The NEA has completed, under Nuclear Regulatory Commission (NRC) sponsorship, a PWR Main Steam Line Break (MSLB) Benchmark against thermal-hydraulic/neutron kinetics codes. Recently another OECD/NRC coupled code benchmark has been initiated for a BWR turbine trip (TT) transient. During the course of defining and co-ordinating the OECD/NRC PWR MSLB and BWR TT benchmarks, a systematic approach has been established to validate best estimate coupled codes. This approach employs a multi-level methodology that not only allows for a consistent and comprehensive validation process, but also contributes to determine additional requirements as well as to prepare a basis of licensing application of the coupled calculations for a specific reactor type and to develop a safety expertise in analysing reactivity transients. Professional communities have been established during the courses of these benchmark activities that allowed in-depth discussions of different aspects of assessing neutron kinetics modeling for a given reactor and how to implement best-estimate methodologies for transient analysis using coupled codes. The above examples demonstrate the benefit of establishing such international coupled standard problems for each type of reactor.
Further continuation of the above activities is the development of a VVER-1000 coolant transient (V1000CT) benchmark set, which 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 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)
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
Kozloduy Nuclear Power Plant
3321 Kozloduy, Bulgaria
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 oK. The temperature differences between the hot and cold legs for the loop with working MCPs vary between 8.3-11.5oK while the same temperature difference for the loop 3 with the MCP out of operation is –3.6oK. 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 ¼ 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.
PSU has completed for V1000CT-1 benchmark specifications and a cross-section library for coupled 3-D kinetics/thermal-hydraulics calculations in a format similar to the PWR MSLB Benchmark Final Specifications. Included in the benchmark specifications are:
The benchmark consists of three exercises:
Exercise 1 - Point Kinetics Plant Simulation
Exercise 2 - Coupled 3-D Kinetics/Core Thermal-Hydraulic Response Evaluation, and
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)
Institute for Nuclear Research and Nuclear Energy (INRNE)
1784 Sofia, Bulgaria
D. Caruge and E. Royer
Service Fluides Numériques, Modélisation et Etudes
Département Modélisation de Systèmes et Structures
Direction de l’Energie Nucleaire
CEA Saclay, 91191 Gif sur Yvette Cedex, France
V1000CT-2 Definition and Objectives
The recent OECD/NEA and AER coupled code benchmarks for light water reactors markedly contribute to the testing of best estimate coupled codes in reactivity transients. At the same time, these benchmarks and flow mixing studies indicate that further improvement of the mixing computation tools in the integrated codes is necessary. For this purpose the coolant mixing & main steam line break (MSLB) benchmark for VVER-1000 (V1000CT-2) was 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. The improved coupled codes will be tested in asymmetric MSLB transients with complex core-plant interactions.
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 and Kalinin-1, 2.
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:
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:
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).
Note: point kinetics plant simulation can be added on request of the participants.Further details are available in
For more information, please contact .
For more information on activities managed/supported by the NSC, please contact .
Last reviewed: 4 October 2017