3. DESCRIPTION: 3-D LWR Rod Ejection and Rod Withdrawal Benchmarks.
NEA-1398/02
- 3DLWRCT-1 represents the first phase of a series of LWR core transient benchmarks. It consists of 2 parts:
. PWR problem : ejection of a control assembly from an initially critical core at hot zero power or at full power, each for three different geometrical configurations. The set of problems offers a variety of reactivity excursions which efficiently test the coupled neutronic/thermal-hydraulic models of the codes.
. BWR problem: reactivity excursions caused by cold water injection and pressurization events.
2-group macroscopic cross sections and their derivatives respect to boron density, moderator temperature, moderator density and fuel temperature for BWR and PWR core materials are included.
For the PWR cases, 63 submitted solutions are analyzed in comparison with a nodal reference solution defined by using a finer spatial and temporal resolution than in standard calculations, which is included. For the BWR benchmark only the cold water injection event is discussed and evaluated in some detail. Lacking a reference solution, the evaluation of the 8 sets of BWR contributions relies on synthetic comparative discussion. - 3DLWRCT-2 represents the second phase of the PWR core transient benchmark and concerns an uncontrolled withdrawal of control rods at zero power. The same cross section data as for phase I is used.
NEA-1398/03
- 3DLWRCT-2 represents the second phase of the benchmark and concerns the results of the "rod withdrawal from zero power" benchmark, including the solutions obtained by ten participants, from ten different countries, and a reference solution, obtained by refining the spatial and time meshing.
The problem is mathematically well defined. The specification provides cross-sections for fuel, reflector and absorbers. Four cases are analysed. The submitted solutions are compared to a reference obtained with a nodal code using finer spatial and temporal resolution than in standard calculation. Besides global information such as fission power evolution, the set of results also includes local maxima, as well as hot pellet enthalpy and cladding temperature.
In general, a good agreement is obtained for most of the codes on the power evolution and its integral, in particular for the core averaged parameters. For the "hot pellet", the spread of the results is more important; however, the assumptions that define and localise this pellet are not the same for each participant: a finer local power profile reconstruction leads to more severe effects. This shows that one must take care of the consistency between the calculation methodology and the criteria applied to the safety analyses.
