Yasunobu Nagaya

JENDL-4.0 Benchmarking for Fission Reactor Applications

Benchmark testing for the newly developed Japanese evaluated nuclear data library JENDL-4.0 is carried out by using a huge amount of integral data. Benchmark calculations are performed with a continuous-energy Monte Carlo code and with the deterministic procedure, which has been developed for fast reactor analyses in Japan. Through the present benchmark testing using a wide range of benchmark data, significant improvement in the performance of JENDL-4.0 for fission reactor applications is clearly demonstrated in comparison with the former library JENDL-3.3. Much more accurate and reliable prediction for neutronic parameters for both thermal and fast reactors becomes possible by using the library JENDL-4.0.

Comparison of Resonance Elastic Scattering Models Newly Implemented in MVP Continuous-Energy Monte Carlo Code

In order to investigate the impact of resonance elastic scattering models on the Doppler reactivity effect, the exact and the constant cross section models with the thermal motion of target nucleus were newly implemented into the MVP-2 continuous-energy Monte Carlo code with and without consideration of energy-dependent resonance cross section, respectively, and the UO2 pin-cell Doppler reactivity benchmark calculations were carried out with the modified code and the JENDL-3.3 library. The present study has revealed that the exact model gives more negative Doppler reactivity coefficients by 7–11% than the conventional asymptotic model, while the constant cross section model gives slightly less negative coefficients than the conventional asymptotic model. Furthermore, it has been found that the impact of resonance elastic scattering models is considerably large around the resonances where elastic scattering has relatively high contribution, whereas capture-dominant resonances have no significant impacts of the models on the Doppler reactivity coefficients.

Development and verification of a new nuclear data processing system FRENDY

ABSTRACT Japan Atomic Energy Agency developed the new nuclear data processing system FRom Evaluated Nuclear Data librarY to any application (FRENDY) in order to solve the problems of the current widely used nuclear data processing systems and process the new evaluated nuclear data file. Verification of FRENDY was carried out by three steps, i.e. verification of each function, comparison of the results, and comparison of the keff values for the 79 benchmark experiments in the ICSBEP handbook using cross section data library processed by FRENDY with those by NJOY99. These results verified that FRENDY generates the ACE (A Compact ENDF) file correctly.

On Effective Delayed Neutron Fraction Calculations with Iterated Fission Probability

The iterated fission probability (IFP) is a quantity proportional to the asymptotic power level originated by a neutron introduced to a reactor. The effective delayed neutron fraction βeff can be accurately calculated by the continuous-energy Monte Carlo method using IFP if a sufficiently large number of generations is considered to obtain the asymptotic state. In order to deterministically quantify the required number of generations in the IFP-based βeff calculations, the concept of the generation-dependent importance functions is introduced to βeff calculations. Furthermore, the most appropriate reactor property used in the IFP calculations, which reduces the required number of generations, is theoretically derived. Through numerical calculations, it is shown that several generations are required in the IFP-based βeff calculations and that the use of the appropriate reactor property can reduce the required number of generations. An efficient procedure for the IFP-based βeff calculations by the Monte Carlo method is also proposed.

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.

Impact of Incident Energy Dependence of Prompt Fission Neutron Spectra on Uncertainty Analyses

This paper investigates the impact of the incident energy dependence of prompt fission neutron spectra (PFNS) on uncertainty propagation calculations. Uncertainty propagation from incident energy-dependent PFNS to criticality is formulated and its impact is evaluated numerically. It is found that the conventional procedure, in which representative PFNS covariance data for a specific incident energy are used, results in a larger PFNS-induced uncertainty than the straightforward procedure, in which different PFNS covariance data are used for each incident energy range given in the nuclear data libraries. The present study suggests that the correlation between different incident energies of PFNS has a large impact on uncertainty propagation calculation results for nuclear characteristics.

An analysis of a heterogeneous core using Green's function

Abstract A nodal method to solve the multi-group diffusion equation is given to treat the heterogeneity of fuel assemblies of reactors. Nodal equations of 3-point form in each coordinates are derived for the diffusion equation in one dimensional slab and x-y geometries using Greens function whose unknowns are only fluxes at the boundary surfaces of assemblies. Coefficients of the nodal equations are given by the boundary values of Greens function, which are also obtained by solving 3-point difference equations in each coordinates for the heterogeneous assembly regarding it as being isolated. In the present method, it is not necessary to use homogenized cross sections for the assembly, and the heterogeneity in the fuel assembly and interaction effect between assemblies can be taken into account.