Validation of Light Water Reactor Ex-Core Calculations with VERA

Abstract

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.

Keywords:

• Reactor ex-core simulation
• VERA
• CADIS
• light water reactor analysis
• fluence calculations

Acknowledgments

The authors would like to acknowledge Herschel Smith and Duke Energy for providing the Shearon Harris MCNP inputs that enabled the code-to-code comparison with VERA, and also Tennessee Valley Authority and Westinghouse for their support in developing the Watts Bar models. The authors also thank all the VERA developers and infrastructure teams for their invaluable support during this work.

This manuscript has been authored by UT-Battelle, LLC, under contract DE-AC05-00OR22725 with the U.S. Department of Energy (DOE). The U.S. government retains and the publisher, by accepting the paper for publication, acknowledges that the U.S. government retains a nonexclusive, paid-up, irrevocable, worldwide license to publish or reproduce the published form of this manuscript, or allow others to do so, for U.S. government purposes. DOE will provide public access to these results of federally sponsored research in accordance with the DOE Public Access Plan (http://energy.gov/downloads/doe-public-access-plan).

Research was sponsored by the Laboratory Directed Research and Development Program of Oak Ridge National Laboratory, managed by UT-Battelle LLC for the DOE and by CASL (www.casl.gov), an Energy Innovation Hub (http://www.energy.gov/hubs) for Modeling and Simulation of Nuclear Reactors under DOE contract no. DE-AC05-00OR22725. This research used resources of the Oak Ridge Leadership Computing Facility, which is a DOE Office of Science User Facility supported under contract DE-AC05-00OR22725.

This research made use of Idaho National Laboratory computing resources that are supported by the Office of Nuclear Energy of the DOE and the Nuclear Science User Facilities under contract no. DE-AC07-05ID14517.

Notes

^a Coupled neutron-photon calculations are performed by MPACT for the in-core VERA calculation, however, only the neutron fission source is passed to Shift for the follow-on fixed-source calculation. During the fixed-source calculation, coupled neutron-photon calculations can also be performed by Shift. For this work, only the results from neutron transport calculations are presented.

^b The full geometry refers to the (1) in-core geometry for the fuel, assemblies, and upper and lower plates as modeled by the reactor toolkit (RTK) geometry native to VERA for modeling LWRs, and (2) ex-core geometry is constructed using combinatorial geometry with Omnibus General Geometry (GG), which is native to Shift. The transport calculation is performed on the entire geometry, which is an amalgamation of the RTK in-core and GG ex-core models. The hybrid calculation performed by Denovo is performed on the entire geometry, which is discretized using a Cartesian grid based on the materials and surface boundaries.