US Nuclear Regulatory Commission and OECD Nuclear Energy Agency (NEA)


Boiling Water Reactor Turbine Trip (BWRTT) Benchmark

Jorge Solis, Kostadin N. Ivanov, and Baris Sarikaya
Nuclear Engineering Program
The Pennsylvania State University
University Park, PA  16802, USA

Andy M. Olson and Kenneth W. Hunt
PECO Nuclear
200 Exelon Way, KSA2-N
Kennett Square, PA 19348


Incorporation of full three-dimensional (3D) models of the reactor core into system transient codes allows for a “best-estimate” calculation of interactions between the core behavior and plant dynamics. Recent progress in the computer technology has made development of coupled system thermal-hydraulic (T-H) and neutron kinetics code systems feasible. Considerable efforts have been made in various countries and organizations in this direction. To verify the capability of the coupled codes to analyze complex transients with coupled core-plant interactions and to fully test thermal-hydraulic coupling, appropriate Light Water Reactor (LWR) transient benchmarks need to be developed on a higher “best-estimate” level. The previous sets of transient benchmark problems addressed separately system transients (designed mainly for thermal-hydraulic (T-H) system codes with point kinetics models) and core transients (designed for T-H core boundary conditions models coupled with a three-dimensional (3-D) neutron kinetics models). The Nuclear Energy Agency (NEA) of the Organization for Economic Cooperation and Development (OECD) has recently completed under the US Nuclear Regulatory Commission (NRC) sponsorship a PWR Main Steam Line (MSLB) Benchmark against coupled T-H and neuron kinetics codes. Small benchmark team from the Pennsylvania State University (PSU) has been responsible for developing the benchmark specification, assisting the participants and coordinating the benchmark activities. The benchmark was very well internationally accepted. It was felt among the participants that there should be a similar benchmark against the codes for a BWR plant transient. The Turbine Trip (TT) transients in a BWR are pressurization events in which the coupling between core phenomena and system dynamics plays an important role. In addition the available real plant experimental data makes the proposed benchmark problem very valuable. NEA, OECD and US NRC have approved it for the purpose of validating advanced system best-estimate analysis codes.

As a result a this benchmark project is established to challenge the coupled system T-H/neutron kinetics codes against a Peach-Bottom-2 (a GE-designed BWR/4) turbine trip transient with a sudden closure of the turbine stop valve. Three-turbine trip (TT) transients at different power levels were performed at the Peach Bottom (PB)-2 BWR/4 Nuclear Power Plant (NPP) prior to shutdown for refueling at the end of Cycle 2 in April 1977. The second test is selected for the benchmark problem to investigate the effect of the pressurization transient, (following the sudden closure of the turbine stop valve) on the neutron flux in the reactor core. In a best-estimate manner the test conditions approached the design basis conditions as closely as possible. The actual data were collected, including a compilation of reactor design and operating data for Cycles 1 and 2 of PB and the plant transient experimental data. The transient was selected for benchmark, because it is a dynamically complex event for which neutron kinetics in the core was coupled with thermal-hydraulics in the reactor primary system. 


The reference problem chosen for simulation in a BWR is a Turbine Trip transient, which begins with a sudden Turbine Stop Valve (TSV) closure. The pressure oscillation generated in the main steam piping propagates with relatively little attenuation into the reactor core. The induced core pressure oscillation results in dramatic changes of the core void distribution and fluid flow. The magnitude of the neutron flux transient taking place in the BWR core is strongly affected by the initial rate of pressure rise caused by pressure oscillation and has a strong spatial variation. The correct simulation of the power response to the pressure pulse and subsequent void collapse requires a 3-D core modeling supplemented by 1-D simulation of the remainder of the reactor coolant system.

The purpose of this proposal is to establish a BWR TT benchmark exercise, based on a well defined problem with complete set of input specifications and reference experimental data, for qualification of the coupled 3-D neutron kinetics/thermal-hydraulic system transient codes. Since this kind of transient is a dynamically complex event with reactor variables changing very rapidly, it constitutes a good benchmark problem to test the coupled codes on both levels: neutronics/thermal-hydraulic coupling and core/plant system coupling. Subsequently, the objectives of the proposed benchmark are: comprehensive feedback testing and examination of the capability of coupled codes to analyze complex transients with coupled core/plant interactions by comparison with actual experimental data. 

Definition of the Benchmark Exercises

The benchmark consists of three separate exercises: 

Exercise 1 - Power vs. Time Plant System Simulation with Fixed Axial Power Profile Table (Obtained from Experimental Data) 

The purpose of the first exercise is to test the thermal-hydraulic system response and to initialize the participants' system models. Core power response is fixed to reproduce the actual test results utilizing either power or reactivity vs. time data. 

Exercise 2  - Coupled 3-D Kinetics/Core Thermal-Hydraulic BC Model and/or 1-D Kinetics Plant System Simulation 

The second exercise consists of two options. Option 1 of the second exercise is to perform a coupled 3-D kinetics/thermal-hydraulic calculation for the reactor core using the PSU-provided boundary conditions at core inlet and exit. The core boundary conditions will be provided utilizing a combination of the calculated PSU results and test data. Option 2 of the second exercise is to perform coupled 1-D neutron kinetics/thermal-hydraulics core boundary condition model calculation for the core using the same boundary conditions provided for option 1. 1-D cross-sections are collapsed from the cross-section libraries generated for 3-D simulation. The participants can participate in either or both options. 

Exercise 3  - Best-Estimate Coupled 3-D Core/Thermal-Hydraulic System Modeling 

The third exercise consists also of 2 options. In Option 1 the participants perform a coupled 3-D core/thermal-hydraulic calculation for the core and 1-D thermal-hydraulics modeling for the balance of the plant. In option 2 the participants perform the calculation using a 1-D kinetics core model and 1-D thermal-hydraulics for the reactor primary system. This exercise combines elements of the first two exercises of this benchmark and is an analysis of the transient in its entirety.

The Listserver of the benchmark is now closed. However the archive of exchanged e-mails sorted according to date, subject, author and thread is accessible for those who wish to find out more about the discussion that has taken place.


For more information, please contact .

For more information on activities managed/supported by the NSC, please contact .

Last reviewed: 15 June 2011