activation rate obtained by using the MCNP code with either heterogeneous or homogeneous models do not affect the calculation result of activation rate outside the active core. As the homogenization method was validated, the calculation of activation rate using the TORT code was performed with the same homogeneous model as in the MCNP calculation. The results of the activation rate calculation using the TORT code gave values 20 to 30% larger than the calculation results obtained via MCNP for 55Mn. This is considered to be caused by thermal energy group structure which is treated as one group.