Small Nuclear Reactors for the Production of Nuclear Medicine Radioisotopes and for Co-Generation

该研究项目旨在开发两个截然不同的小型本质安全核反应堆：SNB25区域供热和热电联产反应堆和均质慢冲堆，并用确定性程序WIMS-AECL和概率(蒙特卡罗)程序MCNP 5对这两个反应堆进行了模拟，两个模拟的结果都非常吻合。SNB25反应堆以浓缩20%的Triso燃料颗粒为基础，通过95摄氏度的水产生25兆瓦的热量，这些热量被送往北极的加拿大军队基地的各种建筑中加热，例如拟建的努纳武特纳尼西维克基地。这些能量也可以供应给基于朗肯循环的系统，以16%的效率发电。之前的工作已经证明了该概念的固有安全特性和反应堆12年的最低核心寿命。对于北极等气候条件恶劣的地区的应用来说，这确实是一项重要的资产。这里拟议的工作旨在改进反应堆的计算机模拟，并通过改进的朗肯循环设计更高效的热传输系统和电力生产系统。提议项目中考虑的另一座核反应堆是基于成功的Slowpoke-2研究反应堆的均质慢冲反应堆。均质慢冲反应堆有双重任务：在现有慢冲-2研究反应堆的核心使用寿命结束时更换它们，并继续它们的研究/教学任务。并生产用于多种应用的放射性同位素，包括核医学中的重要应用。比如Slowpoke-2反应堆，均质反应堆产生20千瓦的热量，在以前的工作中已被证明是本质安全的。为了遵守《不扩散条约》，铀燃料的浓缩度被保持在最高20%。燃料是一种均相的硫酸铀酰水溶液(UO_2SO_4)，对于一个锆-2反应堆容器，其摩尔浓度被确定为1.46摩尔·公斤-1。这个池式反应堆的运行慢化-冷却剂温度为40摄氏度，对反应堆瞬变状态的模拟工作表明，在最坏的情况下，峰值温度和功率分别为830摄氏度和546千瓦。为这一研究项目提议的进一步开发工作包括用三维确定性软件取代二维WIMS-AECL模拟程序，开发考虑燃料在堆芯内运动的模拟程序(中子学和热工-水力学程序的耦合)，以及对反应堆堆芯内饱和裂变产物如135Xe的行为进行建模。

The research project aims at developing two distinct small inherently safe nuclear reactors: the SNB25 district heating and co-generation reactor and the Homogeneous SLOWPOKE reactor.  Both reactors are simulated with the deterministic code WIMS-AECL and the probabilistic (Monte Carlo) code MCNP 5, with the results of both simulations agreeing very well. The SNB25 reactor is based on the TRISO fuel particles with an enrichment of 20% and produces 25 MW of heat by means of water at 95oC that is sent for heating the various buildings of a Canadian Forces base in the Arctic, such as the proposed base at Nanisivik, Nunavut.  This energy can also be supplied to a Rankine cycle-based system to produce electricity with a 16% efficiency.  Previous work has demonstrated the inherent safety characteristics of the concept and the 12-year minimum core life of the reactor, which indeed represents an important asset for applications in areas with harsh weather conditions such as in the Arctic.  The proposed work here aims at improving the computer modelling of the reactor and at designing a more efficient heat transport system and the electricity production system via an improved Rankine cycle.  The other nuclear reactor considered in the proposed project is the Homogeneous SLOWPOKE reactor, based on the successful SLOWPOKE-2 research reactor.  The Homogeneous SLOWPOKE reactor has a double mission: to replace the existing SLOWPOKE-2 research reactors at the end of their core service life and continue their research/teaching mission, and to produce radioisotopes used in several applications including important ones in nuclear medicine.  Like the SLOWPOKE-2 reactor, the Homogeneous reactor produces 20 kW of heat and has been proven as inherently safe in previous work.  The enrichment of the uranium fuel is kept to a maximum of 20% to comply with the Non Proliferation Treaty.  The fuel is an homogeneous aqueous solution of uranyl sulfate (UO2SO4) with a molality determined as 1.46 mol kg-1 for a Zircaloy-2 reactor vessel.  This pool-type nuclear reactor has an operating moderator-coolant temperature of 40oC and simulation work of the reactor in transient states has shown that the worst scenarios result in peak temperatures and power of 83oC and 546 kW, respectively. Further development work proposed for this research project includes replacing the 2-D WIMS-AECL simulation code with a 3-D deterministic software, developing a simulation code accounting for the movement of the fuel within the reactor core (coupling of neutronics and thermal-hydraulics codes), and modelling the behaviour within the reactor core of saturating fission products such as 135Xe .