Speaker
Description
The Japan Atomic Energy Agency (JAEA) has developed and maintained several burnup calculation codes, such as SWAT4 [1] and MVP-BURN [2], which have been widely used for research and various nuclear evaluations. However, recent updates and expansions of evaluated nuclear data libraries have made it difficult to apply new nuclear data to these codes because of limitations in the number of nuclides, nuclear reactions, and decay modes that can be treated.
To address this issue, JAEA has been developing a new burnup calculation code system called SWAT-X. To enable flexible utilization of modern nuclear data, SWAT-X is being developed from scratch using Python 3. As a first step, a basic burnup calculation function was implemented by coupling burnup calculations with CRAMO [3] and neutron transport calculations with MVP3 [4]. The validity of this function was confirmed through comparisons between the burnup calculation results of SWAT-X and SWAT4.
At present, SWAT-X includes a function that can automatically construct arbitrary depletion chains using data from evaluated nuclear data libraries. In this function, detailed depletion chains are generated by reading ENDF-6 formatted decay and fission yield data. For cross sections, reaction paths are defined using the SWAT-X library, which contains multi-group infinite-dilution cross-section data derived from GENDF files produced by FRENDY v2 [5]. One-group cross sections for user-selected major nuclides are obtained by MVP3 using continuous-energy data, while those for other nuclides are calculated by collapsing the multi-group cross sections with neutron fluxes from MVP3. The depletion chain can be systematically simplified by selecting specific nuclides to be included or by applying half-life thresholds to determine whether certain decays are considered. This function enables burnup calculations using a detailed burnup chain based on JENDL-5, comprising approximately 4,070 nuclides.
In parallel, we are developing a fast burnup calculation capability using neutron spectrum reconstruction, as an improved approach to the one-point calculation method of ORIGEN2 [6]. This method employs a reduced-order model (ROM) constructed from neutron spectrum snapshots using the proper orthogonal decomposition and regression models. The ROM allows rapid neutron spectrum estimation at each burnup step, greatly reducing computation time by eliminating repeated neutron transport simulations.
This presentation will introduce the SWAT-X system, describe its calculation capabilities, and present verification results.
This work was supported in part by JSPS KAKENHI Grant Number JP24K08300.
References
[1] Kashima T, Suyama K, Takada T. SWAT4.0-The integrated burnup code system driving continuous energy Monte Carlo codes MVP, MCNP and deterministic calculation code SRAC. Ibaraki: JAEA; 2015. JAEA-Data/Code 2014-028 [in Japanese].
[2] Okumura K, Mori T, Nakagawa M, et al. Validation of a continuous-energy Monte Carlo burn-up code MVP-BURN and its application to analysis of post irradiation experiment. J Nucl Sci Technol. 2000;37(2):128–138.
[3] Yokoyama K, Jin T. Development of burnup/depletion calculation code based on ORIGEN2 cross-section libraries and Chebyshev rational approximation method, CRAMO. Ibaraki: JAEA; 2021. JAEA-Data/Code 2021-001 [in Japanese].
[4] Nagaya Y, Okumura K, Sakurai T, et al. MVP/GMVP version 3: general purpose monte carlo codes for neutron and photon transport calculations based on continuous energy and multigroup methods (Translated document). Ibaraki: JAEA; 2017. JAEA-Data/Code 2016-019.
[5] Tada K, Yamamoto A, Kunieda S, et al. Development of nuclear data processing code FRENDY version 2. J Nucl Sci Technol. 2024;61(6):830–839.
[6] Croff AG. ORIGEN-2: a revised and updated version of Oak Ridge Isotope generation and development code. 1980; ORNL-5621; Oak Ridge National Laboratory.
