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The target segments of the Oak Ridge National Laboratory Second Target Station (STS) neutron production facility become highly activated due to spallation reactions or nuclei transmutation by primary protons and emitted neutrons. Once the target segments are removed from their location within the core vessel, decay dose rates must be accurately quantified to determine the shielding configurations of remote-handling tools and transport casks and to aid in planning maintenance activities. For this analysis, we utilized a hybrid unstructured mesh (UM)/constructive solid geometry approach for calculating spallation products and neutron fluxes, activation calculations using the AARE package that includes the CINDER2008 activation code to calculate the decay photon source at different cooling times, and the ADVANTG code to accelerate the final decay photon transport calculation. Both Type 316 stainless steel (SS-316) and lead were investigated as candidates for shielding materials. The decay photon transport calculation through the thick SS-316 or lead shields exhibited between 25 and 30 orders-of-magnitude attenuations in the radial direction, depending on the shield. Such a difficult shielding calculation required advanced variance reduction. ADVANTG has some missing features, which limits its usability in spallation neutron source applications. It does not support volumetric sources created for MCNP6.2 UM capability. An approximate source was created for this problem. Not only was this approximate source needed for running the ADVANTG calculation to generate the weight windows, but also it was essential to develop source biasing (SB) parameters that were crucial for dramatically accelerating the decay photon transport in this problem. With this approximate source, the analysis was completed in a very reasonable computational time, and the design of the STS remote-handling equipment was finalized. This paper compares the efficiency of Monte Carlo simulations with different weight window and SB parameters calculated using different approximate ADVANTG calculations. [ABSTRACT FROM AUTHOR]

In order to model the activated isotopes and resulting dose from a nuclear detonation in an urban environment, the Activation and Transmutation of Isotopes in an Unstructured Mesh (ACTIUM) Python toolkit has been developed to combine the unstructured mesh–based particle transport capability of MCNP6.2 with the CINDER2008 transmutation code to produce quantities of interest for the post-detonation nuclear forensics and weapons effects communities. The ACTIUM toolkit has been implemented and validated with a number of test cases from a simple analytic model to a case study of the urban detonation in Nagasaki, Japan. The ACTIUM approach is the first of its kind to couple the latest release of CINDER2008 as a part of the Activation in Accelerator Radiation Environments (AARE) package with MCNP6.2 and produce transmuted quantities per time step on an unstructured mesh for the nuclear forensics and weapon effects communities. ACTIUM uses the latest ENDF/B-VIII.0, TENDL2017, and JENDL4 cross-section libraries for the transmutation calculations and includes methods for producing material cards for the initial MCNP6.2 unstructured mesh calculation based on highly detailed materials often found in urban environments on a city-specific basis. [ABSTRACT FROM AUTHOR]