VVER-1000 Nuclear Power Plant Research Articles

A simulation of a steam generator tube rupture in a VVER-1000 plant

In the operation of nuclear power plants, it is very important to evaluate different accident scenarios in actual plant conditions. One of the main accidents is steam generator tube rupture (SGTR), in the field of nuclear safety. In this research the variation of thermo-hydraulics parameters in a primary loop under SGTR accidents at Bushehr VVER-1000 Nuclear Power Plant (NPP) is analyzed by RELAP5/MOD 3.2 thermo-hydraulics code. A model of Bushehr unit 1 has been developed based on RELAP5/MOD 3.2 at Shiraz University in Iran. The results obtained by this code are compared with the Preliminary Safety Assessment Report (PSAR) data of Bushehr NPP. The DINAMICA-97 code was used to evaluate the SGTR accident in Bushehr PSAR by Federal Agency on Nuclear Energy of Russia. In simulation of this accident, it was supposed that after the establishment of a steady-state condition in the system, primary-to-secondary coolant leakage is as a result of an instantaneous break happened with equivalent diameter of 100 mm in the area of lower row heat exchanging tubes in one of the fourth steam generators (SG#2). Loss of off-site power and two diesel generators of in-site power in loops 1and 2 of the primary cycles follow this event. The results show that after initiating the accident, the pressure above the core decreases rapidly from 16 MPa to 8 MPa during 130 s and finally, it will stabilize to about 7.1 MPa and it becomes practically equal to the secondary pressure and it leads to stop leakage completely at the end of transient. The coolant temperature at the reactor inlet decreases from 566 K to 473 K during the transient time and the cooldown of the coolant of the primary loop is performed. The primary-to-secondary coolant leakage reduces from 800 kg/s to approximately 0.0 kg/s during the first 150 s of the accident. As seen from the results, the thermo-hydraulics code correctly predicts the behavior of the main plant parameters in comparison with the reported PSAR data.

Systematic approach for the analytical validation of Kozloduy NPP, VVER-1000/V320 symptom based emergency operating procedures

This paper presents the development and application of methodology used in analytical validation of emergency operating procedures (EOPs) for Kozloduy Nuclear Power Plant (KNPP), VVER-1000/V320 units. EOPs provide generic guidance to a reactor plant operator in maneuvering the plant to a safe, stable condition in the event of an unexpected plant transient or emergency. These procedures have been analytically validated in order to provide technical justification that the prescribed operator actions are reasonable, effective and prudent. This evaluation is accomplished by systematically evaluating the procedures using specialized thermal-hydraulic computer codes designed for nuclear reactor plant simulation. Thermal-hydraulic computer code calculations have been performed to simulate the symptoms presented to the operator to diagnose challenges to the critical safety functions (CSFs). The accomplished emergency operating procedures' (EOPs') analyses are designed to provide the response of monitored plant parameters to identify symptoms available to the operators, timing of the loss of critical safety functions and timing of operator actions to avoid the loss of critical safety functions or core damage. The principal acceptance criteria for EOPs are averting the onset of core damage. The RELAP5/MOD3.2 computer code has been used in performing the analyses in a VVER-1000 Nuclear Power Plant (NPP) model. EOPs' analysis methodology assumes realistic boundary conditions in validating operator actions. A model of VVER-1000 based on Unit 6 of KNPP has been developed for the system's thermal-hydraulic code RELAP5/MOD3.2 at the Institute for Nuclear Research and Nuclear Energy of Bulgarian Academy of Sciences (INRNE-BAS), Sofia, and was used for simulating the bounding scenarios. The work was possible through participation of leading specialist from KNPP and with the assistance of Pacific Northwest National Laboratory (PNNL), under the International Nuclear Safety Program (INSP) of the United States Department of Energy.

RELAP5/MOD3.2 blackout investigation for validation of EOPs for KNPP VVER-1000/V320

This paper presents the results of thermal–hydraulic calculation of the “Blackout” analysis for VVER-1000/V320 units at Kozloduy nuclear power plant (KNPP), done during the development of symptom based emergency operating procedures (SB EOPs) for this plant. These kinds of analyses are designed to provide the response of monitored plant parameters to identify symptoms available to the operators, timing of the loss of critical safety functions and timing of operator actions to avoid the loss of critical safety functions of core damage. The results of the thermal–hydraulic analyses have been used to assist KNPP specialists in analytical validation of EOPs. The principal acceptance criteria for EOPs are averting the onset of core damage. The response of the VVER-1000/V320 nuclear steam supply system to the loss of all ac power initiating event is analyzed to identify the behavior of important VVER process variables. A limited set of potential operator actions have been evaluated. An investigation of plant response to different combinations of component availabilities in which the operator can be effective in averting core melt is presented in the paper. The RELAP5/MOD3.2 computer code has been used in performing the analyses in a VVER-1000 nuclear power plant (NPP) model. A model of VVER-1000 based on Unit 6 of Kozloduy NPP has been developed for the systems thermal–hydraulics code RELAP5/MOD3.2 at the Institute for Nuclear Research and Nuclear Energy – Bulgarian Academy of Sciences (INRNE – BAS), Sofia and was used for simulation of “Blackout” transient. To analyze blackout transient for VVER-1000/V320, the acceptance criterion which was used was fuel cladding temperature not more than 1473 K (1200 °C). This report was possible through the participation of leading specialists from Kozloduy NPP and with the assistance of Pacific Northwest National Laboratory (PNNL), under the International Nuclear Safety Program (INSP) of the United States Department of Energy.

ICONE15-10327 VVER SAFETY ANALYSIS BY THE COUPLED CODE ATHLET/BIPR-VVER

The paper describes the modelling features for a VVER-1000 nuclear power plant by the coupled system code ATHLET/BIPR-VVER. Special attention is paid on the reactor pressure vessel model and its nodalization schema. They play an important role for the realistic prediction of the fluid mixing phenomena in case of asymmetric transients. For such a transient - simultaneous isolation of one steam generator from the steam line and the feed water supply at low power level - validation to measured parameters are presented and analysed. The results of the simulation are compared with experimental local and integral plant loop data from NPP Kozloduy Unit 6 collected during the plant-commissioning phase. Additional examples of coupled code application for VVER-1000 reactor applying detailed reactor pressure vessel nodalization are also presented, choosing Exercise 3 of the CEA-NEA/OECD VVER-1000 Coolant Transient Benchmark (V1000CT-2) - a main steam line break.

Sensitivity calculations of PRZ water level during the Natural Circulation test at Unit 6 of Kozloduy NPP

This paper presents the results of thermal-hydraulic sensitivity calculation of the “Natural Circulation” test at Unit 6, Kozloduy NPP. The RELAP5/MOD3.2 computer code has been used to simulate the Natural Circulation in a VVER-1000 Nuclear Power Plant (NPP) model. The results are an important part of the validation of the RELAP5/MOD3.2 code. A model of VVER-1000 based on Unit 6 of Kozloduy NPP has been developed for the systems thermal-hydraulics code RELAP5/MOD3.2. This model was developed at the Institute for Nuclear Research and Nuclear Energy – Bulgarian Academy of Sciences (INRNE-BAS), Sofia and was used for simulation of the behavior of the real VVER-1000 NPP during the Natural Circulation test. The data included steady-state and operational transient results. The purpose of the experiment, which was done according to the “Program for investigation of primary side Natural Circulation of coolant on NPP”, was to explicitly establish the response of the plant under primary loop Natural Circulation conditions. The possibility to remove approximately 150 MW using the Natural Circulation was also demonstrated during the test. The paper discusses the influence of Pressurizer nodalization on calculation of Pressurizer water level. The comparisons between the RELAP5 results and the test data indicate good general agreement. The report also contains a brief summary of the main NPP systems included in the RELAP5 VVER model and the NC transient sequences. This report was possible through the participation of leading specialists from Kozloduy NPP and with the assistance of Argonne National Laboratory, under the International Nuclear Safety Program (INSP) of the United States Department of Energy.

Analysis of the VVER1000 coolant trip benchmark using the coupled RELAP5/PARCS code

Incorporation of full three-dimensional models of the reactor core into system thermal–hydraulic transient codes allows better estimation of interactions between the core behavior and plant dynamics. Considerable efforts have been made in various countries and organizations to verify and validate the capability of the so-called coupled codes technique. For these purposes appropriate Light Water Reactor (LWR) transient benchmarks based upon programmed transients performed in Nuclear Power Plants (NPP) were recently developed on a higher ‘best-estimate’ level. The reference problem considered in the current framework is a Main Coolant Pump (MCP) switching-on transient in a VVER1000 NPP. This event is characterized by a positive reactivity addition as consequence of the increase of the core flow. In the current study the coupled RELAP5/PARCS code is used to reproduce the considered test. Results of calculation were assessed against experimental data and also through the code-to-code comparison.

06/01632 Simulation of loss-of-flow transient in a VVER-1000 nuclear power plant with RELAP5/MOD3.2: Grudev, P. and Pavlova, M. Progress in Nuclear Energy, 2004, 45, (1), 1–10

06/01632 Simulation of loss-of-flow transient in a VVER-1000 nuclear power plant with RELAP5/MOD3.2: Grudev, P. and Pavlova, M. Progress in Nuclear Energy, 2004, 45, (1), 1–10

RELAP5/MOD3.2 sensitivity calculations of loss-of-feed water (LOFW) transient at Unit 6 of Kozloduy NPP

This paper provides a comparison between the real plant data obtained by Unit 6 of Kozloduy nuclear power plant (NPP) during the loss-of-feed water (LOFW) transient and the calculation results received by RELAP5/MOD3.2 computer model of the same NPP unit. RELAP5/MOD3.2 computer model of the VVER-1000 has been developed at the Institute for Nuclear Research and Nuclear Energy-Bulgarian Academy of Sciences (INRNE-BAS) based on Unit 6 of Kozloduy NPP. This model has been used for simulation the behavior of the real VVER-1000 NPP during the LOFW transient. Several calculations have been provided to describe how the different boundary conditions reflect on the prediction of real plant parameters. This paper discusses the results of the thermal–hydraulic sensitivity calculations of loss-of-feed water transient for VVER-1000 reactor design. The report also contains a brief summary of the main NPP systems included in the RELAP5 VVER model and the LOFW transient sequences. This report was possible through the participation of leading specialists from Kozloduy NPP and with the assistance of Argonne National Laboratory (ANL) for the United States Department of Energy (US DOE), International Nuclear Safety Program (INSP).

05/02505 Proliferation resistance potential and burnup characteristics of an equilibrium core of novel natural uranium fueled nuclear research reactor

During the development of Symptom Based Emergency Operating Procedures (SB-EOPs) for VVER-1000/V320 units at Kozloduy Nuclear Power Plant (NPP), a number of analyses have been performed using the RELAP5/MOD3.2 computer code. One of them is “Investigation of reactor vessel YR line capabilities for primary side depressurization during the Total Loss of Feed Water (TLFW)”. The main purpose of these calculations is to evaluate the capabilities of YR line located at the top of the reactor vessel for primary side depressurization to the set point of High Pressure Injection System (HPIS) actuation and the abilities for successful core cooling after Feed&Bleed procedure initiation. For the purpose of this, operator action with “Reactor vessel off-gas valve – 0.032 m” opening has been investigated. RELAP5/MOD3.2 computer code has been used to simulate the TLFW transient in VVER-1000 NPP model. This model was developed at Institute for Nuclear Research and Nuclear Energy – Bulgarian Academy of Sciences (INRNE-BAS), Sofia, for analyses of operational occurrences, abnormal events, and design basis scenarios. The model provides a significant analytical capability for the specialists working in the field of NPP safety.

05/02511 Optimization of 60Co production using neutron flux trap in the Tehran research reactor

This paper provides a comparison between the real plant data obtained by Unit 6 of Kozloduy nuclear power plant (NPP) during the loss-of-feed water (LOFW) transient and the calculation results received by RELAP5/MOD3.2 computer model of the same NPP unit.RELAP5/MOD3.2 computer model of the VVER-1000 has been developed at the Institute for Nuclear Research and Nuclear Energy-Bulgarian Academy of Sciences (INRNE-BAS) based on Unit 6 of Kozloduy NPP. This model has been used for simulation the behavior of the real VVER-1000 NPP during the LOFW transient. Several calculations have been provided to describe how the different boundary conditions reflect on the prediction of real plant parameters.This paper discusses the results of the thermal–hydraulic sensitivity calculations of loss-of-feed water transient for VVER-1000 reactor design. The report also contains a brief summary of the main NPP systems included in the RELAP5 VVER model and the LOFW transient sequences.This report was possible through the participation of leading specialists from Kozloduy NPP and with the assistance of Argonne National Laboratory (ANL) for the United States Department of Energy (US DOE), International Nuclear Safety Program (INSP).

05/02049 Simulation of loss-of-flow transient in a VVER-1000 nuclear power plant with RELAP5/MOD3.2

05/02049 Simulation of loss-of-flow transient in a VVER-1000 nuclear power plant with RELAP5/MOD3.2

RELAP5/MOD3.2 investigation of reactor vessel YR line capabilities for primary side depressurization during the TLFW in VVER1000/V320

During the development of Symptom Based Emergency Operating Procedures (SB-EOPs) for VVER-1000/V320 units at Kozloduy Nuclear Power Plant (NPP), a number of analyses have been performed using the RELAP5/MOD3.2 computer code. One of them is “Investigation of reactor vessel YR line capabilities for primary side depressurization during the Total Loss of Feed Water (TLFW)”. The main purpose of these calculations is to evaluate the capabilities of YR line located at the top of the reactor vessel for primary side depressurization to the set point of High Pressure Injection System (HPIS) actuation and the abilities for successful core cooling after Feed&Bleed procedure initiation. For the purpose of this, operator action with “Reactor vessel off-gas valve – 0.032 m” opening has been investigated. RELAP5/MOD3.2 computer code has been used to simulate the TLFW transient in VVER-1000 NPP model. This model was developed at Institute for Nuclear Research and Nuclear Energy – Bulgarian Academy of Sciences (INRNE-BAS), Sofia, for analyses of operational occurrences, abnormal events, and design basis scenarios. The model provides a significant analytical capability for the specialists working in the field of NPP safety.

VVER-1000 Model Assessment Using RELAP5-3D

Detailed thermal-hydraulic analysis of commercial power reactors requires modeling of complex multidimensional thermal, fluids, and neutronic behavior. One code that has three-dimensional (3-D) thermal-hydraulic and neutronic capabilities is RELAP-3D. A comparison of RELAP-3D predictions to experimental data obtained during start-up of the Kozloduy VVER-1000 nuclear power plant in Bulgaria is presented here. The experiment has distinguishable 3-D hydraulics effects that allow for code model verification and reasonable agreement with the experimental results obtained.The transient investigated was conducted at 29% reactor power, and it was initiated from a steady state where three out of four reactor coolant pumps were operating. The transient consisted of the start-up of the nonoperating pump. Simulation results were compared to both temperature and pump data from the experiment. Temperature predictions compared reasonably well to the experimental data; however, discrepancies existed between predicted and experimental pump head values.

Simulation of loss-of-flow transient in a VVER-1000nuclear power plant with RELAP5/MOD3.2

A RELAP5/MOD3.2 model of a VVER-1000 (V-320 model nuclear power plant was updated, improved and validated against actual plant data. The data included steady-state and operational transient results from unit 6 of the Kozloduy nuclear power plant. The operational transient was initiated by a loss of flow at partial power conditions caused by the trip of a main coolant pump without reactor scram. The assessment of the model has been performed in two stages: an initial validation against steady-state plant data and then a transient validation by comparison to operational transient data. The comparison between the plant data and the results of the calculation proved the adequacy of the model and demonstrated the capability of the code to reproduce the evolution of the main plant parameters.

Development of a RELAP5-3D Multidimensional Model of a VVER-1000 NPP for Analysis of a Large-Break Loss-of-Coolant Accident

In order to analyze the benefits of the multidimensional hydrodynamic modeling capability of the RELAP5-3D system code for the VVER-1000 nuclear power plant (NPP), a three-dimensional (3-D) model of the core, downcomer, and lower plenum has been created to replace the NPP one-dimensional (1-D) counterparts in a complete plant model. This multidimensional model has been validated with plant operational data and other computer simulations of a thermal-hydraulic transient. The simulated transient considered was a large-break loss-of-coolant accident (LB LOCA).A validated, 1-D control model of the NPP, for the study of the effects of mixed oxide fuel, was modified to include a standard fuel loading of UO2. The development of the 3-D sections of the reactor vessel consisted of ensuring geometrical fidelity with the design of the modeled plant, the Balakovo Unit 4 NPP in Saratov, Russia. A stable operational steady state was obtained and the calculated plant conditions compared well with the design values of the Balakovo plant. Transient results verified that the simulated thermal-hydraulic conditions of the multidimensional model agreed well with both the control and analyses that have been performed separately from this study.It was found that the multidimensional model has shown a reduction in the calculated hot-spot peak-clad temperature (PCT) during the blowdown stage of a LB LOCA and an increase in PCT during the reflood stage. A preliminary uncertainty analysis of the PCT during blowdown stage was performed using a response surface method of the Code Scaling, Applicability, and Uncertainty Method and a significant number of relevant input variables. From the preliminary analysis, the PCT reduction during blowdown appears to be significant, but a further, more detailed analysis should be performed, along with an uncertainty analysis of the PCT during the reflood stage.The enhanced depiction of the flow patterns and temperature distributions in the transient situation allowed the user further understanding of the thermal-hydraulic conditions throughout the transient. The developed model proved to be suitable for analysis of the VVER-1000 plant, but to further the applicability of the model, a 3-D kinetics model of the neutronics and 3-D hydrodynamic models of the horizontal steam generators should be included.

Simulation of Two Loss-of-Flow Transients in a VVER-1000 Nuclear Power Plant with RELAP5/MOD3.2 System Code

A RELAP5/MOD3.2 model of a VVER-1000 (V-320 model) nuclear power plant was updated, improved, and validated against actual power plant data. The data included steady-state and operational transient results from unit 5 of the Kozloduy nuclear power plant in Bulgaria. The first operational transient was initiated by a loss of flow at partial power conditions caused by the trip of a main coolant pump without reactor scram. The second operational transient was a continuation of the first, with the trip of a second main coolant pump.The assessment of the model has been performed in two stages: an initial validation against steady-state plant data and then a transient validation by comparison to operational transient data. The comparison between the plant data and the results of the calculations proved the adequacy of the model and demonstrated the capability of the code to reproduce the evolution of the main plant parameters.

Advanced Designs and Operating Strategies to Enhance the Safety, Operability, and Efficiency of VVER-1000 Reactors

Advanced fuel, burnable absorber, and control rod designs along with advanced fuel management and power distribution control strategies will be implemented in the first operating cycle of the Czech Republic`s Temelin VVER-1000 nuclear power plants. These improvements increase safety margins, enhance operability, and improve fuel efficiency. The Westinghouse VVANTAGE 6 fuel assembly design incorporates many proven advanced fuel and core design features used extensively in western pressurized water reactors. The fuel assembly incorporates mixing vane structural grids, radial enrichment zoning, ZrB{sub 2} integral fuel burnable absorbers, axial blankets, and Zircaloy guide thimbles and structural grids. Low-leakage loading patterns are also used to reduce radial neutron leakage. The rod cluster control assembly (RCCA) design incorporates two absorber materials. The absorber tip uses silver-indium-cadmium material while the remainder of the absorber material is B{sub 4}C enriched in {sup 10}B. This design increases control rod worth as well as the usable lifetime of RCCA. The Westinghouse constant axial offset control operating strategy, improved RCCA design, modified RCCA overlap, and replacement of part-length RCCA by full-length RCCA are used to improve the axial power distribution control capability for VVER-1000 reactors. These design improvements provide thermal margin benefits, increase shutdown margin by almost 1.0% {Delta}{rho},more » reduce fuel cycle costs by nearly 30%, and improve axial power distribution control.« less

Controlling VV�R-1000 power distribution during power level changes

Success in executing optimum methods to control the neutron and power distribution when operating conditions are changed in VVER-1000 nuclear power plants requires a substantial change in the algorithms and instrumentation currently used in the control system. Based on the experience of developing such systems and on model calculations, the control algorithms examined here bring out the following basic questions for changing the planned solutions: (1) The increase in the number of control-rod clusters in the core in order both to increase the total effectiveness for reaction prevention and to use complete submersion of control-rod groups during the level change. (2) Development of on-line computer programs to optimize the level changes. (3) Development and application of the technical equipment to realize this control system.