SaraGR-K: a deterministic transport-based transient analysis code for prismatic gas-cooled micro-reactors

IF 2.5 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Annals of Nuclear Energy Pub Date : 2026-06-01 Epub Date: 2026-01-22 DOI:10.1016/j.anucene.2026.112130

Qiming Yang , Yuan Yuan , Dong Huang , Youqi Zheng , Guoming Liu

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引用次数: 0

Abstract

This paper presents SaraGR-K, a three-dimensional code developed for coupled neutronics/thermal-hydraulics (N/T-H) transient analysis, specifically designed for prismatic gas-cooled microreactors. The code solves the 3D neutron transport equation in the time–space domain using a predictor–corrector quasi-static (PCQS) method. To further enhance computational efficiency in 3D transport simulations, SaraGR-K incorporates a fixed-source scaling factor technique to accelerate the convergence of transient fixed-source equations and utilizes a pipelined parallelization strategy to expedite transport calculations. Temperature distributions within reactor materials are subsequently calculated using a 1D fluid dynamics model coupled with a 1D solid heat conduction model, ensuring consistent integration with the neutronics solution.

With these computational and thermal–hydraulic models in place, SaraGR-K was verified using numerical benchmarks problems and by comparing its results with those from established reference codes. Following the verification, it was applied to the transient analysis of a prismatic gas-cooled microreactor, demonstrating its feasibility for full-core transient simulations in such systems.

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SaraGR-K：基于确定性输运的柱形气冷微堆瞬态分析代码

本文介绍了SaraGR-K，一个用于耦合中子/热工-水力学（N/T-H）瞬态分析的三维代码，专门为柱形气冷微堆设计。该程序采用预测校正准静态（PCQS）方法求解三维中子输运方程。为了进一步提高3D传输模拟的计算效率，SaraGR-K采用了固定源缩放因子技术来加速瞬态固定源方程的收敛，并利用流水线并行化策略来加快传输计算。随后，使用一维流体动力学模型与一维固体热传导模型耦合计算反应堆材料内部的温度分布，确保与中子溶液的一致集成。有了这些计算模型和热水力模型，SaraGR-K就可以通过数值基准问题进行验证，并将其结果与已有的参考代码进行比较。在验证之后，将其应用于柱形气冷微堆的瞬态分析，证明了其在此类系统中进行全堆芯瞬态模拟的可行性。

本文章由计算机程序翻译，如有差异，请以英文原文为准。

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来源期刊

Annals of Nuclear Energy 工程技术-核科学技术

CiteScore

4.30

自引率

21.10%

发文量

632

审稿时长

7.3 months

期刊介绍： Annals of Nuclear Energy provides an international medium for the communication of original research, ideas and developments in all areas of the field of nuclear energy science and technology. Its scope embraces nuclear fuel reserves, fuel cycles and cost, materials, processing, system and component technology (fission only), design and optimization, direct conversion of nuclear energy sources, environmental control, reactor physics, heat transfer and fluid dynamics, structural analysis, fuel management, future developments, nuclear fuel and safety, nuclear aerosol, neutron physics, computer technology (both software and hardware), risk assessment, radioactive waste disposal and reactor thermal hydraulics. Papers submitted to Annals need to demonstrate a clear link to nuclear power generation/nuclear engineering. Papers which deal with pure nuclear physics, pure health physics, imaging, or attenuation and shielding properties of concretes and various geological materials are not within the scope of the journal. Also, papers that deal with policy or economics are not within the scope of the journal.