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In a recent study, we have developed an efficient flux distribution tallying method in the Monte Carlo calculation toward the high-fidelity, large scale multi-physics simulation. In this method, the proper orthogonal decomposition is applied to the flux distribution tallies. While the tallying method was implemented with the collision estimator in the previous study, the track length estimator is implemented in the present study to obtain the tally with lower statistical error. The implementation of the flux distribution tally with the track length estimator is compared with that of the collision estimator and the normal track length estimator in a one-dimensional problem. The numerical results reveal that the distribution tally using the proper orthogonal decomposition with the track length estimator can obtain a more precise solution compared with that with the collision estimator. Therefore, in terms of the statistical error, the relationship between the distribution tally with track length and collision estimators is similar to that between the conventional track length and collision estimators.

A Resonance calculation using energy Spectral Expansion (RSE) method has been recently proposed in order to efficiently treat complicated heterogeneous geometry and resonance interference effect. In the RSE method, ultra-fine group spectra are generated from ultra-fine group calculations in homogeneous geometry, and the spectra are expanded by the orthogonal basis on energy based on the singular value decomposition. Then the transport calculation for expansion coefficients is numerically performed, and the ultra-fine group spectra in the target heterogeneous regions are reconstructed by the expansion coefficients and the orthogonal basis. In this study, the RSE method is applied to multi-cell geometries including UO2, MOX and water cells, in which the resonance interference effect between UO2 and MOX fuel cells appears. The validity of the RSE method is confirmed through comparison with the reference effective multi-group cross sections obtained from the direct ultra-fine group calculation in the target heterogeneous geometry.

In this paper, a new tallying method for a neutron flux distribution using the proper orthogonal decomposition is proposed for dimensionality reduction. The target spatial flux distribution is expanded by orthogonal basis vectors. Expansion coefficients are tallied during the random walk of the Monte Carlo calculation. The orthogonal basis vectors are extracted from the pre-calculated snapshots by the singular value decomposition. The proposed method is verified in the multi-group Monte Carlo calculation with the one-dimensional heterogeneous whole core geometry as a feasibility study. The flux distribution for each of the assemblies and energy groups is expanded by the basis vectors. The fewer basis vectors obtained from snapshots can reconstruct the target distribution well compared with the conventional Legendre polynomials used in the functional expansion tallies. The dimension of the solution in the proposed method is reduced by a factor of twenty compared with the conventional cell tally. In addition, the statistical error is reduced through dimensionality reduction thanks to the methodological feature of the proposed method. The results indicate that the proposed method has the capability of dimensionality reduction to tally the finely discretized flux distribution. [ABSTRACT FROM AUTHOR]

Nuclear data processing is important to connect evaluated nuclear data libraries and radiation transport codes. The nuclear data processing code FRENDY version 1 was released in 2019 to generate ACE-formatted cross section files with simple input data. After we released FRENDY version 1, many functions were developed, e.g. neutron multi-group cross section generation, explicit consideration of the resonance interference effect among different nuclides in a material, adaptive setting of the background cross sections, consideration of the resonance upscattering, ACE file perturbation, statistical uncertainty quantification of probability table, and modification of ENDF-6-formatted file. FRENDY version 2 was released including these new functions. It generates GENDF- and MATXS-formatted neutron multi-group cross section files from an ACE-formatted cross section file or an evaluated nuclear data file. This paper explains the features of the new functions implemented in FRENDY version 2 and the verification of the neutron multigroup cross section generation function of this code. [ABSTRACT FROM AUTHOR]

Surgical intervention for HB with tumor thrombi extending into the IVC and the RA might requires careful planning of the surgical procedures, including vascular reconstruction and extracorporeal circulation. We herein report a successful case of LDLT for HB with atrial extension of a tumor thrombus by extracorporeal circulation with a transdiaphragmatic approach. The patient was a 5‐year‐old boy with PRETEXT IV HB with a tumor thrombus that extended into the IVC and the RA. After 4 cycles of chemotherapy and resection of bilateral lung metastases, the size of the primary HB tumor decreased. As the tumor extension from the LHV to the RA had decreased but was still present, we performed LDLT with tumor thrombectomy. The central part of the diaphragm was sagittally incised to expose the suprahepatic IVC and the RA. Venovenous bypass was achieved from the right femoral vein and IMV to the RA En bloc resection of the native liver with the tumor thrombus was then performed. HV anastomosis was made between the newly created orifice on the IVC and the graft LHV. The duration of LDLT was 10 hours and 44 minutes (extracorporeal circulation time: 78 minutes). Pediatric LT for HB with the extension of tumor thrombi into the RA under extracorporeal circulation is a feasible option and allows for the expansion of the indications for transplantation for children with unresectable liver tumors. [ABSTRACT FROM AUTHOR]