Neutron flux calculation for central channel in first cycle of SPRR-300
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摘要: 基于MCNP程序对300#研究堆首炉堆芯进行精细建模,通过并行计算方式得到了实验临界棒位下堆芯的有效增殖因数为1.002 29,与临界值之间的相对误差为0.229%,验证了物理模型的正确性。探讨并解决了并行计算的中断与接续问题,提出了体通量计数与点探测器计数应用中的合理化建议,即对大体积空间计数时尽量使用体通量计数。计算值与实验值对比结果表明:两者在3 MW功率水平下热中子通量密度相差4.6%,符合得较好。Abstract: The physical model of the 300# swimming pool research reactor(SPRR-300) based on the Monte Carlo code MCNP has been verified. Sophisticated modeling is conducted. An effective multiplication factor value of 1.002 29 is obtained, existing a relative error of 0.229% compared with the critical value. Meanwhile, a problem comes out that the interrupt and continue-run with parallel version of MCNP doesnt work. The problem is solved through trail and error process. A reasonable application of flux tally average over a cell and flux tally at a point is suggested, namely the former is prior to the latter to tally in big volume. Comparison between calculation results and experimental data shows that the thermal neutron flux has a deviation of 4.6% at a power level of 3 MW. That is to say, the calculated value and the experimental value agree well with each other, and the neutron flux result is dependable.
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Key words:
- neutron reactor /
- criticality calculation /
- MCNP code /
- parallel calculation /
- continue-run
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