Development of an Advanced Core Analysis System for Boiling Water Reactor Designs

Maruyama, Hiromi; Koyama, Jun-ichi; Aoyama, Motoo; Ishii, Kazuya; Zukeran, Atsushi; Kiguchi, Takashi; Nishimura, Akira

doi:10.13182/nt97-a35351

Cited by 9 publications

(3 citation statements)

References 8 publications

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“…Figure 11 shows the loading pattern of the core, which was taken from a multi-enrichment initial core of a 1,100 MWe BWR. 14) Three kinds of assemblies, HEA, MEA, and LEA, the same as for the multi-assembly calculation in the previous chapter were used. The neutron collision history was also the same.…”

Section: Application Of Acs Methods For Whole Core Calculation 1 mentioning

confidence: 99%

Multi-Assembly Analysis with an Advanced Correlated Sampling Method

Moriwaki

2003

Journal of Nuclear Science and Technology

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The Advanced Correlated Sampling (ACS) method, which can greatly enhance the calculation efficiency of Monte Carlo (MC) methods using a perturbation technique and a pseudo-scattering technique, is improved so as to be able to handle multi-assembly analysis. Two boundary connection methods-the neutron path connection and neutron current connection methods-are developed, and their accuracies for a 2×2 multi-assembly system are confirmed. The speedup factor against the corresponding direct MC calculation is a maximum of about 100. These boundary connection methods are also applied in pin-wise 2-D quarter core calculations as an example of a large-scale system. A newly developed perturbed neutron source iteration technique allows for very smooth source convergence. Furthermore, by incorporating the Direct Response Matrix (DRM) method as the acceleration means for the source convergence, the number of source iteration cycles is significantly shortened. The results confirm that this improved ACS method has the ability to perform pin-wise whole core calculations very accurately and efficiently.

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Section: Application Of Acs Methods For Whole Core Calculation 1 mentioning

confidence: 99%

Multi-Assembly Analysis with an Advanced Correlated Sampling Method

Moriwaki

2003

Journal of Nuclear Science and Technology

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“…Then, a spectral history method was proposed for the conventional analysis method. 6) In this method, the burn-up average of nodal thermal-to-fast flux ratio is used to represent neutron spectral conditions during fuel burn-up (spectral history). Because the 3D-DRM method does not use nodal flux, the spectral history method using the neutron current was developed.…”

Section: Development Of Spectral History and Pin Power Correction Metmentioning

confidence: 99%

Development of Spectral History and Pin Power Correction Methods for Pin-by-Pin Core Analysis Method Using Three-Dimensional Direct Response Matrix

Mitsuyasu

Ishii

Hino

et al. 2010

J. Nucl. Sci. Technol.

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Spectral history and pin power correction methods have been developed for the pin-by-pin core analysis method using the three-dimensional direct response matrix (3D-DRM). The direct response matrix is formalized using four subresponse matrices in order to respond to a core eigenvalue k and thus it can be recomposed at each outer iteration in the core analysis. For core analysis, it is necessary to take into account the historical effect, which is related to spectral heterogeneity. The spectral history method is used to evaluate the nodal burn-up spectrum obtained by using the outgoing neutron current instead of the nodal flux because the 3D-DRM method does not use the nodal flux. The pin power correction method corrects the fuel rod neutron production rates obtained in the pin-by-pin calculation. These two methods were tested in a heterogeneous system. The test results show that the neutron multiplication factor error and nodal neutron production rate errors can be reduced by half during burn-up. The root-mean-square differences between the relative fuel rod neutron production rate distributions and the maximum error of relative fuel rod production rate can also be reduced by half. This means that the developed methods can reflect the effects of intra-and interassembly heterogeneities during burn-up and can be used for core analysis.

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“…Several core analysis codes of a fuel rod geometry unit (i.e., pin-by-pin) have been developed recently for this purpose [1,2]. However, in general, these codes require longer computation times than conventional core analysis codes of a fuel assembly geometry unit [3,4]. If the accuracy of such conventional core analysis codes can be improved, they would be very useful for the design calculations, which need huge parameter surveys for example.…”

Section: Introductionmentioning

confidence: 99%

Reconstruction method of homogenized cross sections

Ishii

2013

Journal of Nuclear Science and Technology

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The reconstruction method of homogenized cross sections in the direct response matrix method has been developed. In this reconstruction method, homogenized cross sections, which take into consideration the influences of neighboring fuel assemblies, can be reconstructed with the response relationship of incoming neutron partial currents and neutron production rates. Calculations for heterogeneous multi fuel assembly systems were done to verify the developed method. The thermal energy group fuel assembly cell-averaged homogenized cross sections reconstructed by this method agreed with those evaluated by the direct calculation of the whole system using the Monte Carlo method within 0.2%. The effect using the reconstructed fuel assembly cell-averaged homogenized cross sections in a conventional core analysis code using cross sections homogenized in a fuel assembly cell was also investigated. The results obtained showed that the analysis accuracy of k-infinity can be improved by using the cross sections reconstructed by the method. Because almost no influences on the analysis accuracy could be found related to the divided numbers of the surfaces and the angles, and the response relationship with neutron production rates of fuel rods or a fuel assembly cell-averaged neutron production rate, this reconstruction method is applicable to a conventional core analysis code using homogenized cross sections in a fuel assembly cell.

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Development of an Advanced Core Analysis System for Boiling Water Reactor Designs

Cited by 9 publications

References 8 publications

Multi-Assembly Analysis with an Advanced Correlated Sampling Method

Multi-Assembly Analysis with an Advanced Correlated Sampling Method

Development of Spectral History and Pin Power Correction Methods for Pin-by-Pin Core Analysis Method Using Three-Dimensional Direct Response Matrix

Reconstruction method of homogenized cross sections

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