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NEA-1640 ZZ-BWRTT.

ZZ BWRTT, BWR Turbine Trip Transient Benchmark Based on Peach-Bottom 2

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1. NAME OR DESIGNATION OF PROGRAM:  ZZ-BWRTT.
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2. COMPUTERS
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Program name Package id Status Status date
ZZ-BWRTT NEA-1640/08 Tested 12-AUG-2005

Machines used:

Package ID Orig. computer Test computer
NEA-1640/08 Many Computers PC Windows
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3. DESCRIPTION OF PROGRAM OR FUNCTION

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 TTbenchmark 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.
  
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.
NEA-1640/08
The present edition (Revision 5 - 12 August 2005) differs from the previous one (4 December 2003)  in the following:
  
1. The PDF version of the "Boiling Water Reactor Turbine Trip (TT) Benchmark - Volume II : Summary Results of Exercise 1" authored by B. Akdeniz, K. Ivanov, and A. Olson. OECD 2005, NEA/NSC/DOC(2004)21 was added.
  
Note: The present edition contains the cumulative information of all previous meetings and replaces the previous ones.
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9. STATUS
Package ID Status date Status
NEA-1640/08 12-AUG-2005 Tested restricted
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10. REFERENCES
NEA-1640/08, included references:
- Boiling Water Reactor Turbine Trip (TT) Benchmark - Volume I,  Benchmark
Specification (25 June 2001) authored by J. Solis, K. Ivanov, B. Sarikaya, A.
Olson and K.W. Hunt (final printed version), NEA/NSC/DOC(2001)1, ISBN
92-64-18470-8
- Boiling Water Reactor Turbine Trip (TT) Benchmark - Volume II : Summary
Results of Exercise 1,  authored by B. Akdeniz, K. Ivanov, and A. Olson. OECD
2005, NEA/NSC/DOC(2004)21, ISBN 92-64-01064-05
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12. PROGRAMMING LANGUAGE(S) USED

No item found

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15. NAME AND ESTABLISHMENT OF AUTHORS

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, USA
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16. MATERIAL AVAILABLE
NEA-1640/08
Directory 3d-bc - the boundary conditions
Directory Decay-Heat - decay-heat data to be used
Directory EPRIReports - the NP-563 and NP-564 reports
Directory Philadelphia - the papers distributed at the first BWRTT Workshop
Directory PSI - papers distributed at the second BWRTT Workshop
Directory FZR - papers distributed at the third BWRTT Workshop
Directory PHY - papers distributed at the fourth BWRTT Workshop
Directory BARCA - papers distributed at the fifth BWRTT Workshop
Directory Results - Results of the exercises
Directory Specifications - files for problem specifications
- subdirectory jetpump contains a description of the jetpump model
- subdirectory templates contains templates for submitting results
- subdirectory additional contains information / clarifications
- subdirectory volume one contains the Volume 1 = Specification (update)
- subdirectories / unix/ & / nonunix/ contain the RETRAN input decks
directory XS-Lib - the cross section sets
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17. CATEGORIES
  • Y. Integral Experiments Data, Databases, Benchmarks

Keywords: boiling water reactor, neutronics, thermal hydraulics, transients, turbines.