Yasunori Ohoka

AEGIS: AN ADVANCED LATTICE PHYSICS CODE FOR LIGHT WATER REACTOR ANALYSES

AEGIS is a lattice physics code incorporating the latest advances in lattice physics computation, innovative calculation models and efficient numerical algorithms and is mainly used for light water reactor analyses. Though the primary objective of the AEGIS code is the preparation of a cross section set for SCOPE2 that is a three-dimensional pin-by-pin core analysis code, the AEGIS code can handle not only a fuel assembly but also multi-assemblies and a whole core geometry in twodimensional geometry. The present paper summarizes the major calculation models and part of the verification/validation efforts related to the AEGIS code.

Cross section adjustment method based on random sampling technique

A cross section adjustment method based on the random sampling technique is proposed. In the proposed method, correlations among cross sections and core parameters are used instead of sensitivity coefficients of cross sections, which are necessary in the conventional method. The correlations are statistically estimated by the random sampling technique. The proposed method is theoretically consistent with the conventional method and provides comparable adjusted cross sections when sufficient number of random sampling is taken into account. The proposed method would be suitable for practical light water reactor (LWR) core analysis since estimation of sensitivity coefficients, which requires considerable computational cost in typical LWR problems, is not necessary. Through a benchmark problem in simple pin-cell geometry, adjusted cross sections by the present and the conventional cross section adjustment method are compared. The adjusted cross sections by the present method well reproduce the conventional ones, thus the feasibility of the present method is confirmed.

Uncertainty Quantification of LWR Core Characteristics Using Random Sampling Method

Abstract Uncertainties of various neutronics characteristics in commercial boiling water reactor (BWR) and pressurized water reactor (PWR) cores due to cross-section covariance are evaluated by the Latin Hypercube Sampling (LHS) method, which is an efficient random sampling algorithm. Thermal-hydraulic feedback and burnup effects are fully and explicitly taken into account using a licensing-grade core simulator. Uncertainties for various core characteristics are evaluated by the statistical processing of core calculation results based on the LHS method. The calculation results indicate that uncertainty of critical eigenvalue (i.e., core reactivity) in the BWR core is comparable to that of a typical PWR core. On the other hand, uncertainties of assembly relative power distribution and maximum assembly burnup in the present BWR core are much smaller than those of the present PWR core. The strong thermal-hydraulic feedback effect in the BWR core significantly contributes to the difference of uncertainties in BWR and PWR cores.

Important fission product nuclides identification method for simplified burnup chain construction

A method of identifying important fission product (FP) nuclides which are included in a simplified burnup chain is proposed. This method utilizes adjoint nuclide number densities and contribution functions which quantify the importance of nuclide number densities to the target nuclear characteristics: number densities of specific nuclides after burnup. Numerical tests with light water reactor (LWR) fuel pin-cell problems reveal that this method successfully identifies important FP nuclides included in a simplified burnup chain, with which number densities of target nuclides after burnup are well reproduced. A simplified burnup chain consisting of 138 FP nuclides is constructed using this method, and its good performance for predictions of number densities of target nuclides and reactivity is demonstrated against LWR pin-cell problems and multi-cell problem including gadolinium-bearing fuel rod.

Uncertainty estimation of core safety parameters using cross-correlations of covariance matrix

Abstract An uncertainty reduction method for core safety parameters, for which measurement values are not obtained, is proposed. We empirically recognize that there exist some correlations among the prediction errors of core safety parameters, e.g., a correlation between the control rod worth and the assembly relative power at corresponding position. Correlations of errors among core safety parameters are theoretically estimated using the covariance of cross sections and sensitivity coefficients of core parameters. The estimated correlations of errors among core safety parameters are verified through the direct Monte Carlo sampling method. Once the correlation of errors among core safety parameters is known, we can estimate the uncertainty of a safety parameter for which measurement value is not obtained.

Burnup and Temperature Effects on CANDLE Burnup of Block-Type High Temperature Gas Cooled Reactor

The CANDLE burnup strategy is a new reactor burnup concept, where the distributions of fuel nuclide densities, neutron flux, and power density move with the same constant speed along the core axis from bottom to top or from top to bottom of the core and without any change in their shapes. It can be applied easily to the block-type high temperature gas cooled reactor using an appropriate burnable poison mixed with uranium oxide fuel. In the present study, the burnup distribution and the temperature distribution in the core are investigated and their effects on the CANDLE burnup core characteristics are studied. In this study, the natural gadolinium is used as the burnable poison. With the fuel enrichment of 15%, the natural gadolinium concentration of 3.0% and the fuel pin pitch of 6.6cm, the CANDLE burnup is realized with the burning region moving speed of 29 cm/year and the axial half width of power density distribution of 1.5m for uniform group constant case at 900K. When the effect of nuclide change by burnup is considered, the burning region speed becomes 25cm/year and the axial half-width of power density distribution becomes 1.25m. When the temperature distributions effect is considered, the effects on the core characteristics are smaller than the burnup distribution effect. The maximum fuel temperature of the parallel flow case is higher than the counter flow case.Copyright

Verification and Validation of AEGIS/SCOPE2: The State-of-the-Art In-Core Fuel Management System for PWRs

Verification and validation of AEGIS/SCOPE2, a state-of-the-art in-core fuel management system for PWRs, was conducted. Verification of implementations for processes of tabulation and reconstruction of cross sections data in the system was confirmed through benchmark problems to confirm consistency between AEGIS and SCOPE2 in terms of reactivity and pin power distribution with identical computational condition in single- and multi-assembly geometry. In validation study numerical performance of the system was demonstrated in analyses of the B&W’s critical experiment and tracking calculations for commercial PWRs in combination of ENDF/B-VII.0 library. It is found the present system gives stability in prediction of critical boron concentration and radial power distribution.Copyright

Estimation of the Doppler Coefficient from a Lower Power Transient Observed in a Zero-Power Reactor Physics Test of a PWR (I)—Methodology—

A method for estimating the Doppler coefficient from low power transient data that are measured in a routine experiment conducted in a zero-power reactor physics test of a PWR power plant has been developed. A dynamics identification model was introduced by considering two dynamics components for reactor kinetics and heat removal in the primary coolant loops. The dynamics identification method in the time domain was applied under consideration that the effect of reactivity feedback observed in the experiment was very small and measured transient data contained some intermittent intervals unavailable for dynamics identification. Unknown parameters, including the Doppler coefficient, contained in the dynamics model were determined stepwise through five estimation steps. In these steps, gamma ray noises were removed from the NIS signal, the reactivity feedback component was extracted from reactivity transient, and fuel temperature transient was determined in such way as to reproduce the observed low power transient in the digital simulation. In the final step, the Doppler coefficient was estimated from the extracted reactivity feedback component using the quantities evaluated in the preceding steps. The estimated value agreed well with the value evaluated with a core design code.

Hard Spectrum STFR (Spray Type Fast Reactor) Cooled by Water at 7.5 MPa

New concept of fast reactor (FR), i.e. the STFR (Spray Type FR), is proposed here to perform high burn-up of UO2 fuel or MOX fuel by the use of BWR technology, and to improve the backend process of discharged fuel. STFR can be realized by some important changes of BWR system, at 7.50MPa. In case of STFR, heat produced in the core is removed by the evaporation of sprayed hot water jetted to fuel with cross flow at 7.50MPa, and two phases (liquid and vapor) of coolant at high void ratio goes down to the bottom of PV (pressure vessel). This is an improved concept of BWR, which can be regarded as a breakthrough of FBR. This concept has not been listed in GEN IV. Future performance of STFR are as follows, (1) STFR can increase the fraction of direct fission of 238 U with neutron reaction of higher energy than 1MeV, (2) STFR can burn the nuclear fuel to the higher burn-up (80 to 200GWd/Mg-HM) compared with BWR fuel burn-up. (3) Higher burn-up of fuel will reduce the frequency of reprocessing, so STFR can reduce the reprocessing cost per power production. (4) STFR can reduce the remains of Pu and MA (Minor Actinides: Np, Am, Cm etc.) in discharged fuel.Copyright