In the event of an earthquake, it is possible a number of pipes in the steam generator (SG) to rupture, causing a leak from the primary to the secondary side. Due to the sharp increasing of the secondary side pressure in the damaged SG, the conditions for opening of the Steam Dump Valve to the Atmosphere (BRU-A) will be reached. In the case of opening the BRU-A, it is necessary to study the situation in which the valve will remain in the open position. The primary coolant, will leak irreversibly into the environment, leaving the reactor system without the ability to be cooled after a certain time. The calculation is performed until reaching $1200{ }^{\circ} \mathrm{C}$, when steam zirconium reaction will become self-sustaining. The investigated NPP is a generic VVER 1000 /V320 type. The analysis was performed with the integral computer code RELAP5 mod3.3. The purpose of the analysis is to estimate the minimum number of damaged steam generator tubes that will lead to opening of BRU-A. The other purpose is to analyze the behavior of the nuclear power plant and estimate the minimum time that operators have to perform recovery actions such as “bleeding and feeding” of reactor coolant system.

Abstract This paper presents a comparison of the results from simulation of „Main steam line break (MSLB) “accident using RELAP5/mod3.3 and TRACE50p5 computer codes. The work is based on the VVER-1000 MSLB OECD benchmark. The main purpose of this work was to demonstrate the capability of the TRACE computer code for modelling of 3D thermal hydraulic reactor vessel at the integral VVER 1000 model and to predict correctly the plant respond and spatial disturbance in the reactor vessel. In addition, this investigation was focused on validation of the correct predictability of the important integral parameters in VVER1000 TRACE 50p5 models. The other objectives were to provide the elements for the safety, to validate the safety analysis tools and to improve codes and methods for VVER comprehensive safety assessment. The investigated scenario is a “Main steam line break (MSLB) with inside diameter (ID) 580 mm” in a VVER1000 between the steam generator (SG) and the steam isolation valve (SIV), outside of the containment. This event is characterized with significant spatial effects in the core caused by asymmetric cooling and assumed stuck-out of control rods after SCRAM. One of the major concerns for this case is the possible return of power and criticality after SCRAM, as a result of the reactor core overcooling. The reactor state is at the beginning of 1st fuel campaign. For the purpose of the investigation, the simulation of MSLB accident with RELAP5/MOD3.3 and TRACE50p5 computer codes with modelling of reactor physics using point kinetics has been performed initially. Additionally, the 3D thermal hydraulic reactor vessel model with a point kinetic physics has been developed at the integral VVER1000 TRACE 50p5 model. The newly developed integral model of VVER1000 with 3D reactor vessel of TRACE50p5 has been validated with simulation of MSLB accident by code to code comparison with already validated VVER1000 models of RELAP5/MOD3.3 and 1D TRACE50p5 codes. The comparison between the results predicted by the codes demonstrates good agreement, except some of the parameters. The newly developed integral VVER1000 with 3D vessel model of TRACE50p5 predicts reasonable results. The predicted results by used computer codes demonstrate the stability of the reactor system during the accident progression of “Main steam line break”.

This paper presents a study of nuclear energy development in Bulgaria and its contribution to the country 's energy sector. Current status and future prospects of the nuclear power industry have been discussed. Currently, two 1000 MW nuclear power units - Units 5 and 6 with WWER-1000 reactors are in operation at Kozloduy NPP site. Nuclear energy is one of the main pillars of the electricity producing sector with 32, 6 % of current national energy mix. Energy strategy and vision for the development of the power sector in Bulgaria envisage nuclear energy to be supported institutionally as a promising resource for generation of emission -free electricity and to ensure a sustainable energy mix. Extending the operational lifetime of the available capacities, as well as, the construction of new nuclear capacities have been considered to be of strategic importance for preserving energy security, reducing greenhouse gas emissions and reducing dependence on imports of energy resources in a cost-effective and competitive manner. In addition, this investigation is focused on the efforts that have been invested toward current nuclear reactors lifetime extension, thermal power uprate and diversification of fresh nuclear fuel supplies, as well as, construction of new nuclear capacities to ensure the future development of the nuclear industry in Bulgaria. Some key drivers and challenges in nuclear power sector for future development were analysed. Large size nuclear reactors and SMRs in Bulgarian context were considered and the outcomes indicates that SMRs and large reactors generation III/III + should not be opposed but SMRs could be considered as a complementarity technology to the existing large-scale reactors. Many of the currently developed SMRs conceptual designs meet the expectations for Generation IV reactors safety level but in-depth knowledge of the specific technology is required to assess and ensure safety in accordance with the Bulgarian regulatory framework which is oriented towards light water reactors.

This paper presents a comparative analysis of the Main Steam Line Break (MSLB) in a VVER-1000 reactor simulated with RELAP5 using Point Kinetics and the coupled code TRACE5-P05/PARCS using 3D kinetics. In the MSLB-scenario, it is assumed that the main steam line break of 580 mm inner diameter is located between the steam generator (SG) and the steam isolation valve (SIV), outside the containment. In a MSLB, a non-symmetric overcooling of the primary coolant takes place leading to a positive reactivity insertion. Hence, the main safety concern is to assess if the core may become critical despite SCRAM and it there is a considerable power increase (return-to-power). This paper will discuss the capabilities of different computational approaches to simulate the VVER-1000 plant behaviour during a MSLB; one approach based on 1D thermal hydraulics and Point Kinetics while the other one based on 3D thermal hydraulics of the reactor pressure vessel (RPV) and 1D thermal hydraulics for the remaining plant components based on a 3D neutron kinetics model. The analyses are performed for Beginning of Cycle (BOC) conditions i.e., with a fresh core loading when the plant is operated at nominal power. The neutron kinetic parameters for the RELAP5 Point Kinetics model were generated PARCS for the BOC assuming a boron concentration of 1630 ppm. The respective 2 energy group homogenized cross section libraries in PMAXS-format were generated by KIT using the SERPENT2 code. The investigations were performed in the frame of CAMIVVER-project, which focus was the assessment and development of reliable neutron physical and system thermal hydraulic models for safety evaluations of VVER1000 reactors. The comparative analysis for the MSLB has shown that both applied codes are able to qualitatively predicts the plant behaviour under MSLB-conditions in similar manner. Differences are caused by the different approach to represent the core and RPV followed by RELAP5 and TRACE5.05/PARCS as expected.

This article concerns an uncertainty and sensitivity investigation of certain parameters in the In-Vessel Melt Retention (IVMR) test case for Water-Water Energetic Reactor – 1000/v320 (WWER-1000/v320). It has been used the ICARE and CESAR modules of ASTECv2.2b severe accident computer code to describe the basic parameters behaviour and the main phenomena arising during the IVMR in WWER1000 reactor design. The external vessel water cooling has been chosen for IVMR strategy. First, one stand-alone calculation have been done to account the most heat loaded segment from the vessel. After, the uncertainties in two parameters in the deterministic calculation have been investigated additionally to account their influence on the heat flux on this segment. An opportunity for an uncertainty and sensitivity analyses gives SUNSET (Statistical UNcertainty and Sensitivity Evaluation Tool) software which is a part of ASTEC computer code. The SUNSET computational tool developed by IRSN, is a statistical tool designed for uncertainty and sensitivity analysis of mathematical or physical models like computer codes. It have been investigated an influence of the different pressure values inside the vessel and influence of the different temperature values of outside cooling water on the two basic output parameters: heat flux on the most heat loaded segment of the vessel and the minimal vessel wall thickness of this segment. It was found out to what extent each one of the both input parameters effect on the studied output parameters.

This paper presents the comparison of the results from simulation of „Main steam line break“ accident using RELAP5/mod3.3 and TRACE50p5 computer codes. The work is based on the OECD VVER-1000 MSLB benchmark. The main purpose of work was to demonstrate capability of the 3D reactor vessel model of TRACE in integral VVER 1000 model and correctly to predict the plant respond on spatial disturbance in the reactor vessel. In addition, this investigation is focused on validation of correct predictability of the important integral parameters in VVER1000 TRACE 50p5 models. The other objective was to p rovide the elements for the safety, validated safety analysis tools, also improvement of codes and methods for VVER comprehensive safety assessment and to support VVER fuel development and qualification activities. The investigated scenario is a “Main steam line break (MSLB) with ID 580” in a VVER1000 between the steam generator (SG) and the steam isolation valve (SIV), outside the containment. This event is characterized with significant space-time effects in the core caused by asymmetric cooling and assumed stuck-out control rods after scram. One of the major concerns for this case is the possible return of power and criticality after scram, as a result of overcooling. The reactor state is at the beginning of 1-st fuel cycle.For the purpose of investigation, it has been performed initially simulation of MSLB accident with RELAP5/MOD3.3 and TRACE50p5 computer codes with modelling of reactor physics using point kinetics. Additionally, it has been developed the 3D thermal hydraulic reactor vessel of VVER1000 with TRACE 50p5 model. The 3D reactor vessel model of VVER1000 TRACE50p5 have been validated with simulation of MSLB performing code to code comparison with already validated VVER1000 models of RELAP5/MOD3.3 and 1D TRACE50p5 code.The comparison between the results predicted by the codes demonstrate reasonably good agreement, except of some parameters, which would be investigated additionally. The new developed 3D model of TRACE50p5 predict correct results.

The presented article discusses developing and validation of a VVER 1000 TRACE V5.0p5 model. Developing such model for a TRACE computer code is important as it increases the capability of the analyses in simulating nuclear power plant behaviour for different accidents including capabilities for simulating 3D thermal hydraulic phenomena in the reactor core simultaneously with 3D reactor physics calculations. In this way, it is possible to conduct transients involving spatial thermal-hydraulic processes in the reactor core. In developing a VVER 1000 model for TRACE code an existing RELAP5 1D model has been used developed in the INRNE-BAS and widely validated and used for safety assessment of a VVER 1000 reactor system. The validation process includes comparisons of steady-state and transient calculations. The behaviour of selected important parameters calculated from the existing RELAP5 Mod3.3 VVER 1000 model has been compared with results received by using a TRACE V5.0p5 VVER 1000 new computer model. The integral response of the reactor system during a total station blackout (SBO) event has been investigated. The transient includes shutdown of the reactor system, isolating of the turbine, a transition from forced to natural circulation of the coolant in the primary circuit, switching off the feed water system from all steam generators (SGs), activation of safety valves in the secondary and primary circuits, dryout of SGs, loss of natural circulation, reactor core heat up, etc. The performed comparison shows a very good agreement between results received by TRACE model compared to the results received by RELAP5 model.

The presented paper discuses a study of core degradation behavior where a Medium Break Loss of Coolant Accident (MB LOCA) with a 150 mminternal diameter along with station blackout (SBO) was the initiating event. One of the goals of this work was the examination of core degradation at different primary side break locations. Two simulations of MB-LOCA scenarios were performed. The first analysis was performed with the break located in the hot leg and the other with the break in the cold leg. The purpose of the selected study was to investigate plant response with breaks of similar sizes located in different positions in the reactor. This study was focused on in-vessel phase phenomena, such as: core uncovery, core heat up, the beginning of hydrogen generation based on the steam zirconium reaction and the oxidation of other core materials, core degradation, fuel cladding failure, melting of core materials with the formation of a molten pool in the reactor core, relocation of core materials to the bottom of the reactor vessel, and the formation of a molten pool containing corium in the lower head with further vessel failure. The investigation demonstrates that the scenario with a break in the hot leg leads to faster HAs depletion and faster core degradation during first part of the accident with delays in the accidents progression during the later stages of the accident resulting in a later failure of the reactor vessel head due to better core cooling when the results are compared to the evaluation of the accident with a cold leg break. The investigation of the severe accident sequence induced by a MB-LOCA was performed using the computer code ASTEC (Accident Source Term Evaluation Code) v2.1.1.6. The referenced nuclear power plant considered for this analysis is a water-water energetic (WWER)-1000 reactor. The results of the paper could be used for improving severe accident management guidlines (SAMG) as well as for level 2 probabilistic safety analysis (PSA). (C) 2021 Elsevier Ltd. All rights reserved.

CAMIVVER (Codes And Methods Improvements for VVER comprehensive safety assessment) is a three years project launched under the European Union research program HORIZON 2020. The main aim of the Project is to develop and improve computer codes and methods for VVER comprehensive safety assessment. VVER reactors constitute a significant and dynamic part of the European energy market. Their safe Long Term Operation (maintenance, refuelling, safety-upgrade, revamping) relies on the industrial use of neutronics and thermal-hydraulics codes and methods that allow studying the behaviour of the plant in normal and accidental conditions. Computer codes development is strongly required for the following reasons: 1) Current codes and methods used for VVER safety assessment are subjected to growing international export controls from outside EU, threatening the EU sovereignty and security in terms of energy supply 2) A new generation of innovative codes and methods are under development within Europe. They are improving 3Dmultiphysics modelling and uncertainty quantification capabilities and are worth being transferred from lab to industry as they will substantially improve the physics comprehension of PWR and VVER. 3) European codes and methods development for VVER safety assessment will open the VVER market to the European nuclear industry. CAMIVVER will perform developments required for the new generation codes and will generalize 3D-multiphysics coupling, performing benchmark against current industrial codes used for PWR and VVER safety assessment. CAMIVVER will demonstrate CFD assets and compatibility with uncertainty propagation in the frame of a safety assessment.

This paper presents an investigation of VVER 1000 severe accident management guidelines efficiency during "large break loss of coolant accident" simultaneously with station blackout (SBO). The main purpose of this assessment has been focused on the investigation of the efficiency of a second possible entrance at severe accident management guidance (SAMG) strategy (based on Kozloduy nuclear power plant (KNPP)) in case of failure of the first one and an assessment of the possibility to protect the reactor core from significant degradation and the reactor vessel failure. The other goal is the assessment of main plant parameters behavior, like: core uncovery, core heat up, oxidation of core materials, hydrogen generation, core degradation, fuel cladding failure, partial melting of the core materials with the formation of a molten pool in the reactor core, relocation of core materials to the bottom of the reactor vessel, and formation a molten pool containing corium. The scenario included a hot core quenching and recovery of a water level in the reactor core. In the performed work a simulation of operator action, based on a SAMG at VVER 1000 of KNPP is investigated. The selected scenario is a large break loss of coolant accident (LB LOCA) simultaneously with an SBO. The accident starts with a rupture in the cold leg (with inside diameter (ID) 850 mm simultaneously with SBO). All hydro accumulators (HAs) had been available during the accident and prevent earlier damaging of the reactor core. The active safety systems are failing because of loss of all AC and DC power sources. In the selected scenario an operator action based on a SAMG is assumed: quenching of the heated core by an injection of borated water in the reactor vessel when the core exit temperature reaches 980 degrees C by one high pressure injection pump (HPP) and one low pressure injection pump LPP. It is assumed that the first possible entrance into SAMGs at 650 degrees C (923 K) is omitted. The results obtained in this paper could be used for the improvement of SAMG as well as for level 2 probabilistic safety analyses (PSA).

The present work presents an update of Quench 12 experiment calculations using ASTEC V2.2b computer code. The main purpose of the QUENCH program is to investigate fuel behaviour in severe conditions. A special attention has to be paid on a hydrogen generation in result of water or steam injection into overheated and uncovered degraded core of a VVER bundle. The main object of the QUENCH program is to examine the behavior of overheated fuel under different flooding conditions and to create database for model development and improvement of Severe Fuel Damage computer code packages. The QUENCH-12 experiment has been performed to investigate the behavior of VVER fuel assemblies with a hexagonal lattice and fuel rods with claddings made by Zr1%Nb (E 110), which is used in VVER reactors. The ZrO2 pellets as in the LWR-type tests represent the fuel. The test was conducted at Forschungszentrum Karlsruhe on 27 December, 2006 in the frame of EC-supported ISTC (International Science and Technology Center) program. The recently issued (Accident Source Term Evaluation Code) ASTEC2.2b computer code has been use for performing of an update of Quench 12 test calculations. The study was focused on investigation of quenching of overheated VVER fuel behavior, bundle oxidation processes and hydrogen generation. The calculation results have been compared and discussed with a QUENCH test data. The simulation have been performed in the frame of EC (European commission), ASCOM (ASTEC COMmunity) project.

A probabilistic heat transfer model was developed that allows for the gradual evolution of bed-to-surface heat transfer coefficients with increasing superficial gas velocity when bubbling and turbulent flow structures co-exist. The model is based on the packet renewal theory and the probability of particle packets for the specific hydrodynamic regime adjacent to the heat transfer surface. Assuming distinctive two-phase flow, dense packets at minimum fluidization voidage and voids almost free of particles, causes the heat transfer coefficients to be underestimated in the turbulent flow regime. By allowing for contributions of structures of intermediate voidage that reflect typical behavior in the turbulent flow regime of fluidization, heat transfer predictions are improved significantly. The resulting model is shown to give good agreement with experimental heat transfer data for fluidized beds of FCC and Alumina particles, changing smoothly from one flow regime to another in columns of different diameters and from the core of the bed to the wall.

It has been investigated the In-Vessel Melt Retention (IVMR) strategy with external vessel water cooling after the initiation of Large Break LOCA (2 × 850 mm) simultaneously with full station blackout (SBO) at WWER-1000 reactor. The reference nuclear power plant for this analysis is Units 5 and 6 of the Kozloduy NPP. Two calculations have been done with ASTECv2.1.1.3 computer code: In the first calculation ICARE and CESAR modules of ASTECv2.1.1.3 computer code have been used in a coupled mode. The calculation starts at 4910 s. after the initiation of LBLOCA and SBO. At this moment a large amount of molten corium is poured in the lower vessel bottom head. In this calculation just the vessel bottom head was modelled without water and without internals. The external water cooling is initiated also at 4910 s. The second calculation is integral plant simulation of LBLOCA simultaneously with SBO. The modules ’CESAR’, ’ICARE’, ’SOPHAEROS’ and ’CPA’ fromASTECv2.1.1.3 computer codehave beenused in the calculation. LBLOCAandSBOare initiated at 0 s. The external water cooling is initiated at 1025 s. This is the time of the first material and fission products (FP) slump in the lower head vessel. The investigations discussed the maximal values of the internal and the external heat fluxes inside and outside the vessel bottom head for the different segments and other major parameters.

This paper presents an assessment and comparison of the core degradation progression in four Station Blackout (SBO) scenarios analysed by the ASTEC computer code. Two types of SBO scenarios namely with high- and low-pressure conditions are considered. The low-pressure conditions in the SBO scenarios were simulated by introducing a small break loss of coolant accident in the cold leg with an equivalent internal diameter of 80 mm. The selection of the break size was based on the idea to have a significantly faster primary side pressure reduction for simulation of low-pressure conditions and a significantly longer coolant injection from the hydro accumulators (passive safety system). For both types of selected scenarios, the plant behaviour was analysed without operator actions, which allows assessing the times to reach the important set points during the accident progression. The set points include; a) the prediction of the dryout of the steam generators (SG) (or loss of SG effectiveness), b) the loss of natural circulation leading to core uncovery and heat up, c) the beginning of hydrogen generation, d) different stages in core degradation, e) the actuation of the passive safety system etc. The analyses were performed until reactor vessel failure takes place in both of the investigated scenarios. The purpose of these analyses is to study the reactor core behaviour parameters and to estimate the time available to perform operator actions. In addition, this investigation is focused on the assessment of the effectiveness of operator actions as prescribed in the Severe Accident Management Guidlines (SAMGs) for the investigated reactor type. The referenced NPP is KNPP equipped with two VVER 1000 V320 reactors. The ASTECv2.1.1.0 computer code has been used for the investigation. The aim of these analyses is to assess the possibility preserving the reactor core from damage during a severe accident and to assess the hydrogen generation that occurs as a result of the overheated core reflooding at high- and low-pressure initial conditions. The injection of a coolant by an active system will start at the same core exit temperature for each scenario, but at different pressure.

It is investigated the applicability of In-VesselMelt Retention (IVMR) strategywith external vessel water cooling for theVVER-1000 reactor type. The investigation concerns a calculationmadewith ICAREandCESARmodule ofASTECv2.1.1.0 computer code. ASTEC computer code was developed by IRSN (France) and GRS (Germany) to be a European computational tool for simulation of severe accidents in the different reactor designs. The selected reference nuclear power plant for this analysis is Units 5 and 6 of the Kozloduy NPP equipped with VVER-1000 reactor. This type of reactor is a pressurized water reactor with 3000 MW thermal power and 1000 MW electric power. A LBLOCA (double ended guillotine break of the cold leg: 2 × 850 mm) simultaneously with SBO at VVER-1000 reactor design was the postulated SA transient chosen as the most challenging for this type of reactor. The purpose of this calculation is to compute a realistic heat flux profile on the inner side of the vessel and from the vessel to the external water. The periods of maximal heat input from the corium to the vessel steel wall were also accounted.

In case of unsuccessful application of VVER1000 Emergency Operating Procedures (EOPs) for different reasons during the accident in Nuclear Power Plant (NPP), the operators have to leave the EOPs and to start applying Severe Accident Management Guidance (SAMGs). This paper presents calculation results using computer code for validation of operator actions during severe accident conditions in NPP with VVER-1000 type of reactors. The Kozloduy NPP Units 5&6 has been selected as a reference NPP. As an analytical tool has been used ASTECvp2r3p2 computer code developed by IRSN and GRS. The work is oriented on investigation of plant behavior mainly after beginning of reactor core heat up, due to simulation of Large Break Loss of Coolant Accident (LB LOCA) with simultaneous loss of AC and DC power. It is assumed that DGs are missing until a certain plant state, and after that they are available. It have been done calculations without and with operator actions selected based on severe accident management strategies considered in KozloduyNuclear Power Plant (KNPP). Based on the SAMG strategies the operator should depressurize primary circuit by gas removing system (YR) if there is a need of additional primary depressurization and to start to cool down the reactor core at 923 K and 1253 K core exit temperatures as it is pointed in SAMG strategy. The purpose of these analyses is to study the reactor core behavior parameters and to estimate the time available for performing actions. The main goal is to analyze the possibility of preserving the reactor core from damage during a severe accident and to assess hydrogen generation as a result of reflooding of the overheated core. LB LOCA scenario has been selected with break sizes ID 300 mm in the cold leg between main coolant pump (MCP) and reactor pressure vessel. The cold water is injected by a high pressure pump (HPP) in the undamaged cold leg.

This paper presents and discusses the results obtained from ASTEC computer code for investigation of core degradation preservation during specific severe accident transient. The calculations have been performed with ASTECv2.1.1.0 computer code.The purpose of this analysis is to assess the evaluation of ASTEC computer code with modelling of main phenomena arising during hypothetical severe accidents.The performed analyses cover Station Blackout (SBO) scenario with and without injection of passive safety injection systems (hydro-accumulators). The main target of this study is to assess the influence of hydro-accumulators work on core degradation progression in case of simulation of severe accident scenarios (with and without activation of passive safety system). The investigation is focused on investigation of in-vessel phenomena arising during the selected scenario such as dryout of reactor core, hydrogen generation, core material degradation, melting and relocation. It was simulate melt pool formation in the core and on reactor vessel bottom. The analyses have been performed until failure of reactor vessel bottom head in both investigated cases of SBO scenarios.This investigation has been performed in the framework of CESAM project (under Euratom 7-th framework program) by the Institute for Nuclear Research and Nuclear Energy - Bulgarian Academy of Science (INRNE-BAS).