Spatial instability analysis in pressurized water reactors

A simple methodology has been developed to assess the spatial dynamic behavior of large PWRs against xenon spatial instability in different modes. Method of analysis aims to analyze xenon dynamic behavior against anticipated reactivity perturbations. Reactivity perturbations in different modes have been evaluated based on reactivity device movements as well as localized thermal variations in the core. Effect of individual core design and operating parameters on xenon spatial instability has been studied. Behavior of spatial stability index (SI) with core size is investigated. Based on SI-core size curve, a threshold core size has been determined beyond which a PWR core tends to become spatially unstable. Methodology has been used to assess the spatial xenon dynamic behavior of different modes of oscillations in VVER1000 and AP1000 reactor cores. (C) 2010 Elsevier Ltd. All rights reserved

Towards an efficient multiphysics model for nuclear reactor dynamics

Availability of fast computer resources nowadays has facilitated more in-depth modeling of complex engineering systems which involve strong multiphysics interactions. This multiphysics modeling is an important necessity in nuclear reactor safety studies where efforts are being made worldwide to combine the knowledge from all associated disciplines at one place to accomplish the most realistic simulation of involved phenomenon. On these lines coupled modeling of nuclear reactor neutron kinetics, fuel heat transfer and coolant transport is a regular practice nowadays for transient analysis of reactor core. However optimization between modeling accuracy and computational economy has always been a challenging task to ensure the adequate degree of reliability in such extensive numerical exercises. Complex reactor core modeling involves estimation of evolving 3-D core thermal state, which in turn demands an expensive multichannel based detailed core thermal hydraulics model. A novel approach of power weighted coupling between core neutronics and thermal hydraulics presented in this work aims to reduce the bulk of core thermal calculations in core dynamics modeling to a significant extent without compromising accuracy of computation. Coupled core model has been validated against a series of international benchmarks. Accuracy and computational efficiency of the proposed multiphysics model has been demonstrated by analyzing a reactivity initiated transient

Investigations on Neutronic Decoupling Phenomenon in Large Nuclear Reactors

AbstractCharacteristic size (size of core expressed in terms of neutron migration length) of a nuclear reactor has been used as a simple thumb rule to assess the degree of neutronic coupling the reactor core. In the present paper, neutronic decoupling phenomena of reactor cores is studied using a sophisticated technique, called eigenvalue separation (EVS). A large eigenvalue separation reflects a distant higher harmonic and ensures neutronic coupling in respective neutron flux mode of the core. The EVS analysis technique has been used to optimize core geometry of pressurized water reactors, i.e., Height to Diameter (H/D) ratio of the reactor has been optimized to minimize the deep influence of higher harmonics on the stability. The study brings out that a typical H/D rat io beyond 1.3 should be avoided in large sized cores to maximize core neutronic coupling and hence the stability of the reactor

DEVELOPMENT AND VALIDATION OF COUPLED DYNAMICS CODE 'TRIKIN' FOR VVER REACTORS

New generation nuclear reactors are designed using advanced safety analysis methods. A thorough understanding of different interacting physical phenomena is necessary to avoid underestimation and overestimation of consequences of off-normal transients in the reactor safety analysis results. This feature requires a multiphysics reactor simulation model. In this context, a coupled dynamics model based on a multiphysics formulation is developed indigenously for the transient analysis of large pressurized VVER reactors. Major simplifications are employed in the model by making several assumptions based on the physics of individual phenomenon. Space and time grids are optimized to minimize the computational bulk. The capability of the model is demonstrated by solving a series of international (AER) benchmark problems for VVER reactors. The developed model was used to analyze a number of reactivity transients that are likely to occur in VVER reactors

Development, validation and application of multi-point kinetics model in RELAP5 for analysis of asymmetric nuclear transients

Point kinetics approach in system code RELAP5 limits its use for many of the reactivity induced transients, which involve asymmetric core behaviour. Development of fully coupled 3D core kinetics code with system thermal-hydraulics is the ultimate requirement in this regard; however coupling and validation of 3D kinetics module with system code is cumbersome and it also requires access to source code. An intermediate approach with multi-point kinetics is appropriate and relatively easy to implement for analysis of several asymmetric transients for large cores. Multi-point kinetics formulation is based on dividing the entire core into several regions and solving ODEs describing kinetics in each region. These regions are interconnected by spatial coupling coefficients which are estimated from diffusion theory approximation. This model offers an advantage that associated ordinary differential equations (ODEs) governing multi-point kinetics formulation can be solved using numerical methods to the desired level of accuracy and thus allows formulation based on user defined control variables, i.e., without disturbing the source code and hence also avoiding associated coupling issues. Euler's method has been used in the present formulation to solve several coupled ODEs internally at each time step. The results have been verified against inbuilt point-kinetics models of RELAP5 and validated against 3D kinetics code TRIKIN. The model was used to identify the critical break in RIH of a typical large PHWR core. The neutronic asymmetry produced in the core due to the system induced transient was effectively handled by the multi-point kinetics model overcoming the limitation of in-built point kinetics model of RELAP5 and standalone 3D core kinetics codes. (C) 2016 Elsevier B.V. All rights reserved

Linear stability analysis of a nuclear reactor using the lumped model

The stability analysis of a nuclear reactor is an important aspect in the design and operation of the reactor. A stable neutronic response to perturbations is essential from the safety point of view. In this paper, a general methodology has been developed for the linear stability analysis of nuclear reactors using the lumped reactor model. The reactor kinetics has been modelled using the point kinetics equations and the reactivity feedbacks from fuel, coolant and xenon have been modelled through the appropriate time dependent equations. These governing equations are linearized considering small perturbations in the reactor state around a steady operating point. The characteristic equation of the system is used to establish the stability zone of the reactor considering the reactivity coefficients as parameters. This methodology has been used to identify the stability region of a typical pressurized heavy water reactor. It is shown that the positive reactivity feedback from xenon narrows down the stability region. Further, it is observed that the neutron kinetics parameters (such as the number of delayed neutron precursor groups considered, the neutron generation time, the delayed neutron fractions, etc.) do not have a significant influence on the location of the stability boundary. The stability boundary is largely influenced by the parameters governing the evolution of the fuel and coolant temperature and xenon concentration