海洋核动力平台堆芯子通道分析

针对海洋核动力平台的堆芯结构和组件形式，使用成熟的子通道分析程序COBRA验证了堆芯稳态热工的安全性。通过计算得出，14.8 MPa压力下堆芯稳态最小烧毁比（DNBR）为2.342，燃料棒包壳表面最高温度为342 ℃，芯块中心最高温度为1 545 ℃。计算结果表明，改进后堆芯热工特性能满足当代反应堆安全性要求，并为海洋不利条件的影响留有足够的安全裕量。同时自主开发了计算机子通道分析程序，与COBRA程序的计算结果进行对比验证，两种计算方法的计算结果一致，从一定程度上说明了计算结果的可靠性。通过以上分析过程证明了燃料组件在稳态下的热工特性是安全和可靠的。

Sub-channel Analysis of Reactor Core for Marine Nuclear Power Platform

For structures and assembly forms of the nuclear reactor core used in marine nuclear power platform ,COBRA which is a proved sub‐channel analysis code was used to verify the thermal‐hydraulic security characteristics of the reactor core . When the reactor works at the pressure of 14.8 MPa ,the calculated results of the steady state show that the minimum value of departure from nucleate boiling ratio (DNBR) is 2.342 , the highest surface temperature of fuel rod cladding is about 342 ℃ ,and the highest temperature of the center for pellet is about 1 545 ℃ .It is show n that thermal‐hydraulic characteristics of the reactor can meet the requirements of security criterion and leave enough safety allowance for the impact of complicated marine conditions .At the same time ,a sub‐channel analysis code was developed to check the COBRA code results and it is ensured that the results are enough reliable and accurate . Through comparing the results with each other ,the good agreements of the comparison proved in some degree that the results are reliable .According to above calculations ,we can draw a conclusion that the steady thermal‐hydraulic characteristics of the reactor assembly are safe and the results are reliable .