We present a systematic and detailed approach to simulate the University of Utah TRIGA Reactor (UTR) in support of core criticality calculations to profile the entire reactor in detail. In performing this work, we utilized both a 3-D Cartesian S_N deterministic code, PENTRAN, and a Monte Carlo code, MCNP6, to calculate complimentary, high accuracy 3-D transport derived neutron flux distributions at reactor full power. For deterministic models, our study shows that a 14-group CONTRIBUTON weighted multi-group cross-section library with up-scattering compared extremely well with continuous energy Monte Carlo results. We also modeled results to confirm the activity of an activated arsenic sample placed in the heavy water moderated thermal irradiation chamber of the UUTR (University of Utah TRIGA Reactor).

A 138 pcm difference in the system eigenvalue for the full 3-D core PENTRAN discrete ordinates (S_N) UUTR model utilizing 2$\times$10⁹ equations, in 14 energy groups, is compared to the reference case of a similar full core MCNP6 Monte Carlo calculation with continuous energy cross-sections. On average, deterministic and Monte Carlo core neutron group flux values differ by less than 1%, with some local maximum relative differences between 10% and 20% for the 3-D core flux distribution in some energy groups are observed. The differences of neutron fluxes from group 3/14 to group 10/14, with energies spanning from 1.11 eV to 13.8 MeV, are comparable to within 5% in the UUTR core active fuel region. For energy groups with energies greater than 13.8 MeV, MCNP6 is not able to resolve the flux distributions with reasonable statistical uncertainties due to the extremely low sampling probability for neutrons in those energies; orders of magnitude more histories would be required to do so. In contrast, the PENTRAN 3D S_N calculation shows a very detailed flux distribution for energies larger than 13.8 MeV, highlighting the complimentary utility of applying both methods. Detailed approaches for the PENTRAN calculation, including energy group collapsing, homogenization of fuel + gap + cladding, and analysis are presented in the narrative.

我们提供了一种系统且详细的方法来模拟犹他大学的TRIGA反应堆（UTR），以支持堆芯临界计算以详细描述整个反应堆。在执行这项工作时，我们利用了3-D笛卡尔S _N确定性代码PENTRAN和蒙特卡洛代码MCNP6来计算在反应堆全功率下互补的高精度3-D传输派生的中子通量分布。对于确定性模型，我们的研究表明，具有向上散射的14组CONTRIBUTON加权多组横截面库与连续能量蒙特卡洛结果非常好。我们还对结果进行建模，以确认放置在UUTR（犹他大学TRIGA反应堆）的重水缓和的热辐射室中的活化砷样品的活性。

使用2的完整3-D核心PENTRAN离散坐标（S _N）UUTR模型的系统特征值有138 pcm的差异$\times$10 ⁹将14个能量组中的方程与具有连续能量横截面的类似全核MCNP6蒙特卡洛计算的参考案例进行比较。平均而言，确定性和蒙特卡罗核心中子群通量值相差不到1％，对于某些能量组中的3-D核心通量分布，观察到一些局部最大相对差在10％和20％之间。能量范围从1.11 eV到13.8 MeV的3/14组到10/14组的中子通量差异在UUTR堆芯活性燃料区域中可比不到5％。对于能量大于13.8 MeV的能量组，由于中子在这些能量中的采样概率极低，因此MCNP6无法解决具有合理统计不确定性的通量分布。这样做需要更多的历史记录。相比之下，PENTRAN 3D S_N计算显示了大于13.8 MeV的能量的非常详细的通量分布，突出了应用这两种方法的互补效用。叙述中介绍了用于PENTRAN计算的详细方法，包括能量组崩溃，燃料+间隙+包层的均质化以及分析。