Development of Fast Helium Cooled Reactors in Russia
Development of Fast Helium Cooled Reactors in Russia
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发表时间：
2003
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通讯作者：
N. Ponomarev-stepnoi;S. Glushkov
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作者：
N. Ponomarev-stepnoi;S. Glushkov
21世纪核能的特征将在于在其结构中使用快堆，其中核燃料增殖沿着热反应堆来完成。高温气冷堆（HTGR）和快中子堆（BGR）的联合使用可能被证明是解决未来核能燃料供应问题的良好解决方案之一。这种反应堆出口处的高氦气温度允许发电和使用热量用于各种过程，例如制氢。本文介绍了分析的结果，努力发展的快速氦冷却反应堆的概念，以前在俄罗斯进行。论证了快氦冷堆（BGR）相对于快液态金属冷堆的优越性。各种BGR的概念进行了分析。其中一个概念是通过使用一个模块式反应堆（钚堆芯反应堆）来实现最大增殖比，该反应堆有一个含有239钚的小堆芯，但没有增殖材料。在第二个概念中，在保持燃料增殖参数的同时增加反应堆功率是在具有多钚堆芯的反应堆中实现的，该反应堆基于在共同的铀包层内以一定的周期性结构放置多个钚堆芯。在第三个概念中，通过使用238 U稀释的钚（混合氧化物燃料）增加堆芯体积来增加反应堆功率。
Nuclear energy of the 21 century will be characterized by the use in its structure of fast reactors wherein nuclear fuel breeding is accomplished along with thermal reactors. A combined use of high-temperature gas cooled reactors on thermal (HTGR) and fast (BGR) neutrons may prove to be one of good solutions to the problem of providing fuel for the future nuclear energy. The high helium temperature at the outlet of such reactors allows both electricity generation and using heat for various processes, such as hydrogen production. The paper presents results of the analysis of efforts on development of fast helium cooled reactor concepts previously undertaken in Russia. Advantages of fast helium cooled reactors (BGR) over fast liquid metal cooled reactors are demonstrated. Various BGR concepts are analyzed. One of the concepts consists in attaining the maximum breeding ratio through the use of a modular reactor with a small core containing 239 Pu without breeder material (the plutonium core reactor). In the second concept, an increase in the reactor power while maintaining the fuel breeding parameters is accomplished in a reactor with a multiplutonium core based on placement of several plutonium cores in a certain periodic structure inside a common uranium blanket. In the third concept, the reactor power is increased through an increase of the core volume using plutonium diluted with 238 U (MOX fuel).