Power series solution method for solving point kinetics equations with lumped model temperature and feedback
Creators
- 1. Reactor Physics Division, Indira Gandhi Centre for Atomic Research, Kalpakkam, Tamilnadu 603 102 (India)
Description
Point kinetics equations are stiff differential equations, and their solution by the conventional explicit methods will give a stable consistent result only for very small time steps. Since the neutron lifetime in a LMFBR is very short, the point kinetics equations for LMFBRs become even stiffer. In this study the power series solution (PWS) method is applied for solving the point kinetics equations for a typical LMFBR. A Fortran program is developed for accident analysis of LMFBRs with the PWS method for solving the point kinetics and a lumped model for solving the heat transfer equations. A new technique is developed with fixing factor to find out the average temperature at the peak power node (PPN) without performing temperature calculations at all axial nodes in a reactor fuel pin. The temperature at PPN also decides whether the reactor is within the design safety limit (DSL) or it has entered a serious transient that may lead to an accident. The coupled heat transfer and point kinetics models for a peak power node give the average fuel, clad and coolant temperatures. For the transient over power accidents (TOPA), this is the best way for calculating the temperature, with minimum amount of computations. TOPA analyses are carried out with PWS method. It is found that the PWS methodology uses a small number of numerical operations, while the computational time and the accuracy are comparable with the available fast computational tools. This methodology can be used in nuclear reactor simulation studies and accident analysis
Availability note (English)
Available from http://dx.doi.org/10.1016/j.anucene.2008.11.005Additional details
Identifiers
- DOI
- 10.1016/j.anucene.2008.11.005;
- PII
- S0306-4549(08)00288-0;
Publishing Information
- Journal Title
- Annals of Nuclear Energy (Oxford)
- Journal Volume
- 36
- Journal Issue
- 2
- Journal Page Range
- p. 246-250
- ISSN
- 0306-4549
- CODEN
- ANENDJ
INIS
- Country of Publication
- United Kingdom
- Country of Input or Organization
- International Atomic Energy Agency (IAEA)
- INIS RN
- 40045568
- Subject category
- S11: NUCLEAR FUEL CYCLE AND FUEL MATERIALS; S22: GENERAL STUDIES OF NUCLEAR REACTORS;
- Descriptors DEI
- ACCURACY; CALCULATION METHODS; COOLANTS; DESIGN; DIFFERENTIAL EQUATIONS; FORTRAN; HEAT TRANSFER; LMFBR TYPE REACTORS; NUCLEAR FUELS; PEAK LOAD; POWER SERIES; RADIATION ACCIDENTS; REACTOR FUELING; SIMULATION
- Descriptors DEC
- ACCIDENTS; BREEDER REACTORS; ENERGY SOURCES; ENERGY TRANSFER; EPITHERMAL REACTORS; EQUATIONS; FAST REACTORS; FBR TYPE REACTORS; FUELS; LIQUID METAL COOLED REACTORS; MATERIALS; PROGRAMMING LANGUAGES; REACTOR MATERIALS; REACTORS; SERIES EXPANSION
Optional Information
- Copyright
- Copyright (c) 2008 Elsevier Science B.V., Amsterdam, The Netherlands, All rights reserved.