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The Harris Nuclear Plant reactor model includes pin-cell, baffles, core barrel, neutron pads, downcomer, reactor vessel, concrete and detector.
Source publication
Previous work focused on the use of automated variance reduction parameters for use in a fixed source reactor shielding problem to solve for ex-core particle transport results. The current work assesses that strategy in light of a fully-coupled Monte Carlo problem run as a critical system with parallel processing and the use of moderate variance re...
Contexts in source publication
Context 1
... MCNP model consists of fuel, cladding, guide tubes, baffles, core barrel, neutron pads, moderator downcomer, reactor pressure vessel, concrete shield wall, and ex-core neutron detector, all shown in Figure 1. ...
Context 2
... concrete shield wall is also a necessary feature of the problem geometry since it has been determined that a significant portion of the neutron detection occurs because of thermalization and backscatter from the concrete. In Figure 1 above, it can be seen that the ex-core neutron detector sits inside a pipe (which is included in this model), and the geometric divisions in the problem exist for particle splitting to aid variance reduction. Finally, the "forced collisions" variance reduction feature is used in the concrete to aid problem convergence. ...
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Citations
... A study of the effect of these three different neutron source spectra sampling schemes (discussed in Section 2) on the power range detector relative 10 B response was performed on the Shearon Harris (Harris) model, previously published in [8]. Figure 1a shows the in-and ex-core geometry. ...
... The relative detector responses from these cases are all shown in Fig. 2, where the results generated with the modified geometry are denoted in the legend with M od.P ad. Note that a comparison of the absolute values of the detector response from MCNP and Shift (standalone), generated with the original Harris geometry, were done previously in [8] and are not presented here. The absolute values from both the codes compared well giving further confidence in the validity of the relative detector response being calculated by Shift. Figure 2 shows that the VERA results generated with the 235 U spectrum and nuclide-dependent Watt spectra sampling schemes are in good agreement with the MCNP results. ...
The Consortium for Advanced Simulation of Light Water Reactors (CASL) Virtual Environment for Reactor Applications (VERA) offers unique capabilities to combine highfidelity in-core radiation transport with temperature feedback using MPACT and CTF with a follow-on fixed source transport calculation using the Shift Monte Carlo code to calculate ex-core quantities of interest. In these coupled calculations, MPACT provides a fission source to Shift for the follow-on radiation transport calculation. In past VERA releases, MPACT passed a spatially dependent source without the energy distribution to Shift. Shift then assumed a ²³⁵ U Watt spectrum to sample the neutron source energies. There were concerns that, in cases with burned or mixed oxide (MOX) fuel near the periphery of the core, the assumption of a ²³⁵ U Watt spectrum for the source neutron energies would not be accurate for studying ex-core quantities of interest, such as pressure vessel fluence or detector response. Therefore, two additional options were implemented in VERA for Shift to sample neutron source energies: (1) a nuclide-dependent Watt spectra for ²³⁵ U, ²³⁸ U, ²³⁹ Pu, and ²⁴¹ Pu, and (2) to use the standard 51-energy group MPACT spectrum. Results show that the 51-group MPACT spectrum is not suitable for ex-core calculations because the groups have been fine-tuned for in-core calculations. Differences in relative detector response due to ²³⁵ U and nuclide-dependent Watt spectra sampling schemes were negligible; however, the use of nuclide-dependent Watt spectra for vessel fluence calculations was found to be important for fuel cycles with burned and fresh fuel.
... A 5 cm radius spherical tally region containing air was added to the excore region, as shown in green in Fig. 9. This tally was modified with 10 B total reaction cross section to simulate a BF 3 neutron detector (Smith et al., 2018). Since there are no streaming channels in this problem, it more closely resembles the diffusion base case. ...
Monte Carlo (MC) shielding calculations often use weight windows (WWs) and biased sources formed from a deterministic estimate of the adjoint flux to improve the convergence rate of tallies. This requires a significant amount of computer memory, which can limit the memory available for high-resolution tally output. A new method is proposed for reducing these memory requirements by using singular value decomposition (SVD) in linear or logarithmic space to approximate the adjoint flux. This method's performance is evaluated using the Shift and Denovo codes for streaming and diffusion base case problems, followed by problems using the Westinghouse AP1000 and the Joint European Torus. The log SVD reduced WW memory requirements by an order of magnitude in all cases without a significant performance penalty. Additionally, the linear SVD reduced biased source memory requirements by an order of magnitude, but further investigation is needed to account for observed limitations.
The Consortium for Advanced Simulation of Light Water Reactors (CASL) Virtual Environment for Reactor Applications (VERA) is a reactor simulation software. It offers unique capabilities by combining high-fidelity in-core radiation transport with temperature feedback by using MPACT (a deterministic neutron transport code) and COBRA-TF (a thermal-hydraulic code) with follow-on, fixed-source transport calculations using the Shift Monte Carlo code to calculate ex-core quantities of interest. In these coupled calculations, MPACT provides Shift with the fission source for follow-on ex-core calculations. These ex-core simulations can be set up to calculate detector responses, as well as the flux and fluence in ex-core regions of interest, such as the reactor pressure vessel, nozzle, and irradiated capsules. A Watts Bar Nuclear Plant Unit 1 (WBN1) ex-core model was developed, as described in this paper, and this model was used to perform coupon calculations. The results for the coupon flux calculations show close agreement with the reference values for cycle 1 produced by the two-dimensional Discrete Ordinates Transport (DORT) code and presented in a BWXT Services Inc. report. However, differences in the results (10%) seen in cycles 2 and 3 and the reasons for these differences are discussed in this paper. The VERA WBN1 model was also used to perform a vessel fluence calculation for cycle 1. Additionally, a collaboration between CASL and Duke Energy led to the first code-to-code validation of VERA for reactor ex-core applications that used a model for the Shearon Harris reactor. Results from this collaboration show excellent agreement between VERA and the Monte Carlo N-Particle Transport Code for the detector response calculations. The work performed under this collaboration is also detailed in this paper.
The private sector’s recent interest in the active development of molten salt reactors has led to the need to develop and test advanced modeling and simulation tools to analyze various advanced reactor types under numerous conditions. This paper discusses the effort undertaken to model the Oak Ridge National Laboratory (ORNL) Molten Salt Breeder Reactor (MSBR) design using ORNL’s Shift Monte Carlo code. The MSBR model integrates a Monte Carlo N-Particle (MCNP) MSBR core model with an MCNP model that was generated from a CAD model of the external components and the reactor building, which was subsequently run in Shift. This paper focuses on development of the fully integrated model and its use in performing neutron transport calculations in the reactor cell area. This model is intended to aid in understanding radiological dose conditions during operation, as well as the iron dpa rates in the reactor vessel. The neutron biological dose rates and flux calculated in the reactor cell are much higher in the MSBR than in typical light-water reactors. The implications of these results and future work are also discussed in this paper.
