Development and verification of pin-by-pin homogenized simplified transport solver Tortin for PWR core analysis
Creators
- 1. Ecole Polytechnique Federale de Lausanne, Lausanne (Switzerland)
Description
Currently, the pin-by-pin homogenized solvers are a very active research field as they can, unlike the nodal codes, directly predict the local power, while requiring significantly less computational resources than the heterogeneous transport codes. This paper presents a recently developed pin-by-pin diffusion/SP3 solver Tortin, its spatial discretization method and the reflector treatment. Regarding the spatial discretization, it was observed that the finite difference method applied on pin-cell size mesh does not properly capture the big flux change between MOX and uranium fuel, while the nodal expansion method is more accurate but too slow. If the finite difference method is used with a finer mesh in the outer two pin rows of the fuel assembly, it increases the required computation time by only 50%, but decreases the pin power errors below 1% with respect to lattice code reference solutions. The paper further describes the coupling of Tortin with a microscopic depletion solver. Several verification tests show that the SP3 pin-by-pin solver can reproduce the heterogeneous transport solvers results with very good accuracy, even for fuel cycle depletion of very heterogeneous core employing MOX fuel or inserted control rods, while being two orders of magnitude faster
Additional details
Publishing Information
- Journal Title
- Nuclear Engineering and Technology
- Journal Volume
- 52
- Journal Issue
- 11
- Series
- 27 refs, 15 figs, 2 tabs
- Journal Page Range
- p. 2431-2441
- ISSN
- 1738-5733
INIS
- Country of Publication
- Korea, Republic of
- Country of Input or Organization
- Korea, Republic of
- INIS RN
- 52113468
- Subject category
- S21: SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS; S97: MATHEMATICAL METHODS AND COMPUTING;
- Descriptors DEI
- ACCURACY; CONTROL ELEMENTS; DIFFUSION; FINITE DIFFERENCE METHOD; FUEL ASSEMBLIES; FUEL CYCLE; MIXED OXIDE FUELS; NODAL EXPANSION METHOD; PWR TYPE REACTORS; TRANSPORT; URANIUM; VERIFICATION
- Descriptors DEC
- ACTINIDES; CALCULATION METHODS; ELEMENTS; ENERGY SOURCES; ENRICHED URANIUM REACTORS; FUELS; ITERATIVE METHODS; MATERIALS; MATHEMATICAL SOLUTIONS; METALS; NUCLEAR FUELS; NUMERICAL SOLUTION; POWER REACTORS; REACTOR COMPONENTS; REACTOR MATERIALS; REACTORS; SOLID FUELS; THERMAL REACTORS; WATER COOLED REACTORS; WATER MODERATED REACTORS