Masashi Tsuji

Subcriticality Measurement by Neutron Source Multiplication Method with a Fundamental Mode Extraction

A new neutron multiplication method has been proposed for an accurate measurement of subcriticality. The proposed method consists of two correction processes for (1) extraction of the fundamental mode from measuring data of a neutron detector that contains higher modes as well as the fundamental mode feeding from an external neutron source and (2) spatial corrections for perturbations induced by a reactivity addition in the distributions of the fundamental mode and a neutron importance field. Feasibility of the proposed method has been verified from a numerical study, although under some limitations such that the neutron multiplying system to be analyzed is small-sized and a reactivity change takes place homogeneously in a fuel loaded region. With extraction of the fundamental mode and the spatial corrections, the subcriticality can be estimated accurately even with measuring data highly contaminated with higher modes due to a detector position near to an external point neutron source. For a future application to measurement of control rod bank worth of a nuclear power plant from measuring data during a reactor physical testing, some useful guidelines have been obtained.

A genetic algorithm approach to optimization for the radiological worker allocation problem.

This paper describes a new approach to the radiological worker allocation problem using a multiple objective genetic algorithm. The worker allocation problem in radiological facilities involves various types of constraints and even mutually conflicting ones, such as individual dose limits, working time limits, etc. A major difficulty of this highly working time limits, etc. A major difficulty of this highly constrained problem is the way of finding an optimal solution in the huge search space where a large proportion of solutions are not feasible because some of the constraints cannot be satisfied. The paper proposes a model of evolution to establish an optimal assignment efficiently, based on the biological insights into the evolutionary process and heuristic ideas. The experimental results show a very rapid evolution to produce feasible solutions, and the application of multiple evaluation functions converges the feasible solutions to good ones. The genetic algorithm approach was found to be superior to the goal programming and simplex methods.

Uncertainty quantification of neutronic parameters of light water reactor fuel cells with JENDL-4.0 covariance data

Neutronic parameter uncertainty induced by nuclear data uncertainty is quantified for several light water reactor fuel cells composed of different combinations of fissile/fertile nuclides. The covariance data given in JENDL-4.0 are used as the nuclear data uncertainty, and uncertainty propagation calculations are carried out using sensitivity coefficients calculated with the generalized perturbation theory for burnup-related neutronic parameters. It is found that main contributors of nuclear data uncertainty to the neutronic parameter uncertainty are the uranium-238 capture cross section in a uranium-oxide fuel cell, and the plutonium-240 and plutonium-241 capture cross sections and fission spectrum of fissile plutonium isotopes in a uranium–plutonium mixed-oxide fuel cell. It is also found that thorium-232 capture cross section uncertainty is a dominant source of neutronic parameter uncertainty in thorium–uranium and thorium–plutonium mixed-oxide fuel cells. It should be emphasized that precise and detail information of component-wise uncertainties can be obtained by virtue of the adjoint-based sensitivity calculation methodology. Furthermore, cross-correlations are evaluated for each fuel cell, and strong correlations among the same parameters at the beginning of cycle and at the end of cycle and among different parameters are observed.

A hierarchical domain decomposition boundary element method applied to the multiregion problems of neutron diffusion equations

AbstractA technique is presented for solving neutron diffusion equations with the boundary element method (BEM) based on a hierarchical domain decomposition technique. In this method, the reactor domain is decomposed into homogeneous regions and the boundary condition on the common boundary of regions is initially assumed. The neutron diffusion equation is solved iteratively at two levels of hierarchical structure: First, BEM is applied to solve the neutron diffusion equation of each homogeneous region under the given assumed boundary conditions and an assumed multiplication factor. Then, these assumed values are modified to satisfy the continuity conditions for the neutron flux and neutron current.The proposed technique is useful for multiregion problems with a large number of regions of complex geometry, where the finite difference approximation cannot be applied properly.

Subcriticality Measurement of Pressurized Water Reactors during Criticality Approach using a Digital Reactivity Meter

A mockup experiment of subcriticality measurement with a digital reactivity meter during criticality approach in a PWR was carried out using plant data. There are some difficulties in the direct application of a digital reactivity meter to the subcriticality measurement. Some treatments were needed in using the count rate of Source Range (SR) detector as input signal to the digital reactivity meter. To overcome these difficulties, we proposed a digital reactivity meter combined with a methodology of the modified Neutron Source Multiplication (NSM) method with the fundamental mode extraction. In this study, the neutron source strength was defined as a constant in terms of a known initial stable subcriticality and the neutron signal from a steady state condition. Even though initial stable subcriticality might be contained some errors, the errors of estimated subcriticalities do not depend on the error contribution of initial stable subcriticality when we normalize the estimated subcriticalities with exactly known reference subcriticality. It was proved that the digital reactivity meter could be used not only for the subcriticality measurement or control rod worth measurement but also for the continuous monitoring of the subcriticality during criticality approach.

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.

JENDL-4.0 Benchmarking for Effective Delayed Neutron Fraction of Fast Neutron Systems

The performance of the latest Japanese evaluated nuclear data library JENDL-4.0 for the prediction of effective delayed neutron fraction βeff is assessed using experimental data of a wide range of fast neutron systems. Covariance data of JENDL-4.0 are used to quantify nuclear-data-induced uncertainties. Calculations with other libraries, JENDL-3.3, ENDF/B-VII.0, and JEFF-3.1, are also carried out for a quantitative comparison. JENDL-4.0 results in good agreement between calculation and experimental values within total uncertainties, and consistency between the differential nuclear data and integral experimental data is confirmed. While the other libraries also show good performance for βeff prediction, there are small differences in the predicted values of βeff among different libraries and ENDF/B-VII.0 gives the most stable results. Furthermore, a simple and convenient procedure to calculate sensitivity profiles of βeff to nuclear data is proposed.

Feasibility Study on Subcriticality Monitoring with a Digital Reactivity Meter

A conventional digital reactivity meter is based on a simple principle to solve inverse point reactor kinetics equations and it can monitor reactivity continuously on a real time basis. Then, feasibility was studied for a conventional digital reactivity meter to be used as a subcriticality monitor. It was necessary to overcome some problems; for example, the applicability of the point reactor kinetics equations must be verified for the system where neutron distribution is dependent on the subcriticality. We showed that the problems can be solved or can be taken into account. The subcriticality calculated by the reactivity meter might not be accurate for the measurement of the actual value of the subcriticality itself, however, it is accurate enough for the purpose of subcriticality monitoring. We believe that the monitoring on a real time basis is more important for subcriticality monitoring than the accuracy of the value of the monitored subcriticality. Based on the study, we proposed that a digital reactivity meter can be used as a subcriticality monitor.

A hierarchical domain decomposition boundary element method with a higher order polynomial expansion for solving 2-D multiregion neutron diffusion equations

Abstract A hierarchical domain decomposition boundary element method (HDD-BEM) for solving the multiregion neutron diffusion equation (NDE) has been developed to reduce computation time. The boundary integral equations derived from NDEs defined in homogeneous subregions are discretized with higher order boundary elements. The neutron flux and the neutron currents on boundary elements are expanded by quadratic or cubic polynomials. This expansion allows a large decrease in the number of unknown variables compared with the conventional HDD-BEM with constant boundary elements and reduces the computation time greatly. To obtain high accuracy with a small number of unknowns it is important to assign suitable nodal points on the non-conforming boundary elements. Guidelines for the assignment of nodal points is presented through numerical analysis. The HDD-BEM with higher order boundary elements calculates at least 5 times faster than the conventional HDD-BEM with constant boundary elements and 30 times faster than the finite difference method. The improvements in computation time will enable an extension of the scope of application to a wider variety of problems in reactor analysis.

An application of the domain decomposition method into the boundary element method for solving the multi-region neutron diffusion equation

This paper presents a technique for solving the neutron diffusion equation with the boundary element method based on the domain decomposition method. In this technique, the domain region is decomposed into homogeneous regions. The boundary conditions on the common boundary of decomposed regions and the multiplication factor are initially assumed. The boundary conditions on the other boundary are given. The neutron diffusion equation is solved iteratively at two levels of a hierarchical structure: first, the boundary element method is applied to solve the neutron diffusion equation of each homogeneous region under given assumed boundary conditions and multiplication factor. These assumed values are then modified to satisfy the continuity conditions for the neutron flux and neutron current. The proposed technique is useful for multi-region problems with a large number of regions.