This benchmark incorporated full 3-D 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 included 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.
The benchmark was based on a well-defined problem concerning a pressurised water reactor 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 included a description of the event sequence with set points of all activated system functions and typical plant conditions during the transient.
It consisted of three exercises:
The first two exercises helped to tune the models 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 suite of statistical methods had 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 had been modified to correctly analyse relative normalized parameters.
The exercise was co-organised by the OECD/NEA Nuclear Science Committee and the Committee on the Safety of Nuclear Installations and the US Nuclear Regulatory Commission. It involved around 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. They worked together on this project. It was co-ordinated by the United States Pennsylvania State University Nuclear Engineering Programme Team.
The exercise had shown that:
Data related to the MSLB Benchmark can be requested from the NEA Databank.
J.Bryce Taylor, Kostadin N Ivanov, "Statistical Methods used for Code-to-code Comparisons in the OECD/NRC PWR MSLB Benchmark", Annals of Nuclear Energy, Volume 27, Issue 17, 2000, pages 1589-1605, doi: doi.org/10.1016/S0306-4549(00)00014-1.
B. Taylor, K. Ivanov, D. Ebert: Comparative Analysis of Best Estimate Solutions of the OECD/NRC MSLB Benchmark, Proceedings of the ANS/ENS Embedded Topical Meeting BE2000 – “Best Estimate Methods in Nuclear Installation Safety Analyses”, November 12-16, 2000, Washington D.C., USA, www.ans.org/store/item-700282.
The goal of the Benchmark for Uncertainty Analysis in Best-Estimate Modelling for Design, Operation and Safety Analysis of Light Water Reactors (LWR-UAM) was to determine the uncertainty in light water reactor (LWR) systems and processes in all stages of calculations. It was estimated through a simulation process of nine exercises in three phases provided by the benchmarking framework.
This benchmark was a continuation of the V1000CT activities and defined a coupled code problem for further validation of thermal-hydraulics system codes for application with Russian-designed VVER-1000 reactors based on actual plant data from the Russian nuclear power plant Kalinin Unit 3 (Kalinin-3)
The Working Party on Scientific Issues and Uncertainty Analysis of Reactor Systems (WPRS) studies the reactor physics, fuel performance, and radiation transport and shielding in present and future nuclear power systems.