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Here, the Chebyshev Rational Approximation Method (CRAM) algorithm has become the method of choice for modern nuclear depletion calculations. A simple modification of the CRAM yields not only depleted nuclide amounts but also time-averaged nuclide amounts, with little additional calculation cost.
This presentation begins by providing an overview of the PERT card. The PERT card uses differential operator method to compute first- and second-order tally variations due to density, composition, and reaction cross-sections. It is possible to have multiple PERT cards in one MCNP input deck to study tally variations for several sets of nuclides, reactions, and energy ranges. Furthermore, the METHOD option tells MCNP to calculate either the perturbed tally (METHOD=-1, -2, -3) or the change in the unperturbed tally (METHOD=1, 2, 3). In summation, a powerful use-case for the MCNP code PERT card is that it facilitates calculating tally sensitivities to nuclear data. Writing PERT card entries and parsing output MCTAL files is tedious and error prone. however, Easy_PERT makes use of existing tools (Faust and MCNPTools) to handle writing PERT card entries and parsing the output MCTAL files. The PERT card is early in the development process and planned upcoming capabilities include calculating sensitivities and combining MCTAL files from separate runs into one JSON file.
The Taylor series (differential operator) perturbation method, as implemented in the PERT capability in MCNP6 (Ref. 1), can be used for first- and second-order sensitivity analyses in fixed-source problems. Unfortunately, the PERT capability does not yet work with point-detector (F5) or pulse-height (F8) tallies. Also, the second-order Taylor series term lacks the cross terms that would allow the calculation of mixed sensitivities. Finally, the PERT capability does not compute any sensitivities involving the source in a fixed-source problem. When neutrons or gamma rays are emitted due to radioactive decay, the source rate density depends intrinsically on the nuclide densities in the source material. When the PERT capability in MCNP6 is used to compute sensitivities of a response to nuclide and material densities, those sensitivities do not include the effect of the source.
The relative sensitivity of k eff to the densities of nuclides in a material are combined to compute relative sensitivities to other inputs. When computed in a single Monte Carlo run, the nuclide density sensitivities are correlated, and the statistical uncertainties propagated to other inputs will be incorrect unless those correlations are accounted for. Equations are presented to apply correlated sampling using batch statistics for the sum of an arbitrary number of random tallies and the difference of two random tallies when each is multiplied by a different constant. When correlated sampling is used on a recent benchmark evaluation, the correct statistical uncertainties for certain combinations of sensitivities are dramatically smaller than the incorrect uncertainties. A user-controlled, regular output of MCNP6’s KSEN sensitivities that allows batch statistics to be applied to combinations would be extremely valuable.
Monte Carlo simulations are the gold standard in particle transport simulations, for shielding and nuclear reactor simulations Full core simulations can require terabytes of memory for tracking pin-resolved (up to 50k pins), nuclide (200 tracked), axial (20-200 zones) reaction rates Domain decomposition, splitting the calculation and the memory storage, among several nodes is challenging as it involves communication of particles across domain boundaries We created a new MOOSE application, MaCaw, to study this problem Algorithms developed for ray tracing in MOOSE were adapted for simulating neutral particle transport Physics and tally routines in OpenMC, dynamically linked, are called from the application
For measuring a waste drum containing multi-nuclide radioactive material calorimetric assay is often combined with gamma ray and neutron measurements. To reduce the methodological uncertainties of the inventory determination in a multi-nuclide matrix, the escaping radiation must be quantized as much as the encapsulated gamma-rays and α-, β-, n- particles depositing radiological heat inside the drum. To differentiate the possible variety of heat sources in a radioactive waste drum, Monte Carlo simulations are utilized resembling the waste drum and the calorimeter in both, their geometries as well as the material compositions. MCNP6-based numerical imaging of the Large Volume Calorimeter (LVC) designed by KEP Nuclear(France) allows simulating various scenarios for different particles types, energies and source locations to accompany and support the interpretation of calorimetric KEP-LVC measurements anticipated within the CHANCE project. For the analyses, the particle flux and energy deposition in each layer of the calorimeter are displayed. The results yield that a significant part of the emitted radiation leaves the system and does not contribute to the heat deposited inside the drum. As an example, the particle flux study of a 200 L mock-up drum placed inside the KEP-LVC calorimeter is discussed. A tally 4 with mesh option for both gamma and neutrons up to 5 MeV simulates the mean flux in each cell and for various source location and source distribution patterns. The expected energy deposition is cumulated for each layer, and over the whole energy range it reveals that actually a small but well known fraction of particles escapes the LVC calorimeter. This makes it a very suitable apparatus for the anticipated experiments on large and heterogeneous waste drums. The results suggest that the most interesting cases would be polyethylene, bitumen and concrete matrices [3, 5]. However, it must be kept in mind that the high-energy part of the gamma and neutron radiation flux can reach the reference chamber of the calorimeter and deposit some energy there, which may compromise the calibration and cause double-bias. The simulations showed that, depending on source location and power, neutron and high-energy gamma radiation leakage can influence the heat measurements differently. In general, part of the simulation results suggest that for waste drums with a high emission rate the accuracy of calorimetric assay can be decreased due to radiation escape. (authors)
Here, the Comprehensive Nuclear-Test-Ban Treaty bans all nuclear tests worldwide. As part of treaty compliance, the concentration of radioactive nuclides in the atmosphere is monitored to detect nuclear weapons tests. Radioactive noble gas fission products, specifically radioxenon, can vent into the atmosphere after a nuclear weapons test, even if the test is well contained underground or underwater. Radioxenon thus serves as a signal for nuclear weapons tests. All atmospheric monitoring systems require samples of radioxenon isotopes for detector calibration, quality control, and certification. Here, we present a novel, improved method for creating samples of 133m Xe via neutron irradiation of 132 Xe in the Washington State University TRIGA reactor. 132 Xe neutron absorption results in either 133 Xe or 133m Xe—thermal neutron absorption results in 133m Xe 12% of the time, while fast neutron absorption (above ~1 MeV) results in 133m Xe ~50% of the time. To optimize the production of 133m Xe via neutron absorption in 132 Xe in the thermal TRIGA reactor, spectral tuning using an irradiation chamber is required to maximize the fraction of fast neutrons being absorbed and minimize the number of thermal neutrons interacting with the 132 Xe. We used MCNP to tally 132 Xe absorptions with the isotopic tally function, flux tallies and neutron attenuation to estimate the number of neutrons reaching the 132 Xe through the irradiation chamber, and the adjoint importance function to improve the source strength estimate. Additionally, we performed a heat transfer analysis for safety considerations. It was determined that the use of a 96% enriched 10 B boron carbide chamber, placed next to the fuel elements in reactor position D8, increases the 133m Xe/ 133 Xe activity ratio from a baseline value of 0.3 to 1.0, a 233% increase. Additionally, it was determined that the alpha heating produced in the boron does not become an unmanageable problem in the Washington State University reactor.