POWER LEVEL EFFECTS ON THORIUM-BASED FUELS IN PRESSURE-TUBE HEAVY WATER REACTORS

Lattice and core physics modeling and calculations have been performed to quantify the impact of power/flux levels on the reactivity and achievable burnup for 35-element fuel bundles made with Pu/Th or U-233/Th. The fissile content in these bundles has been adjusted to produce on the order of 20 MWd/kg burnup in homogeneous cores in a 700 MWe-class pressure-tube heavy water reactor, operating on a once-through thorium cycle. Results demonstrate that the impact of the power/flux level is modest for Pu/Th fuels but significant for U-233/Th fuels. In particular, high power/flux reduces the breeding and burnup potential of U-233/Th fuels. Thus, there may be an incentive to operate reactors with U-233/Th fuels at a lower power density or to develop alternative refueling schemes that will lower the time-average specific power, thereby increasing burnup.

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Sensitivity Studies to Assess the Impact of Geometry and Operating/Boundary Condition Perturbations on Thermal-Hydraulic Behavior of Advanced Fuel Channels in Pressure Tube Heavy Water Reactors with Uranium and Thorium-Based Fuels

Nuclear Technology ◽

10.1080/00295450.2020.1827658 ◽

2021 ◽

pp. 1-27

Author(s):

B. P. Bromley ◽

Z. Cheng ◽

A. Nava Dominguez ◽

A. V. Colton

Keyword(s):

Boundary Condition ◽

Heavy Water ◽

Pressure Tube ◽

Hydraulic Behavior ◽

Sensitivity Studies ◽

Advanced Fuel ◽

Water Reactors ◽

The Impact

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Simulations of a Pressure-Tube Heavy Water Reactor Operating on Near-Breeding Thorium Cycles

Nuclear Technology ◽

10.13182/nt15-30 ◽

2016 ◽

Vol 194 (2) ◽

pp. 178-191 ◽

Cited By ~ 3

Author(s):

Sourena Golesorkhi ◽

Blair P. Bromley ◽

Matthew H. Kaye

Keyword(s):

Heavy Water ◽

Pressure Tube ◽

Water Reactor ◽

Heavy Water Reactor ◽

Reactor Operating

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MODELING STUDIES AND CODE-TO-CODE COMPARISONS FOR PRESSURE TUBE HEAVY WATER REACTOR CORES

CNL Nuclear Review ◽

10.12943/cnr.2018.00003 ◽

2018 ◽

Vol 7 (2) ◽

pp. 177-200

Author(s):

Huiping Yan ◽

Blair Patrick Bromley ◽

Cliff Dugal ◽

Ashlea V. Colton

Keyword(s):

Heavy Water ◽

Neutron Transport ◽

Software Tool ◽

Natural Uranium ◽

Pressure Tube ◽

Core Model ◽

Test Case ◽

Reactor Physics ◽

Water Reactors ◽

The Impact

Preliminary, conceptual studies have been performed previously using deterministic lattice physics (WIMS-AECL) and core physics codes (RFSP) to estimate performance and safety characteristics of various thorium-based fuels and uranium-based fuels augmented by small amounts of thorium for use in pressure tube heavy-water reactors (PT-HWRs). To confirm the validity of the results, the WIMS-AECL/RFSP results are compared against predictions made with the stochastic neutron transport code MCNP. This paper describes the development of a method for setting up an MCNP core model of at PT-HWR for comparison with WIMS-AECL/RFSP results, using a core with 37-element natural uranium fuel bundles as a test case for sensitivity studies. These studies included evaluating the sensitivity of the bias of the effective neutron multiplication factor (keff), a source convergence study, uncertainties correction with multiple independent simulations, the impact of irradiation map binning methods, and the impact of reflector models. A Python-based software scripting tool was developed to automate the creation, execution, and post-processing of reactor physics data from the MCNP models. The software tool and algorithm for creating an MCNP core model using data from the WIMS-AECL and RFSP models are described in this paper. Based on the preliminary evaluations of the simulation parameters with the base model, reactor physics analyses were performed for PT-HWR cores with thorium-based fuels in a 35-element bundle type. Code-to-code results demonstrate good agreement between MCNP and RFSP, giving confidence in the method developed and its applicability to other fuels and core types.

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Effect of Inlet Configuration on Moderator Velocity and Temperature Distribution Inside the Calandria of a Heavy Water Reactor

Volume 2: Fuel Cycle and High Level Waste Management; Computational Fluid Dynamics, Neutronics Methods and Coupled Codes; Student Paper Competition ◽

10.1115/icone16-48347 ◽

2008 ◽

Author(s):

Abhijeet Mohan Vaidya ◽

Naresh Kumar Maheshwari ◽

Pallippattu Krishnan Vijayan ◽

Dilip Saha ◽

Ratan Kumar Sinha

Keyword(s):

Temperature Distribution ◽

Impact Velocity ◽

Heavy Water ◽

Computational Study ◽

Water Reactor ◽

Heavy Water Reactor ◽

The Impact

Computational study of the moderator flow in calandria vessel of a heavy water reactor is carried out for three different inlet nozzle configurations. For the computations, PHOENICS CFD code is used. The flow and temperature distribution for all the configurations are determined. The impact of moderator inlet jets on adjacent calandria tubes is studied. Based on these studies, it is found that the inlet nozzles can be designed in such a way that it can keep the impact velocity on calandria tubes within limit while keeping maximum moderator temperature well below its boiling limit.

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