Kozloduy Nuclear Power Plant Research Articles

Steam Line Break investigation at full power reactor for VVER-1000/V320

This paper presents the results of thermal-hydraulic calculation of “Steam Line Break” analysis at full power reactor 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. The RELAP5/MOD 3.2 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/MOD 3.2 at the Institute for Nuclear Research and Nuclear Energy–Bulgarian Academy of Sciences (INRNE–BAS), Sofia.The main purpose of the analysis is to estimate the parameters of the monitored plant which are used to identify symptoms that are used by operators to identify the plant's state and the critical safety function (CSF). The results of the thermal-hydraulic analyses have been used to assist KNPP specialists in analytical validation of EOPs.The performed analysis is based on a previously used bounding approach in analytical validation of SB EOPs. Based on this approach a list of scenarios has been performed, involving a different number of safety systems with or without operator actions. The presented thermal-hydraulic calculations of the accident scenarios involve the loss of CSF “Subcriticality” for VVER-1000/V320 units at KNPP.

Analytical validation of operator actions based on SAMG for VVER 1000 with ASTECv2r3 computer code

This paper presents the analytical validation of operator action based on severe accident management guidelines (SAMG) for Kozloduy NPP VVER1000 with severe accident computer code ASTECv2r3.The work is oriented on investigation of plant behavior during total loss of power and the operator actions performed based on strategies considered in severe accident management guidelines (SAMG) in Kozloduy nuclear power plant (KNPP). Using the SAMG strategies the operator depressurize primary circuit by gas removing system (YR) and try to cool down the reactor core by high pressure injection system (HPIS).The purpose of these analyses is to examine the possibility of keeping the core from further damage during a severe accident and to assess the likelihood of additional generation of hydrogen by additional flooding of the heated core. For this purpose it have been simulated a SBO scenario with injection of cold water by a high pressure pump (HPP) in cold leg at different core exit temperatures at 923K and 1253K. The selection of investigated analyses was based on severe accident management strategy of KNPP VVER1000.The presented work is important for analytical validation, verification, and further improvements of SAMG as well as for assessment of Level 2 probabilistic safety analyses (L2 PSA). The work was performed at the Institute for Nuclear Research and Nuclear Energy (INRNE) in the frame of severe accident research.

Investigation of main coolant pump trip problem in case of SB LOCA for Kozloduy Nuclear Power Plant, WWER-440/V230

This paper presents the results of thermal-hydraulic calculation of accident scenarios that involve the trip of main coolant pump (MCP) in case of Small break loss of coolant accident (SB LOCA) for WWER-440/V230 units at Kozloduy Nuclear Power Plant (KNPP), done in support of the development of Symptom Based Emergency Operating Procedures (SB EOPs) for this plant. The main purpose of these analyses is to show how the different time of MCP switching off results in primary inventory depletion in case of SB LOCA and it is reflect on peak cladding temperature. According to this, the SB LOCA scenario is regarded from the point of view of an inadequate core cooling. Therefore, the primary concern is Critical Safety Function (CSF) “Core cooling” and “Primary inventory”. High core residual heat, minimal safety injection flow and other initial conditions challenging the mentioned CSFs are the main particularities of the accepted scenarios.The RELAP5/MOD3.2 computer code has been used to perform the analyses in a WWER-440 Nuclear Power Plant (NPP) model. A model of WWER-440 based on Unit 4 of Kozloduy NPP has been developed for the system’s thermal-hydraulics code RELAP5/MOD3.2 at the Institute for Nuclear Research and Nuclear Energy – Bulgarian Academy of Sciences (INRNE-BAS), Sofia.

Health monitoring of NPP workers

The aim of the study is to assess the health status of Kozloduy Nuclear Power Plant personnel in association with their occupational exposure to low doses of ionising radiation (IR) as well as with other factors of their working environment and to determine the probability of the occupational exposure to be a reason for the diagnosed malignant diseases (Probability of Causation – PC). More than 1000 occupationally exposed workers and nonexposed administrative staff have been studied during 1993–2008. No statistically significant variation in the average cumulative dose was found between the healthy people and the persons with at least one disorder. The highest rate of morbidity was registered in persons with doses up to 1 mSv. The analysis of parameters of peripheral blood showed that at occupational lowdose exposure the haemopoietic system completely preserves the tri-linear cell proliferation capacity of the bone marrow. The analysis of results of PC demonstrated that this is spontaneously arising malignant diseases, regardless of condition of work.

Investigation of nuclear power plant behaviour at low power and cold conditions during an overpressurization in primary circuit

This paper discusses the results obtained during an investigation of WWER-1000 Nuclear Power Plant (NPP) behaviour at low power and cold conditions during an overpressurization in the primary circuit. The reference nuclear power plant for this analysis is Unit 6 at Kozloduy NPP (KNPP) site. The systems and equipment of the KNPP, Unit 6 operate according to the design requirements for the corresponding level of reactor power. In this paper is also presented an analysis of the additionally installed equipment for situations such as cold overpressurization. This equipment was mounted during the Modernization program in KNPP. In the paper is analysed the response of the new equipment together with other NPP available safety systems during the postulated transient in shutdown state. For the purpose of this analysis a RELAP5/MOD3.2 computer code has been used to simulate the transient for WWER-1000/V320 NPP model. A model of WWER-1000 based on Unit 6 of Kozloduy NPP has been developed for the RELAP5/MOD3.2 code at the Institute for Nuclear Research and Nuclear Energy – Bulgarian Academy of Sciences (INRNE-BAS), Sofia. The low power and cold conditions and the modifications after the modernization program, have been taken into account.The main purpose of this analysis is to estimate the parameters of the monitored plant which are used to identify symptoms that are used by operators to identify the plant’s state and the Critical Safety Function (CSF). The analysis is also used to define the timing needed to reach the following stages during the progression of processes in the reactor system:•Reaching the saturated temperature at the outlet of the assembly.•Beginning of reactor core uncover.•Heating up of fuel.•Defining the transition time between Emergency Operating Procedures (EOPs) and Severe Accident Management Guidance (SAMG) at temperature of 923.15K.•Restoring of water level in the core.•Defining the CSF “Integrity” status and the time of its loss.The results of the thermal–hydraulic analysis have been used to assist KNPP specialists in analytical validation of EOPs at low power and cold conditions.The performed analysis is based on a previously used bounding approach in analytical validation of Symptom Based EOPs (SB EOPs). The presented thermal–hydraulic calculations of the accident scenarios involve the loss of Critical Safety Function (CSF) “Integrity” for WWER-1000/V320 units at Kozloduy Nuclear Power Plant (KNPP). During the analysis is also discussed the behaviour of CSF “Core cooling”. The most specific in this analysis compared to the analyses of NPP accidents at full power is the lack of some important safety systems due to NPP regulations. Based on this approach a list of scenarios has been performed, involving a different number of safety systems with or without operator actions. The other specific characteristic of the accidents in NPP at low power and cold conditions is that even though all process are progressing significantly slower compared to full power, the operator actions should always bring the reactor system under safety conditions.

Pilot determinations of 3H and 14C in gaseous releases from Kozloduy nuclear power plant

Methods for routine assessment of 3H and 14C content in gaseous releases from ventilation stacks of Kozloduy Nuclear Power Plant (Bulgaria) were developed. Technique for correction of incomplete desorption of tritium from exposed silica gel was proposed. The distribution and the concentrations of both nuclides in various chemical forms were constantly monitored for a period of 1 year. The results for annual normalized gaseous discharges were assessed for the fifth unit at 173 GBq/(GW.a) for 3H and 369 GBq/(GW.a) for 14C, while for the sixth unit—3H—98 GBq/(GW.a) and 14C—289 GBq/(GW.a).

Investigation of primary to secondary leakage at low power during “Hydro-test” for VVER440/V230 at Kozloduy NPP

This paper presents the thermal–hydraulic investigation “Primary to secondary leakage at low power” for analytical validation of Symptom Based Emergency Operating Procedures (SBEOPs) at low power for VVER440/V230 units 3 and 4 at Kozloduy NPP. The main purpose of this analysis is to provide the response of monitored plant parameters to identify symptoms available to the operators and define timing for reaching the following stages during the development of processes in the reactor system, after primary to secondary leakage: • Reaching the saturated temperature at the outlet of the assembly. • Beginning of reactor core uncovery. • Define the time for reaching maximum level in the damaged SG. • Termination of natural circulation after decreasing of water level in reactor vessel below the hot nozzles. • Defining the time for opening SVs of damaged SG. • Define the CSF status and the time of loss of Critical Safety Functions. The thermal–hydraulic analysis for analytical validation of Symptom Based Emergency Operating Procedures (SBEOPs) at low power of reactor for VVER440/V230 units 3 and 4 at Kozloduy NPP is performed with RELAP5/MOD3.2 computer code. The RELAP5/MOD3.2 model of Kozloduy NPP VVER-1000 for investigation of operational occurrences, abnormal events, and design basis scenarios have been developed and validated in the Institute for Nuclear Research and Nuclear Energy – Bulgarian Academy of Sciences (INRNE–BAS) Sofia, and Kozloduy NPP. The model provides a significant analytical capability for the specialists working in the field of NPP safety.

Primary loop study of a VVER-1000 reactor with special focus on coolant mixing

The intention of this work is to demonstrate the ability of modern computational fluid dynamics (CFD) tools like ANSYS CFX to model technical facilities with complex geometry such as a reactor pressure vessel with its primary loops over all relevant scales from a range of approximately 1 mm up to a largest dimension of about 50 m. For this purpose, a detailed model of a VVER1000 reactor pressure vessel (RPV) and an extended model with simplified primary loops (Loop model) are presented. The primary loop components like steam generators and pumps are modelled by the outer shapes and additional source terms for energy and momentum exchange. The RPV model part is tested without primary loops; results are compared with former results obtained with a coarser model. The improved model shows less sensitivity on the discretization scheme, and recalculated hot leg temperatures transients are improved. As an example case for the Loop model, the swirl of flow patterns at the core inlet derived from experimental data obtained at the Kozloduy Nuclear Power Plant (Unit 6) is simulated. As progress to the RPV model the Loop model is able to predict a swirl, but due to the lack of technical details, like the pump impellers, the swirl is overestimated.

Probabilistic safety analyses developments resurgence in Bulgaria—Kozloduy nuclear power plant example

This paper presents some of the main technical features and insights of the Kozloduy nuclear power plant (NPP) units 5 and 6 probabilistic safety analysis (PSA) level 1. Probabilistic analyses and their applications in Bulgaria were given further impetus in recent years. More than 17 years after the first PSA study in Bulgaria in 1992 today probabilistic analyses receive increasing attention and application than ever before. The Bulgarian regulatory body (BNRA) is also interested in expanding their capability of reviewing and using PSA in plant safety assessments. In November 2008 within the framework of the program financed by European Union (PHARE), a project for assisting the BNRA in establishing the regulatory requirements on the base of PSA was completed. One of the objectives of this project was performance of the independent review of Kozloduy NPP units 5 and 6 PSA. This review was a new impulse for the authors to present in more details of Kozloduy NPP probabilistic assessment studies in the present paper.

Loss of ‘Core cooling’ at low power and cold condition of VVER-1000/V320

This paper presents the results of thermal–hydraulic calculation of accident scenarios that involve the loss of critical safety function (CSF) “Core cooling” for VVER-1000/V320 units at Kozloduy nuclear power plant done during the development of symptom based emergency operating procedures (SB EOPs) for this plant at low power and cold condition. The main purpose of this analysis is to provide the response of monitored plant parameters to identify symptoms available to the operators and define timing for reaching the following stages during the development of processes in the reactor system: ○ Reaching the saturated temperature at the outlet of the assembly; ○ Beginning of reactor core uncovery; ○ Heating up of fuel; ○ Defining the transition time between EOPs and SAMG at temperature of 923 K; ○ Restoring of water level in the reactor; ○ Defining the CSF “Core cooling” status and the time of its loss. The results of the thermal–hydraulic analyses have been used to assist KzNPP specialists in analytical validation of EOPs at low power and cold condition. 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. A model of VVER-1000 based on Unit 6 of KzNPP 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. The low power and cold condition and the modifications after the modernization program are taken into account.

Assessment of passive autocatalytic recombiners' capabilities in case of Zr oxidation during station blackout scenario for TVSA and TVSM fuel assemblies of WWER 1000

This paper presents the investigation of Passive Autocatalytic Recombiners' (PARs) capabilities for hydrogen recombination in case of Station blackout scenario. The assessment was performed for both types of WWER fuel assemblies – the old Modernistic type of fuel assemblies (TVSM) and recently installed new Alternative type of fuel assemblies (TVSA) in Kozloduy NPP. The main difference between both types of fuel assemblies is the different geometries, masses of internals materials as well as different burnable poisons. To investigate the PARs' capabilities it has been performed comparison of fuel behaviour of both types of fuel assemblies. To perform this analysis it has been used MELCOR “input model” for Kozloduy Nuclear Power Plant (KNPP) WWER-1000. The model was developed at the Institute for Nuclear Research and Nuclear Energy (INRNE) for investigation of severe accident scenarios. The model provides a significant analytical capability for the Bulgarian technical specialists, working in the field of the NPP safety, for analysis of core and containment damaged states and the estimation of radionuclides' release outside fuel cladding. To assess the PARs' capabilities it was used the acceptance criterion for containment integrity to be 8% hydrogen concentration. This criterion was based on the decision of RSK (Germany commission for reactor safety). Generally, based on the performed analyses it was made a conclusion that using both types of fuel assemblies it was not disturbance of PARs' capabilities and safety criterion of NPP.

Comparison of hydrogen generation for TVSM and TVSA fuel assemblies for water water energy reactor (VVER)-1000

This paper presents the results received during investigation of hydrogen generation for both types fuel assemblies—the old modernistic type of fuel assemblies (TVSM) and recently installed new one alternative type of fuel assemblies (TVSA) in case of severe accident. There are some differences between both types FAs. They have different geometry as well as different burnable poisons. To investigate behavior of new fuel assemblies during the severe conditions it have been performed comparison of fuel behavior of old type TVSM fuel assembly to new one TVSA. To perform this investigation it has been used MELCOR “input model” for Kozloduy Nuclear Power Plant (KNPP) VVER 1000. The model was developed by Institute for Nuclear Research and Nuclear Energy-Bulgarian Academy of Sciences (INRNE-BAS) for investigation of severe accident scenarios and Probabilistic Safety Analyses (PSA) level 2. The model provides a significant analytical capability for the Bulgarian technical specialists, working in the field of the NPP safety, for analysis of core and containment damaged states and the estimation of radionuclides release outside fuel cladding. It was accepted criteria for vessel integrity about hydrogen concentration to be 8%. This criterion was based on the decision of RSK (Germany commission for reactor safety). Generally based on the received results it was made conclusion that using both types of fuel assemblies it was not disturbance safety conditions of NPP.

Separation of strontium from calcium by the use of sodium hydroxide and its application for the determination of long-term background activity concentrations of 90Sr in 100 km area around Kozloduy Nuclear Power Plant (Bulgaria)

The method for the determination of 90 Sr which employs sodium hydroxide for the separation of strontium from calcium was further improved introducing the use of elevated temperatures. The results from 11-year study of background activity concentrations of 90 Sr in different environmental objects in 100 km zone around Kozloduy Nuclear Power Plant (Bulgaria) are presented as an application of the analytical method. The measured mean values are as follows: air precipitation – 0.0015±0.0009 Bq/(m 2. d), tap water – 0.0017±0.0012 Bq/L, soil – 1.90±1.26 Bq/kg, grass – 1.54±0.80 Bq/kg, milk – 0.023±0.012 Bq/L and for the Danube river: water – 0.0046±0.0026 Bq/L, bottom sediments – 0.64±0.60 Bq/kg, algae – 1.99±1.56 Bq/kg. The calculated transfer coefficients (soil-grass) are in the range of 0.33–0.84. Between 2 and 5 times reduction in actual background activities of 90 Sr is observed compared to 1972–1974.

Validation Cases of CATHARE 2 for VVER-1000 Main Steam Line Break Analysis

Recent coupled code benchmarks identified coolant mixing in the reactor vessel as an unresolved issue in the analysis of complex plant transients with reactivity insertion. Thus, Phase 2 of the OECD VVER-1000 Coolant Transient Benchmark (V1000CT-2) was defined. The benchmark includes calculation of vessel mixing tests and main steam line break (MSLB) analysis. The reference plant is Kozloduy-6 in Bulgaria. The general objective is the assessment of system codes for VVER safety analysis and specifically for their use in the analysis of reactivity transients. A specific objective is the testing of different scale mixing models (mixing matrix, multi-1D, coarse-3D and CFD), and analysis of MSLB transients with improved vessel thermal hydraulic models. The benchmark is sponsored by CEA-France and OECD and is jointly prepared by CEA and INRNE, in collaboration with the Kozloduy NPP, IRSN and PSU. This paper summarizes CATHARE2 code assessment calculations using multi-1D vessel thermal hydraulics with cross flow. Test cases are the OECD V1000CT-1 pump start-up benchmark and the V1000CT-2 benchmarks. Emphasis is put on vessel mixing aspects. Separate effects in the lower plenum as well as component and integral system tests are considered. The comparison shows that a six-sector vessel mixing model informed by plant data or validated CFD calculations in the initial state was able to correctly reproduce the channel average temperatures at the core inlet as well as the vessel outlet temperatures. Testing at system level including code-to-experiment and CATHARE-ATHLET comparison shows that the considered CATHARE VVER-1000 system model is capable of MSLB simulation.

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.

Simulation of mixing effects in a VVER-1000 reactor

This work has been performed in the framework of the OECD/NEA thermalhydraulic benchmark V1000CT-2. This benchmark is related to fluid mixing in the reactor vessel during a MSLB accident scenario in a VVER-1000 reactor. Coolant mixing in a VVER-1000 V320 reactor was investigated in plant experiments during the commissioning of the Unit 6 of the Kozloduy nuclear power plant. Non-uniform and asymmetric loop flow mixing in the reactor vessel has been observed in the event of symmetric main coolant pump operation. For certain flow conditions, the experimental evidence of an azimuthal shift of the main loop flows with respect to the cold leg axes (swirl) was found. Such asymmetric flow distribution was analyzed with the Trio_U code. Trio_U is a CFD code developed by the CEA Grenoble, aimed to supply an efficient computational tool to simulate transient thermalhydraulic turbulent flows encountered in nuclear systems. For the presented study, a LES approach was used to simulate turbulent mixing. Therefore, a very precise tetrahedral mesh with more than 10 million control volumes has been created. The Trio_U calculation has correctly reproduced the measured rotation of the flow when the CAD data of the constructed reactor pressure vessel where used. This is also true for the comparison of cold leg to assembly mixing coefficients. Using the design data, the calculated swirl was significantly underestimated. Due to this result, it might be possible to improve with CFD calculations the lower plenum flow mixing matrices which are usually used in system codes.

Calculation of the VVER-1000 coolant transient benchmark using the coupled code systems DYN3D/RELAP5 and DYN3D/ATHLET

Plant-measured data provided by the OECD/NEA VVER-1000 coolant transient benchmark programme were used to validate the DYN3D/RELAP5 and DYN3D/ATHLET coupled code systems. Phase 1 of the benchmark (V1000CT-1) refers to an experiment that was conducted during the commissioning of the Kozloduy NPP Unit 6 in Bulgaria. In this experiment, the fourth main coolant pump was switched on whilst the remaining three were running normal operating conditions. The experiment was conducted at 27.5% of the nominal level of the reactor power. The transient is characterized by a rapid increase in the primary coolant flow through the core, and as a consequence, a decrease of the space-dependent core inlet temperature. The control rods were kept in their original positions during the entire transient. The coupled simulations performed on both DYN3D/RELAP5 and DYN3D/ATHLET were based on the same reactor model, including identical main coolant pump characteristics, boundary conditions, benchmark-specified nuclear data library and nearly identical nodalization schemes. In addition to validation of the coupled code systems against measured data, a code-to-code comparison between simulation results has also been performed to evaluate the respective thermal hydraulic models of the system codes RELAP5 and ATHLET.

Neutron fluence attenuation through VVERs and mock-ups of reactor pressure vessels

The calculational results of neutron flux attenuation through the Reactor Pressure Vessel (RPV) of VVER type of reactors of Kozloduy NPP and their LR-0 mock-ups have been compared. The neutron flux with energy above 0.5 MeV has been considered. For the VVER-440 the neutron flux attenuation through the RPV and its mock-up has consisted within the calculation uncertainty. For the VVER-1000 the neutron flux attenuation through the RPV and its mock-up has consisted also within the calculation uncertainty except the region behind RPV where the attenuation difference has been significant. It has been demonstrated that this difference is due to the difference in the biological shielding materials. The mock-ups of VVER-1000 and VVER-440 types of reactor could be used for experimental simulating of the neutron flux attenuation in the nuclear power plants taking into account the existing peculiarity of the biological shielding and the attenuation dependence on the RPV azimuth.

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.