Elsevier

Annals of Nuclear Energy

Volume 120, October 2018, Pages 178-185
Annals of Nuclear Energy

A stylized 3D Advanced High Temperature Reactor (AHTR) benchmark problem

https://doi.org/10.1016/j.anucene.2018.05.019Get rights and content

Abstract

A set of Stylized 3D Advanced High Temperature Reactor (AHTR) benchmark problems in full core and single fuel assembly configurations is developed in this paper. The configurations include the lower support plate, the bottom reflector, the fuel zone (AHTR assemblies), the top reflector, and the upper support plate. The benchmark problems retain the multiple heterogeneities and other important neutronics features such as the detailed geometric and material distributions of the Tristructural-Isotropic (TRISO) fuel and burnable poison particles, the fluoride salt coolant, the graphite moderator, and the reflectors. Monte Carlo results are presented for the benchmark problem in both the uncontrolled and controlled single assembly configurations. These benchmark problems can be used for evaluating the performance of neutronics codes.

Introduction

The Advanced High Temperature Reactor (AHTR) (Varma et al., 2012, and Holcomb et al., 2011) is a fluoride-salt-cooled high-temperature reactor (FHR) concept that provides inherent safety through passive safety systems and improved economics through higher operating temperatures. One of the challenges that remain before the commercialization and deployment of this class of reactors is the verification and validation (V&V) of neutronics tools and methodologies for design optimization and safety analysis in support of licensing this type of reactors. To verify neutronics tools, it is necessary to create new heterogeneous benchmark problems that retain the important neutronic characteristics of the AHTR such as the detailed geometric and material configurations of the Tristructural-Isotropic (TRISO) fuel and burnable poison particles, fluoride salt coolant, graphite moderator, and reflectors. However, the structural details and the randomness of the fuel particles distribution are simplified for ease of modeling and the lack of the capability of existing codes (including stochastic transport) to model random particle distributions in large systems (e.g., AHTR). These simplifications have relatively small neutronic effects as compared to that of the multiple heterogeneity. This stylized benchmark problem set facilitates numerical verification of deterministic transport and diffusion methods in which the multigroup approximation is a common practice, cross section sensitivity analysis, and evaluation of various approximations used in neutronic modeling at both lattice (fuel assembly) and core levels.

The paper is organized as follows. The assumptions and simplifications used to develop the stylized AHTR benchmark problems are summarized in Section 2. Then, the specification of the AHTR full core problems is presented in Section 3. The description of the assembly configuration and the reflectors as well as support plates is given in Sections 4 AHTR fuel assembly specification, 5 Specification of reflectors and support plates, respectively. Reference solutions for controlled and uncontrolled single assembly benchmark problems are presented in Section 6. The paper is summarized in Section 7. The isotopic composition of the materials in the core are given in the appendix.

Section snippets

Assumptions and simplifications

In developing any benchmark problem, it is desirable to simplify as much as possible without compromising the underlying physics (neutronic characteristics). The benchmark problem presented in the paper is based on the description of the AHTR conceptual design found in Varma et al., 2012, Holcomb et al., 2011. In addition to the usual simplifications, assumptions had to be made to fill gaps and resolve inconsistency in the data found in these references.

The assumptions and simplifications that

Core specification

The AHTR concept was designed to have a thermal power of 3400 MW with the overall thermal efficiency of 45%. The core parameters are shown in Table 1. The primary coolant used in the core is FLiBe (2LiF-BeF2). The TRISO particles with an enrichment of 9 wt% were used for the most recent AHTR design (Varma et al., 2012).

AHTR fuel assembly specification

As shown in Fig. 4, the AHTR fuel assembly has 120° rotation symmetry and is separated into three identical sectors by the 4 cm thick Y-shaped supporting structure which is made of C-C composite with a density of 1.95 g/cm3. Each sector consists of six fuel plates. The gap between two fuel plates is 0.7 cm while the gap between a fuel plate and the Y-shaped supporting structure is 0.35 cm. Each of those gaps, maintained by two semi-cylinder spacers, is filled with the primary coolant FLiBe. The

Reflectors

Reflectors used in the AHTR conceptual design can be classified in four types based on their location and geometric configurations as found below. Their geometrical configuations can be found in Fig. 7. The outer apothem of all reflectors is 23.4 cm.

  • Top reflector: The top reflector is immediately above the fuel assemblies as shown in Fig. 3. Like the fuel assembly, the top reflector consists of 18 plates, 42 spacers, a Y-shaped supporting structure, a hexagonal C-C channel box, a 0.9-cm thick

Solution to benchmark problems

MCNP5 was used to calculate the eigenvalue and the stripe-wise fission densities for the uncontrolled and controlled single assembly benchmark problems. The problems were run on a 40-core computer cluster with 80 million particle histories (80,000 particles in each cycle for 1000 cycles). The eigenvalue (keff) results for the two single assembly benchmark problems are provided in Tables 10, respectively. Since the assembly consisting of three identical sectors has the 120° rotation symmetry,

Summary

A 3D full core benchmark problem typical of fluoride-salt-cooled high-temperature reactor cores in prismatic configuration has been developed and fully described in this paper. The problem was derived from the AHTR concept by simplifying the geometry and material specification of the original conceptual design while retaining the multiple heterogeneities and major physical characteristics of the core that are important from the neutronics point of view. The fuel assembly and core descriptions

Acknowledgements

This research is being performed using funding received from the DOE Office of Nuclear Energy’s Nuclear Energy University Program.

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