The high-temperature gas-cooled reactor pebble-bed module (HTR-PM) nuclear power plant consists of two nuclear steam supply system modules, each of which drives the steam turbine by the superheated steam flow and is fed by the heated-up water flow. The shared steam/water system induces mutual effects on normal operation conditions and transients of the nuclear power plant, which is worthy of safety concerns and intensive study. In this paper, a coupling code package was developed with the TINTE and vPower codes to understand how the HTR-PM operated. The TINTE code was used to analyze the reactor core and primary circuit, while the vPower code simulated the steam/water flow in the conventional island. Two TINTE models were built and coupled to one vPower model through the data exchange in the steam generator models. Using this code package, two typical transients were simulated by decreasing the primary flow rate or introducing the negative reactivity of one module. Important parameters, including the reactor power, the fuel temperature, and the reactor inlet and outlet helium temperatures of two modules, had been studied. The calculation results preliminarily proved that this code package can be further used to evaluate working performance of the HTR-PM.

中文翻译：

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带有TINTE-vPower耦合代码包的HTR-PM的瞬态研究

高温气冷堆卵石床模块（HTR-PM）核电站由两个核蒸汽供应系统模块组成，每个模块都通过过热蒸汽流驱动蒸汽轮机，并由加热的水给水流。共享的蒸汽/水系统对正常运行条件和核电厂的瞬变产生相互影响，这值得安全方面的关注和深入研究。在本文中，使用TINTE和vPower代码开发了一个耦合代码包，以了解HTR-PM的工作方式。TINTE代码用于分析反应堆堆芯和一次回路，而vPower代码用于模拟常规孤岛中的蒸汽/水流。建立了两个TINTE模型，并通过蒸汽发生器模型中的数据交换将其耦合到一个vPower模型。使用此代码包，通过降低一次流速或引入一个模块的负反应性，模拟了两个典型的瞬态现象。研究了重要参数，包括反应堆功率，燃料温度以及两个模块的反应堆入口和出口氦气温度。计算结果初步证明，该代码包可进一步用于评估HTR-PM的工作性能。