A Study of Superhomogenization Applied to PHWR Lattices

2019 ◽  
Vol 6 (1) ◽  
Author(s):  
Thomas A. Ferguson ◽  
Eleodor M. Nichita
Keyword(s):  
Cross Sections ◽  
Homogenization Method ◽  
Single Node ◽  
Rectangular Array ◽  
Water Reactor ◽  
Production Type ◽  
Fuel Pin ◽  
Heavy Water Reactor ◽  
Fuel Elements ◽  
Full Core

Abstract To reduce computational expenses, full-core production-type neutronics calculations are customarily performed using a simplified core-model whereby large regions of the core, called nodes, are assumed to be homogeneous. The process of generating the few-group homogenized-node macroscopic cross sections is called lattice homogenization. The simplest homogenization method is standard homogenization (SH) and full-core models based on it do not usually reproduce heterogeneous-core calculations too closely. To improve agreement between node-homogenized core results and heterogeneous-core results, advanced homogenization techniques are used. Such techniques tend to use additional parameters besides homogenized macroscopic cross sections. Superhomogenization (SPH) is an advanced lattice homogenization method, which has been developed initially for light-water-reactor (LWR) lattices whereby fuel elements are arranged in a rectangular array. It has the advantage of not requiring any modification to the full-core diffusion code for its implementation. For LWRs, SPH establishes neutronic equivalence between detailed-geometry heterogeneous fuel-pin cells and homogenized fuel-pin cells by adjusting homogenized multigroup macroscopic cross sections and diffusion coefficients. This work investigates the possible use of the SPH methodology for pressurized heavy-water reactor (PHWR) lattices whose fuel pins are arranged in concentric rings rather than in a rectangular array. Results for single-node (SN) as well as multinode (MN) lattice-calculation models are presented. Results show that, with proper region definition, the SPH methodology can be used for PHWR lattices but that improvement in homogenization accuracy is only marginal compared with SH when comparing results for the same type of lattice model (SN or MN).

Critical experiments on the stand of a heavy-water reactor with rod fuel elements

Atomic Energy ◽  
1962 ◽  
Vol 12 (2) ◽  
pp. 171-174
Author(s):  
Yu. G. Abov ◽  
V. F. Belkin ◽  
P. A. Krupchitskii
Keyword(s):  
Heavy Water ◽  
Water Reactor ◽  
Heavy Water Reactor ◽  
Fuel Elements

scholarly journals Evaluation of PBMR-400 Core Design Steady State Condition with Serpent and AGREE

2021 ◽  
Vol 2048 (1) ◽  
pp. 012024
Author(s):  
H Ardiansyah ◽  
V Seker ◽  
T Downar ◽  
S Skutnik ◽  
W Wieselquist
Keyword(s):  
Monte Carlo ◽  
Cross Sections ◽  
Reactor Core ◽  
Light Water Reactor ◽  
Steady State Condition ◽  
Light Water ◽  
Water Reactor ◽  
The Core ◽  
Core Simulation ◽  
Full Core

Abstract The significant recent advances in computer speed and memory have made possible an increasing fidelity and accuracy in reactor core simulation with minimal increase in the computational burden. This has been important for modeling some of the smaller advanced reactor designs for which simplified approximations such as few groups homogenized diffusion theory are not as accurate as they were for large light water reactor cores. For narrow cylindrical cores with large surface to volume ratios such the Ped Bed Modular Reactor (PBMR), neutron leakage from the core can be significant, particularly with the harder neutron spectrum and longer mean free path than a light water reactor. In this paper the core from the OECD PBMR-400 benchmark was analyzed using multigroup Monte Carlo cross sections in the HTR reactor core simulation code AGREE. Homogenized cross sections were generated for each of the discrete regions of the AGREE model using a full core SERPENT Monte Carlo model. The cross sections were generated for a variety of group structures in AGREE to assess the importance of finer group discretization on the accuracy of the core eigenvalue and flux predictions compared to the SERPENT full core Monte Carlo solution. A significant increase in the accuracy was observed by increasing the number of energy groups, with as much as a 530 pcm improvement in the eigenvalue calculation when increasing the number of energy groups from 2 to 14. Significant improvements were also observed in the AGREE neutron flux distributions compared to the SERPENT full core calculation.

Further Evidence of Zircaloy Corrosion in Fuel Elements Irradiated in a Steam Generating Heavy Water Reactor

2009 ◽  
pp. 107-107-15 ◽  
Author(s):  
R Sumerling ◽  
A Garlick ◽  
A Stuttard ◽  
JM Hartog ◽  
FW Trowse ◽  
...  
Keyword(s):  
Heavy Water ◽  
Water Reactor ◽  
Heavy Water Reactor ◽  
Fuel Elements

Application of Cross Sections Uncertainty Propagation Framework to Light and Heavy Water Reactor Systems

2019 ◽  
Vol 6 (1) ◽  
Author(s):  
Dongli Huang ◽  
Hany S. Abdel-Khalik
Keyword(s):  
Cross Sections ◽  
Power Distribution ◽  
Degrees Of Freedom ◽  
Uncertainty Propagation ◽  
Nuclear Data ◽  
Companion Paper ◽  
Water Reactor ◽  
Reduction Techniques ◽  
Uncertainty Space ◽  
Heavy Water Reactor

Abstract Uncertainty quantification has been recognized by the community as an essential component of best-estimate reactor analysis simulation because it provides a measure by which the credibility of the simulation can be assessed. In a companion paper, a framework for the propagation of nuclear data uncertainties from the multigroup level through lattice physics and core calculations and ultimately to core responses of interest has been developed. The overarching goal of this framework is to automate the propagation, prioritization, mapping, and reduction of uncertainties for reactor analysis core simulation. This paper employs both heavy and light water reactor systems to exemplify the application of this framework. Specifically, the paper is limited to the propagation of the nuclear data starting with the multigroup cross section covariance matrix and down to core responses, e.g., eigenvalue and power distribution, in steady-state core wide calculations. The goal is to demonstrate how the framework employs reduction techniques to compress the uncertainty space into a very small number of active degrees-of-freedom (DOFs), which renders the overall process computationally feasible for day-to-day engineering evaluations.

scholarly journals Nondestructive verification of the exposure of heavy-water reactor fuel elements

1982 ◽  
Author(s):  
J.R. Phillips ◽  
T.R. Bement ◽  
C.R. Hatcher ◽  
S.T. Hsue ◽  
D.M. Lee
Keyword(s):  
Heavy Water ◽  
Reactor Fuel ◽  
Water Reactor ◽  
Heavy Water Reactor ◽  
Fuel Elements

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

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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