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The OECD/NEA - US/NRC
PWR Main Steam-Line Break Benchmark Summar
Scope / Objective
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.
Description of the problem
The 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.
It consists of three exercises:
1. Point kinetics simulation to test the primary and secondary system model responses
2. Coupled 3D neutronics / core thermal-hydraulics response evaluation using inlet and outlet core transient boundary conditions
3. Best-estimate coupled core – plant transient modelling
The first two exercises have helped tuning the models in the different codes in order to ensure they all solve the same problem and parametric studies and scenarios were developed to help understanding the source of uncertainties. Suite of statistical methods has been applied to analyse code-to-code comparisons involving different type of data – single values, 1-D and 2-D distributions, and time histories. The statistical methods have been modified to analyse correctly relative normalized parameters.
a proof of principle that coupling 3D neutronics with thermal-hydraulics is feasible and working;
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 details during transients;
“best-estimate” methods provide margins to safety limits, allowing more flexibility in plant operation;
“best-estimate” 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;
there is a need to develop a common approach for sensitivity/uncertainty analysis in neutronics and thermal-hydraulics.
K.N. Ivanov, T.M. Beam, A.J. Baratta, A. Irani, N. Trikouros: Pressurised Water Reactor Main Steam Line Break (MSLB) Benchmark, Volume I: Final Specifications: OECD NEA/NSC/DOC(99)8, April 1999.
T.M. Beam, K.N. Ivanov, B. Taylor, A.J. Baratta: Pressurised Water Reactor Main Steam Line Break (MSLB) Benchmark, Volume II: Summary results of Phase I (Point Kinetics): OECD NEA/NSC/DOC(2000)21, December 2000.
B. Taylor, N. K. Todorova, K.N. Ivanov: Pressurised Water Reactor Main Steam Line Break (MSLB) Benchmark, Volume III: Results of Phase 2 on 3-D Core Boundary Conditions Model: OECD NEA/NSC/DOC(2002)12, October 2002.
N.K. Todorova, B. Taylor, K.N. Ivanov: Pressurised Water Reactor Main Steam Line Break (MSLB) Benchmark, Volume IV: Summary results of Phase III (Best Estimate Coupled Simulation), OECD NEA/NSC/DOC(2003)21, 2003.
K.N. Ivanov, A.J. Baratta, E. Sartori: OECD/NRC MSLB Benchmark - A Systematic Approach to Validate Best-Estimate Coupled Codes Using a Multi-Level Methodology, OECD/CSNI Workshop on "Advanced Thermal-Hydraulic and Neutronic Codes: Current and Future Applications", April 10-13, 2000, Barcelona, Spain.
B. Taylor, K. Ivanov: Statistical Methods used for Code-to-Code Comparisons in the OECD/NRC PWR MSLB Benchmark, Annals of Nuclear Energy 27 (2000), 1589-1605
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.