[en] A continuous-energy sensitivity coefficient capability has been introduced into MCNP6. The methods for generating energy-resolved and energy-integrated sensitivity profiles are discussed. Results from the verification exercises that were performed are given, and these show that MCNP6 compares favorably with analytic solutions, direct density perturbations, and comparisons to TSUNAMI-3D and MONK. Run-time and memory requirements are assessed for typical applications, and these are shown to be reasonable with modern computing resources. (authors)

[en] A research version of the Monte Carlo software package MCNP6 is modified to incorporate advection and diffusion of delayed neutron precursors, resulting in the emission of delayed neutrons at locations different from the original fission sites. Results of two test problems, a pipe carrying flowing fissile solution and a sphere of fissile solution with precursor diffusion, show that the fission product mobility tends to perturb the fundamental mode, has a negative reactivity effect, and, perhaps most importantly, causes a decrease in the effective delayed neutron fraction. (authors)

[en] MCNP6 has the capability to produce energy-resolved sensitivity profiles for secondary distributions (fission Χ and scattering laws). Computing both unconstrained and constrained profiles are possible. Verification is performed with analytic test problems and a comparison to TSUNAMI-3D, and the comparisons show MCNP6 calculates correct or consistent results. Continuous-energy calculations are performed for three fast critical experiments: Jezebel, Flattop, and copper-reflected Zeus. The sensitivities to the secondary distributions (integrated over chosen energy ranges) are of similar magnitude to those of many of the cross sections, demonstrating the possibility that integral experiments are useful for assessing the fidelity of these data as well. (authors)

[en] Traditionally, Monte Carlo radiation transport codes have been unable to tally adjoint-weighted quantities in a robust and consistent way. A new framework based on the iterated fission probability interpretation of the adjoint flux is developed that solves this problem at minimal increase in CPU time. A derivation of adjoint-weighted tallies is presented along with a methodology for computing adjoint-weighted reactor kinetics parameters and first-order reactivity perturbations. Details are given for the implementation in the production MCNP code. The calculations are benchmarked to both experimental measurements and calculations with the discrete ordinates code, Partisn. (authors)

[en] The variances estimated by Monte Carlo codes in k-eigenvalue calculations are under-predicted due to inter-cycle correlation between fission sources. The mutual information serves as a diagnostic to measure the correlation between fission source distributions in different cycles. There is a definite observed relationship between the variance bias and the mutual information of the source distributions. Furthermore, using the mutual information in conjunction with the Wielandt method shows how effective a particular Wielandt shift is at removing variance bias. Finally, the dominance ratio and the mutual information are related to MacMillan's correction. (authors)

[en] The ENDF71x library [1] is the most thoroughly tested set of ACE-format data tables ever released by the Nuclear Data Team at Los Alamos National Laboratory (LANL). It is based on ENDF/B-VII. 1, the most recently released set of evaluated nuclear data files produced by the US Cross Section Evaluation Working Group (CSEWG). A variety of techniques were used to test and verify the ENDF7 1x library before its public release. These include the use of automated checking codes written by members of the Nuclear Data Team, visual inspections of key neutron data, MCNP6 calculations designed to test data for every included combination of isotope and temperature as comprehensively as possible, and direct comparisons between ENDF71x and previous ACE library releases. Visual inspection of some of the most important neutron data revealed energy balance problems and unphysical discontinuities in the cross sections for some nuclides. Doppler broadening of the total cross sections with increasing temperature was found to be qualitatively correct. Test calculations performed using MCNP prompted two modifications to the MCNP6 source code and also exposed bad secondary neutron yields for ^231,233Pa that are present in both ENDF/B-VII.1 and ENDF/B-VII.0. A comparison of ENDF71x with its predecessor ACE library, ENDF70, showed that dramatic changes have been made in the neutron cross section data for a number of isotopes between ENDF/B-VII.0 and ENDF/B-VII.1. Based on the results of these verification tests and the validation tests performed by Kahler, et al. [2], the ENDF71x library is recommended for use in all Monte Carlo applications. (authors)

[en] The element F_ij of the fission matrix is equal to the number of fission neutrons born in region i due to one average fission neutron starting in region j. The fission matrix is a spatially discretized Green's function for the next generation fission neutron source. We describe recent experience and results from implementing a fission matrix capability into the MCNP Monte Carlo code. The fission matrix can be used to provide estimates of the fundamental mode fission distribution, the dominance ratio, the eigenvalue spectrum, and higher mode forward and adjoint eigenfunctions of the fission neutron source distribution. It can also be used to accelerate the convergence of the power method iterations and to provide basis functions for higher-order perturbation theory. The higher-mode fission sources can be used in MCNP to determine higher-mode forward fluxes and tallies, and work is underway to provide higher-mode adjoint-weighted fluxes and tallies. Past difficulties and limitations of the fission matrix approach are overcome with a new sparse representation of the matrix, permitting much larger and more accurate fission matrix representations. The new fission matrix capabilities provide a significant advance in the state-of-the-art for Monte Carlo criticality calculations

[en] Fusion systems are, in general, geometrically complex requiring detailed three-dimensional (3-D) nuclear analysis. This analysis is required to address tritium self-sufficiency, nuclear heating, radiation damage, shielding, and radiation streaming issues. To facilitate such calculations, we developed an innovative computational tool that is based on the continuous energy Monte Carlo code MCNP and permits the direct use of CAD-based solid models in the ray-tracing. This allows performing the neutronics calculations in a model that preserves the geometrical details without any simplification, eliminates possible human error in modeling the geometry for MCNP, and allows faster design iterations. In addition to improving the work flow for simulating complex 3- D geometries, it allows a richer representation of the geometry compared to the standard 2nd order polynomial representation. This newly developed tool has been successfully tested for a detailed 40 degree sector benchmark of the International Thermonuclear Experimental Reactor (ITER). The calculations included determining the poloidal variation of the neutron wall loading, flux and nuclear heating in the divertor components, nuclear heating in toroidal field coils, and radiation streaming in the mid-plane port. The tool has been applied to perform 3-D nuclear analysis for several fusion designs including the ARIES Compact Stellarator (ARIES-CS), the High Average Power Laser (HAPL) inertial fusion power plant, and ITER first wall/shield (FWS) modules. The ARIES-CS stellarator has a first wall shape and a plasma profile that varies toroidally within each field period compared to the uniform toroidal shape in tokamaks. Such variation cannot be modeled analytically in the standard MCNP code. The impact of the complex helical geometry and the non-uniform blanket and divertor on the overall tritium breeding ratio and total nuclear heating was determined. In addition, we calculated the neutron wall loading variation in both the poloidal and toroidal directions. The final optics system of the HAPL power plant includes several metallic and dielectric mirrors that are sensitive to radiation. Although some of these mirrors are not in the direct line-of-sight of the neutron source, radiation scattering and streaming through the laser beam ports requires an assessment of the nuclear environment at the final optics to predict their lifetime. Detailed CAD models of the ITER FWS modules were analyzed to produce high resolution maps of nuclear heating, radiation damage and helium production. These clearly show the impact of the design heterogeneity details with the many coolant channels embedded in the module. In addition, hot spots produced in the vacuum vessel behind the module as a result of streaming through these coolant channels were evaluated. These examples will be presented to demonstrate the applicability of the tool to nuclear analysis of complex fusion systems

[en] The recent development of the compact stellarator concept delivered ARIES-CS - a compact stellarator with 7.75 m average major radius, approaching that of tokamaks. In stellarators, the most influential engineering parameter that determines the machine size and cost is the minimum distance between the plasma boundary and mid-coil (Δ_min). Accommodating the breeding blanket and necessary shield within this distance to protect the superconducting magnet represents a challenging task. Selecting the nuclear and engineering parameters to produce an economic optimum, modeling the complex geometry for 3-D nuclear analysis to confirm the key engineering parameters, and minimizing the radwaste stream received considerable attention during the ARIES-CS design process. This paper provides a perspective on the successful integration of the nuclear activity, economics, and safety constraints into the final ARIES-CS configuration. (author)

[en] Epithermal neutron elastic scattering can be significantly affected by the thermal motion of target nuclides. Since the 1950's continuous-energy Monte Carlo codes have generally accounted for the target motion using a free gas scattering model, with the assumption that the scattering cross-section is constant in energy. Recent work has shown the importance of resonance scattering, and several methods for an improved free-gas treatment have been developed. We have implemented a rejection-based sampling scheme in the MCNP free-gas treatment to account for cross-section variation. The modified MCNP code was used to investigate a number of practical concerns: results for an LWR Doppler defect benchmark; computational costs; and energy limits for the free-gas treatment. Additionally, the impact on a suite of ICSBEP criticality benchmark problems (at room temperature) was determined to be negligible, an important result since such problems are used extensively in testing and evaluating revisions to ENDF/B-VII nuclear data. (authors)