Yoshihira Ando

Multigroup Scattering Matrix Generation Method Using Weight-to-Flux Ratio Based on a Continuous Energy Monte Carlo Technique

We have developed a deterministic group constant generation method based on the calculation results of a continuous energy Monte Carlo technique. This method features multigroup scattering matrix generation via a weight-to-flux ratio. We performed both diffusion and transport core calculations with this set of multigroup constants generated by the proposed method, which we then validated by both a comparison with a conventional method and a critical experiment analysis. The developed method is particularly useful for innovative fuel and future core designs as Monte Carlo calculations are applicable to any heavy material and the geometrical heterogeneity thereof.

Analysis of Core Physics Experiments on Fresh and Irradiated BR3 MOX Fuel in REBUS Program

As part of an international experimental program REBUS, core physics experiments have been implemented on a UO2 core, which consists of 3.3 and 4.0 wt% UO2 fuel rods in a square pitch of 1.26 cm, and two partial MOX cores, which replace 7 × 7 UO2 fuel rods in the center of the UO2 core by fuel bundles made of fresh BR3 MOX fuel or irradiated BR3 MOX fuel with an average burnup of 20GWd/t. Burnup calculations of the BR3 MOX fuel were performed using a general-purpose neutronic calculation code SRAC, and core calculations of the three critical cores were carried out using SRAC, a transport calculation code THREEDANT, and a continuous-energy Monte Carlo code MVP. The measured inventories of major U and Pu isotopes on a sample taken from the BR3 MOX fuel agree with the results of the burnup calculations within 3% deviation. The k effs of the three cores are from 0.985 to 1.002. The measured burnup reactivity of the irradiated BR3 MOX fuel was well reproduced by the three types of core calculations. The influence of the accuracy of the inventory calculations on burnup reactivity was studied by comparing between the calculated and measured inventories. The result indicates that the biases in the inventory and reactivity calculations compensate each other, and it makes the total biases of the burnup reactivity small.

Analysis of Core Physics Experiments on Fresh and Irradiated PWR UO2 Fuels in the REBUS Program

Critical experiments of two cores each loaded with fresh 5 × 5 test PWR-type fuel rods of 235U enrichment of 3.8 wt% or irradiated 5 × 5 test rods of rod average burnup of 55 GWd/t in the REBUS program were analyzed using diffusion, transport, and continuous-energy Monte Carlo calculation codes coupled with nuclear data libraries based on JENDL-3.2 and JENDL-3.3. Biases in effective multiplication factors k effs of the critical cores were about −1:2%Δk for the diffusion calculations (JENDL-3.2), −0:5%Δk for the transport calculations (JENDL-3.3), and −0:5 and 0.1%Δk for the Monte Carlo calculations (JENDL-3.3 and JENDL-3.2, respectively). The measured core fission rate and Sc- or Co-activation rate distributions were generally well reproduced using the three types of calculation. The burnup reactivity determined using the measured water level reactivity coefficients was −2:35 ± 0:07Δk/kk′. The calculated result of the Monte Carlo calculations agreed with it; however, the diffusion and transport calculations overestimated the absolute value by about 7%, which would be mainly attributed to the errors in the calculation of the reactivity caused by changing the fuel compositions from fresh fuel to irradiated fuel.

Neutronics Analysis of Full MOX BWR Core Simulation Experiments FUBILA

A part of the experimental program FUBILA was dedicated to the study of core physics characteristics of full MOX BWR cores by testing five experimental cores: a core inserted with a B4C control blade, a core loaded with UO2 fuel rods, the core loaded with Gd2O3-UO2 fuel rods, a core loaded with 10 × 10 MOX assemblies and a core loaded with time-elapsed MOX fuel. The present article describes analysis results of the experimental data with deterministic analysis codes and a continuous energy Monte Carlo code coupled with major nuclear data libraries. The calculated critical k effs with the Monte Carlo calculations range from 0.999 to 1.007. Those of the transport calculations with sixteen energy groups are close to those of the Monte Carlo calculations while those of diffusion calculations with the same sixteen energy groups are systematically smaller by −0.3 to −0.5% Δk than those of the Monte Carlo calculations. The RMSs of differences between the calculated and measured core radial fission rates are 2 to 3%, 1 to 2%, and 1 to 2% for the diffusion, transport and Monte Carlo calculations, respectively. For the analysis of the Gd2O3-UO2 fuel rod loaded core, the C/Es of the radial fission rates were improved by adopting a detailed lattice calculation model and a newly measured thermal and resonance cross-sections of Gadolinium.

Analysis of measured isotopic compositions of high-burnup PWR MOX and UO2 fuels in the MALIBU program

The measured isotopic compositions of fuel samples taken from high-burnup spent PWR MOX and UO2 assemblies in the MALIBU program has been analyzed by lattice physics codes. The measured isotopes were U, Np, Pu, Am, and Cm isotopes and about 30 major fission product nuclides. The codes used in the present study were a continuous-energy Monte Carlo burnup calculation code (MVP-BURN) and a deterministic burnup calculation code (SRAC) based on the collision probability method. A two-dimensional multi-assembly geometrical model (2 × 2 model) was mainly adopted in the analysis in order to include the fuel assemblies adjoining the relevant fuel assembly, from which the samples were taken. For the MOX sample, the 2 × 2 model significantly reduces the deviations of the calculated results from the measurements compared with a single assembly model. The calculation results of MVP-BURN in the 2 × 2 model reproduce the measurements of U, Np, and Pu isotopes within 5% for the MOX sample of 67 GWd/t. The deviations of their calculated results of U, Np, and Pu isotopes from the measurements are less than 7% for the UO2 sample of 72 GWd/t.

Analysis of Measured Isotopic Compositions of High-Burnup BWR MOX Fuel

Analysis of measured isotopic compositions of four high-burnup BWR MOX fuel samples was performed by using a general-purpose neutronic calculation code SRAC and a continuous-energy Monte Carlo burnup code MVP-BURN. The initial Pu fissile content of the samples was 5.52 wt%, and the burnups ranged from 50 to 80 GWd/t. It is confirmed that a geometrical model including the effect of UO2 assemblies adjacent to the MOX assembly is necessary in the burnup calculations to obtain accurate calculated isotopic compositions. The calculated results of MVP-BURN with JENDL-3.3 taking such effect into account show more accurate results for major actinides (U, Pu, and Am isotopes) and most fission products than those of infinite assembly calculations. The paper also shows the results calculated using SRAC with JENDL-3.3, ENDF/B-VII, and JEFF-3.1.

Analysis of Reactivity Measurements in Core Physics Experiments on Full-MOX BWR

A core physics experimental program FUBILA has been performed to study core physics characteristics of full-MOX BWR cores consisting of high Pu-enriched MOX assemblies for high burnups. The program includes the measurement of reactivity worth, which is essential in operating BWR cores. The reactivity worth is due to the reactivity caused by (1) changes in the in-channel void fraction of assemblies, (2) the insertion of a B4C control blade, (3) Gd2O3-UO2 rods and UO2 rods in assemblies, and (4) the mixing of boron in a moderator related to a stand-by liquid control system. The reactivity worth was measured by the modified neutron source multiplication method using the reactivity worth of a pilot rod as a reference worth. The measured reactivity worth was determined by processing count rates of neutron detectors taking into account the detector efficiency and effective neutron source intensity analyzed by three-dimensional transport calculations. Comparing the measured reactivity worth with the results obtained by deterministic calculation methods and a Monte Carlo calculation method, the accuracy of the calculations was evaluated. The calculated results generally well reproduce the measurements, except for the boron reactivity worth in the moderator, for which the calculations overestimate the reactivity worth.

Analysis of High Moderation PWR MOX Core MISTRAL-4 with SRAC and MVP

An extensive experimental program, MISTRAL, is undertaken by NUPEC and CEA in order to measure the main core physics parameters of high moderation 100% MOX LWRs. The analysis of the MISTRAL-4, a mock-up of a full MOX PWR core was carried out by diffusion/transport calculations with the SRAC code system and by continuous-energy Monte Carlo calculations with the MVP code. The calculation results agreed with the experimental results of both calculations within experimental uncertainties and calculation results showed no specific trend caused by heterogeneity in highly-moderated mock-up core configurations.

A New Estimation Method for Nuclide Number Densities in Equilibrium Cycle

A new method is proposed for estimating nuclide number densities of LWR equilibrium cycle by multi-recycling calculation. Conventionally, it is necessary to spend a large computation time for attaining the ultimate equilibrium state. Hence, the cycle in nearly constant fuel composition has been considered as an equilibrium state which can be achieved by a few of recycling calculations on a simulated cycle operation under a specific fuel core design. The present method uses steady state fuel nuclide number densities as the initial guess for multi-recycling burnup calculation obtained by a continuously fuel supplied core model. The number densities are modified to be the initial number densities for nuclides of a batch supplied fuel. It was found that the calculated number densities could attain to more precise equilibrium state than that of a conventional multi-recycling calculation with a small number of recyclings. In particular, the present method could give the ultimate equilibrium number densities of ...

Analysis of MISTRAL Experiments with JENDL-3.2

Nuclear Power Engineering Corporation has been analyzing measurement data of MISTRAL, a LWR MOX core physics experimental program, with a deterministic code system of the diffusion and the transport theories, SRAC, and a Monte Carlo transport code, MVP, based on JENDL-3.2. The program consists of one reference U02 core, two homogeneous full MOX cores and one full MOX PWR mock-up core that have higher moderation ratio than the conventional lattice. The measurement parameters cover critical masses, boron concentrations, radial and axial core power distribution, spectrum indexes, conversion factors, effective delayed neutron fractions, iso-thermal temperature coefficients, boron efficiency, single absorber worth, water hole worth, absorber cluster worth and void worth. Critical keff calculations show a slight overestimate for MOX cores and a small systematic trend with aging of MOX fuel. Power distribution calculations show good agreement with the measurements in both homogeneous and mock-up cores. Measured spectrum indexes of fissile isotopes are well reproduced by the calculation. For other measurements, calculated results well reproduce the measurements within one or two times of the experimental error. The analysis verifies the validity of JENDL-3.2 and the analysis methods for the high moderation LWR MOX cores.