Nuclear Energy Agency
- US/NRC PWR Main Steam Line Break Benchmark
This benchmark incorporates
full 3D modelling of the reactor core into system transient codes for "best-estimate"
simulations of the interactions between reactor core behaviour and plant dynamics
and their testing on a number of transients of importance for plant behaviour
and safety analysis. This includes verifying the capability of codes to analyse
complex transients with coupled core-plant interactions, to fully test the 3D-neutronics/thermal-hydraulics
coupling, and to evaluate discrepancies between predictions of coupled codes in
best-estimate transient simulations.
of the problem
benchmark is based on a well-defined problem concerning a PWR Main Steam Line
Break (MSLB), which may occur as a consequence of the rupture of one steam line
upstream of the main steam isolation valves. This event is characterised by
significant space-time effects in the core caused by asymmetric cooling and
an assumed stuck-out control rod during reactor trip. It is based on reference
design and data from the Three Mile Island Unit 1 Nuclear Power Plant (TMI-1).
It includes a description of the event sequence with set points of all activated
system functions and typical plant conditions during the transient.
benchmark consists of three exercises:
- point kinetics
simulation to test the primary and secondary system model responses;
- coupled 3D
neutronics/core thermal-hydraulics response evaluation using inlet and outlet
core transient boundary conditions;
coupled core-plant transient modelling.
first two exercises have helped to fine-tune the models used in the different
codes in order to ensure they all solve the same problem. Parametric studies
and scenarios were developed to help understand the source of uncertainties.
A series of statistical methods has been applied to analyse code-to-code comparisons
involving different types of data single-values, 1-D and 2-D distributions,
and time histories. The statistical methods have been modified to analyse correctly
relative normalised parameters.
exercise was co-organised by the OECD/NEA Nuclear Science Committee, the OECD/NEA
Committee on the Safety of Nuclear Installations, and the US Nuclear Regulatory
Commission. It involved about 70 experts from 15 countries representing 30 organisations.
It brought together specialists in neutronics and thermal hydraulics from universities,
research centres, utilities, engineering companies and vendors. It was co-ordinated
by the Penn State University Nuclear Engineering Program Team.
exercise has shown:
- a proof of
principle that coupling 3D neutronics with thermal hydraulics is feasible
- 3D coupling
provides more detailed insight into phenomena occurring in the core during
transients, required for engineering simulations: power plant operators seek
to know what happens in detail during transients;
methods provide margins for safety limits, allowing more flexibility in plant
methods will be used both for reactor operation and safety analysis and that
tools common to both will emerge;
- its timely
organisation has not only achieved a comparison of the performance of different
codes but has driven the development of coupled 3D neutronics/thermal hydraulics
codes, in particular optimal coupling schemes through parametric studies;
- the exercise
has also provided a template for multi-level benchmark methodology to be used
for complex problems;
is a need to
develop a common approach for sensitivity/uncertainty analysis in neutronics
and thermal hydraulics.