Abstract
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 a235U 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 a235U 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 for235U,238U,239Pu, and241Pu, 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 to235U 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.
Highlights
The Consortium for Advanced Simulation of Light Water Reactors (CASL) Virtual Environment for Reactor Applications (VERA) [1] offers unique capabilities to combine high-fidelity in-core radiation transport with temperature feedback using the MPACT [2] deterministic solver and CTF [3] subchannel thermal hydraulics code with in-core and ex-core transport using the Shift [4] Monte Carlo (MC) code
The in-core geometry was set up using VERA common input [9], and the ex-core geometry was modeled using General Geometry (GG) with a supplemental file
For the study presented in this paper, a detailed analysis on the effect of the fission source energy spectra for ex-core detector calculations was performed using the Shearon Harris and Watts Bar Nuclear Plant Unit 1 (WBN1) VERA models
Summary
The Consortium for Advanced Simulation of Light Water Reactors (CASL) Virtual Environment for Reactor Applications (VERA) [1] offers unique capabilities to combine high-fidelity in-core radiation transport with temperature feedback using the MPACT [2] deterministic solver and CTF [3] subchannel thermal hydraulics code with in-core and ex-core transport using the Shift [4] Monte Carlo (MC) code. For the remainder of this paper, the focus is on the effect of the approximation used for the fission source spectra in Shift for the fixed-source MC calculation. In these coupled calculations, MPACT provides a fission source to Shift for a follow-on fixed source radiation transport calculation to determine ex-core quantities of interest. There were concerns that, in cases with burned fuel or mixed oxide (MOX) fuel near the core periphery, the assumption of a 235U Watt spectrum for the neutron source energies would not be accurate for studying excore quantities of interest such as pressure vessel fluence or detector response. An investigation of the neutron source energy spectra used by Shift in VERA was undertaken
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