last modified: 07-DEC-2011 | catalog | categories | new | search |

NEA-1693 ZZ-V1000CT1&2.

ZZ V1000CT-1&2, VVER-1000 Main Coolant Pump Switching-on, Coolant Mixing Tests, Main Steam-Line Break Benchmarks

top ]
1. NAME OR DESIGNATION:  ZZ-V1000CT1&2.
top ]
2. COMPUTERS
To submit a request, click below on the link of the version you wish to order. Only liaison officers are authorised to submit online requests. Rules for requesters are available here.
Program name Package id Status Status date
ZZ-V1000CT1&2 NEA-1693/07 Arrived 07-DEC-2011

Machines used:

Package ID Orig. computer Test computer
NEA-1693/07 Many Computers
top ]
3. DESCRIPTION

This VVER-1000 coolant transient (V1000CT) benchmark set, defines coupled code standard problems based on actual plant data. The overall objective is to assess computer codes used in the safety analysis of VVER power plants, specifically for their use in reactivity transients in VVER-1000. In performing this work the Pennsylvania State University (PSU), USA and CEA/DEN, France have collaborated with Bulgarian organisations, in particular with the Institute of Nuclear Research and Nuclear Energy (INRNE) and the Kozloduy nuclear power plant (KNPP).
  
The V1000CT benchmark set consists of two parts: V1000CT-1 is a simulation of the switching on of one main coolant pump (MCP) when the other three MCPs are in operation, and V1000CT-2 is a calculation of coolant mixing experiments and a main steam line break (MSLB) transient. Each of the two parts contains three exercises.
  
V1000CT-1: Main Coolant Pump (MCP) Switching On
  
V1000CT-1 Definition and Objectives
  
The reference problem chosen for simulation in a VVER-1000 is a MCP switching on when the other three main coolant pumps are in operation. It is an experiment that was conducted by Bulgarian and Russian engineers during the plant-commissioning phase at the Kozloduy NPP Unit 6 as a part of the start-up tests. The test was done because it is important for the safety of the NPP with VVER-1000, model 320. The reactor is at the beginning of cycle (BOC) with average core exposure of 30.7 EFPD and boron concentration 5.95 g/kg H2O. At the beginning of the experiment there are three pumps in operation - 1st, 2nd and 4th main coolant pumps and the reactor power is at 29.45% of nominal power level according to the equipment that measures the neutron flux. MCPs 1, 2 and 4 are operating under stable conditions and MCP 3 is out of operation. The initial conditions are as follows. The average inlet temperature in the reactor core is about 555.00  deg.K. The temperature differences between the hot and cold legs for the loop with working MCPs vary between 8.3-11.5 deg.K while the same temperature difference for the loop 3 with the MCP out of operation is -3.6 deg.K. The total mass flow through the core is about 13611 kg/s with an average flow of 5000 kg/s through each of the working loops and negative (reverse) flow of -1544 kg/s in loop 3. There is a core axial non-symmetry as can be seen from the value of the axial offset of the core power distribution at the initial state - 28.5 %. The control rod group 10 is inserted into the core at about 36% of the reactor core height. Analysis of the initial 3-D relative power distribution showed that this insertion introduced axial power asymmetry in the core. At the beginning of the transient there is also a radial thermal-hydraulic asymmetry coming from the colder water introduced in 1/4 of the core when MCP 3 is switched on. This causes a spatial asymmetry in the reactivity feedback, which is propagated through the transient and combined with insertion of positive reactivity.
  
In summary, this event is characterized by rapid increase in the flow through the core resulting in a coolant temperature decrease, which is spatially dependent. This leads to insertion of spatially distributed positive reactivity due to the modeled feedback mechanisms and non-symmetric power distribution. Simulation of the transient requires evaluation of core response from a multi-dimensional perspective (coupled three-dimensional neutronics/core thermal-hydraulics) supplemented by a one-dimensional simulation of the remainder of the reactor coolant system.
  
Definition of the V1000CT-1 Benchmark Exercises
  
The benchmark consists of three exercises.
Exercise 1 - Point Kinetics Plant Simulation
Exercise 2  - Coupled 3-D Kinetics/Core Thermal-Hydraulic Response Evaluation,
Exercise 3  - Best-Estimate Coupled 3-D Core/Thermal-Hydraulic Plant Transient Modeling. In addition, an extreme version of Exercise 3 is defined as follows: the control rod of group 10 located in the sector of the core cooled by MCP 3 is ejected after switching on of the MCP 3. The ejection of the rod begins at the 13th second of the transient and the velocity of ejection is 17.75 m/sec. The scram is activated upon reaching the high neutron flux set point, which is 107% of the initial level. This extreme scenario will develop very peaked spatial power distribution and nonlinear asymmetric feedback effects. It is designed to test and compare better the predictions of coupled 3-D kinetics / thermal-hydraulic codes.
  
The benchmark specifications for Phase 1 include a complete set of the initial and boundary conditions that are needed for the participants to perform the aforementioned three exercises.
==============================
  
V1000CT-2: Coolant Mixing Tests and Main Steam-Line Break (MSLB)
  
V1000CT-2 Definition and Objectives
  
This benchmark aims at improving the flow mixing computation tools in integrated codes y. For this purpose the coolant mixing & main steam line break (MSLB) benchmark for VVER-1000 (V1000CT-2) is defined. Reference plant is Kozloduy-6 and the multilevel testing approach is employed. Measured data from NPP mixing experiments and code-to-code comparison will be used to validate the reactor vessel thermal hydraulic models.
  
V1000CT-2A
  
Plant measured data from VVER-1000 coolant mixing experiments will be used to test and validate vessel mixing models (CFD, coarse-mesh and mixing matrix). The task is to compare the various calculations with measured data, using specified vessel boundary conditions and core power distribution. The validated models will be used to analyze the main steam line break (MSLB) transient. The experiments include single loop cooling down or heating-up by disturbing the heat transfer in the steam generator (SG) through the steam valves, at low reactor power in the range of 5-14% and with all MCP in operation. They were conducted during the plant commissioning phase at Kozloduy-6.
  
V1000CT-2B
  
The transient to be analysed is initiated by a main steam line break in a VVER-1000 between the SG and the steam isolation valve (SIV), outside the containment. This event is characterised by large asymmetric cooling of the core, stuck control rods and large primary coolant flow variations.
  
Because of a possible return to power and criticality after reactor scram due to overcooling, the main objective of the study is to clarify the local 3D feedback effects depending on the vessel mixing. Special emphasis is put on testing 3D vessel thermal hydraulics models and the coupling of 3D neutronics/vessel thermal hydraulics. For the purpose of this benchmark two versions of the MSLB scenario are defined. The first scenario is taken from the current licensing practice and the second is derived from the original one using aggravating assumptions to enhance the code-to-code comparison.
  
The original scenario is based on conservative assumptions that maximise the consequences for a return to criticality:
  
1. The reactor is at the end of cycle (EOC) and at hot full power (HFP). The steam generator (SG) inventory is near the maximum value.
2. The two most reactive control assemblies remain stuck out of the core and are assumed to be in the affected sector. The scram worth is reduced supposing that one control rod bank does not fall in the core.
3. The feedwater pump trip setpoint is reached but the pumps fail to trip. The main feedwater flow to the faulted SG is terminated by closure of the feedwater isolation valve.
  
Other major assumptions are that off-site electric power is available; the main coolant pump (MCP) of the affected loop trips, the other MCP operate normally during the transient and the pressurizer is connected to the affected loop.
  
The second scenario is derived from the original one by assuming that all MCP remain in operation and the scram worth is additionally reduced through adjustment of the absorber cross sections.
  
The V1000CT-2 specifications are in preparation. They will include:
  
1. Case-specific geometry data;
2. Thermal hydraulic plant data from mixing experiments;
3. Thermal hydraulic plant data - MSLB related;
4. Neutron kinetics core specifications;
5. Macroscopic cross-section library;
6. Setpoints and interlocks;
7. Complete set of initial and boundary conditions needed for the exercises.
Definition of the V1000CT-2 Exercises
  
The benchmark consists of three exercises.
  
Calculation of VVER-1000 mixing experiments (2 sets)
  
Exercise 1. Comparison of CFD and coarse mesh calculations with measured data, using specified vessel boundary conditions and core power distribution.
  
VVER-1000 MSLB transient (2 scenarios)
  
Exercise 2. Coupled 3D neutronics/vessel TH simulation using specified vessel thermal hydraulic boundary conditions. Compare the different mixing models (coarse mesh, CFD, mixing matrix).
  
Exercise 3. Best estimate coupled simulation (plant, 3D vessel and core).
top ]
9. STATUS
Package ID Status date Status
NEA-1693/07 07-DEC-2011 Masterfiled restricted
top ]
10. REFERENCES
top ]
12. PROGRAMMING LANGUAGE(S) USED

No item found

top ]
15. NAME AND ESTABLISHMENT OF AUTHORS

Boyan Ivanov, Kostadin N. Ivanov
Nuclear Engineering Program
The Pennsylvania State University
University Park, PA  16802, USA
  
Pavlin Grudev and Malinka Pavlova
Institute of Nuclear Research and Nuclear Energy (INRNE)
Tsarigradsko shausse 72
1784 Sofia, Bulgaria
  
Vasil Hadjiev
Kozloduy Nuclear Power Plant
3321 Kozloduy, Bulgaria
  
Mr. Eric ROYER
Centre d'Etudes de Saclay
CEA/DEN/DM2S/SFME
91191 Gif-sur-Yvette Cedex
France
  
N. Kolev
Institute for Nuclear Research and Nuclear Energy (INRNE)
Tsarigradsko shausse 72
1784 Sofia, Bulgaria
-------
Sponsors:
US Department of Energy, Commissariat a l'Energie Atomique (CEA), and OECD Nuclear Energy Agency (NEA)
  
Co-organisers:
NEA Nuclear Science Committee (NSC), NEA Committee on Safety of Nuclear Installations (CSNI), and Atomic Energy Research (AER)
------
This package is available to participants of the V1000CT benchmarks.
top ]
16. MATERIAL AVAILABLE
NEA-1693/07
/Specification/ original and latest version of the specification & data
/Saclay/ papers distributed at the first V1000CT Workshop held at CEA Saclay
(12-13 May 2003)
/Sofia/ papers distributed at the second V1000CT Workshop held at Sofia,
Bulgaria (5-6 April 2004)
/Garching/ papers distributed at the third V1000CT Workshop held at Garching,
Germany (4-5 April 2005)
/Pisa/ papers distributed at the fourth V1000CT Workshop held at Pisa (24-25
April 2006)
/Vancouver/ papers and videoclips presented at the ad-hoc meeting during the
Physor-2006 conference (V1000CT4.5) (12 September 2006)
/Issy/ papers and videoclips presented at the fifth workshop (V1000CT5) (7 May
2007)
top ]
17. CATEGORIES
  • Y. Integral Experiments Data, Databases, Benchmarks

Keywords: VVER type reactor, coolants, main-steamline break, neutronics, pressurized water reactor, thermal hydraulics, transients.