Kozloduy Nuclear Power Plant Research Articles

Status and future prospects of nuclear industry development in Bulgaria

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 indicatesthat 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.

Cooldown of a water-cooled reactor during the natural circulation mode using decay heat of the core and a low-power steam turbine

The Fukushima-1 nuclear power plant accident in 2011 confirms that NPP safety is one of the most urgent present-day problems. Moreover, a rapidly growing number of NPP units increases the risk of serious radiation accidents. The authors have developed a novel cooldown method for water-cooled reactors to be used under loss of connection with the power system using decay heat from the core and a low-power steam turbine. A distinctive feature of the method is financial viability of the steam turbine due to its possibility to generate additional electricity at the normal conditions of NPP operation. The aim of the invention is to ensure a high level of NPP safety while maintaining and increasing commercial efficiency and competitiveness of nuclear power plants. To implement the developed method, it is necessary to provide guarantee for reliability of NPP units at natural circulation of the primary coolant, since the primary circulation pumps will be shut off caused by a trip of a reactor. For this reason, the authors carried out an investigation into international practice theory and actual usage of natural circulation of water-cooled reactors. Experiments conducted at Kozloduy NPP and research of a number of scientists demonstrate a possibility to maintain constant decay heat removal that will ensure obtaining steam with stable parameters in the secondary circuit. The second part of the research deals with a new method aimed at beneficial usage of steam obtained by using the decay heat from the core under natural circulation of the primary coolant. The generated steam can be used in the additional low-power steam turbine to power supply the NPP's auxiliary needs. In our earlier research we performed a preliminary assessment of reliability of the developed system. The results of the research showed that a combination of a steam turbine and a standard three-channel emergency power supply system with diesel generators enables a safety level comparable to that of external heat exchangers in the passive heat removal systems. At the same time, passive heat removal systems require significant capital investments and maintenance costs, whereas according to the method developed by the authors, a low-power steam turbine can operate in the normal regime to generate additional electricity, as well as in combination with energy storage systems. This will ensure full cost recovery of the developed installations. To assess a reduction of risks of nuclear accidents in terms of cash equivalents when using the novel method, the authors analyzed the influence of various factors on damage expectancy in case of an accident at NPP, based on the example of Fukushima and Chernobyl NPPs. The investigation showed that a reduction of nuclear accident risk in terms of cash equivalents, when implementing the authors’ method aimed to supply the auxiliary NPP needs in blackout situations, reaches tens of millions of US dollars per year.

Efficiency Assessment of Sanitary-Technical Barriers in Nuclear Facilities by Means of Adhesive Decontaminating Foils

The efficiency of sanitary-technical barriers in Kozloduy NPP, Bulgaria, is investigated for the first time with adhesive decontaminating foils. The duration of use and retention coefficient are assessed using gamma-ray spectroscopy. The retained activity increases linearly up to a saturation point at 250 controlled persons. About 6% of the radioactive mechanical fraction is fixed on the adhesive. For three successive sanitary-technical barriers in the controlled zone, the activity is reduced by one to two orders of magnitude on each barrier towards the sanitary complex exit. The total activity released during annual maintenance is estimated to be 105 Bq. The radioactive contamination in the controlled zone is studied and correlations are established between two groups of isotopes, generated by the release of fission products and by neutron activation.Our investigation demonstrates that the sanitary technical barriers prevent efficiently the release of radioactivity from the controlled zone to the environment.

Geoelectric studies of the Kozloduy nuclear power plant region, Bulgaria

The task of the work was geoelectrical studies using variations of the magnetotelluric (MT) field of the Kozloduy nuclear power plant (KNPP) region and the integration of its results with other geological and geophysical knowledge. This paper presents the determined interpretation parameters of the MT field. The KNPP is located on the right bank of the Danube River in close proximity to the river. This fact, together with the location of electrified railways determined the unique network of locations of observation points for MT field variations. Based on the analysis of Earthquake Catalogs of Bulgaria and international seismicity databases, a map of the seismicity of nuclear power plant areas was built. Over the past 50 years, about 750 earthquakes (mainly south of KNPP) have been recorded at a distance of 40—80 km from the KNPP. Two magnetotelluric stations GEOMAG-02 were used at measurement sites, but equipment for recording electrical channels was available only for one station (due to the lack of another set of non-polarizable electrodes). The MT field variations were observed at 21 points, which are located on the territory with sides approximately 30—35 km from east to west and 40—50 km from north to south. For all observation points on the profile, only the parameters of the vertical magnetic transfer function (VMPF) were determined, in the form of the real (Cu) and imaginary (Cv) parts of the induction vector. The steadily induction vector was defined for periods from 10—20 to 4900—10 800 s. For most points it was possible to estimate the values Cu, Cv with an error of 0.02—0.04 and AzCu, AzCv 3—5°. The analysis showed the presence of anomalous behavior of Cu, Cv in different intervals of periods at some points. In the shortest (about 20 s) and longest periods (600 to 1000 s), the Cu directions completely coincide and indicate the presence of anomalous conductivity of the quasi-longitudinal strike to the west of the study area. This behavior of the Cu vector is in good agreement with power isohypsum strike of the Cenozoic deposits. At intermediate periods of 50—200 s, the behavior of Cu is more complex. Approaching the zone of high seismicity, the direction of the Cu differs from the previous ones by almost 90°. On the Geoelectrical sections, obtained as a result of 1D inversions of MTS curves at 4 points located in the southern part of the region, anomalous layers are identified (ρ about 10 ohm · m, the depth of the center of the object is 15—20 km). It can be assumed that well-conducting objects in the Earth’s crust of the region, apparently, prevent the propagation of seismic waves from nearby earthquakes to the north towards the KNPP.

EVALUATION OF INTERNAL EXPOSURE TO RADIOACTIVE AEROSOL GENERATED FROM PLASMA MELTING SYSTEM USING THE BIDAS CODE

The radioactive aerosol generated by the Nuclear Power Plant (NPP) decommissioning process can be inhaled by workers and deposited inside the human body, resulting in internal exposure. Because internal exposure, unlike external exposure, is difficult to measure directly, it is all the more necessary to assess the dose workers receive as a result of internal exposure. Precise assessment of the internal exposure necessitates actual measurements in the work environment such as the workers' respiration rate, kind of nuclide and amount of captured nuclide. However, in the event of difficulties in securing these measurements, the internal exposure dose can be estimated based upon the recommended values by the ICRP (International Commission on Radiological Protection) such as the intake fraction and particle size. In this study, 5μm was selected as the particle size as recommended by the ICRP, and both heavy and light respiratory rates were used in the calculation. With respect to the nuclides contained in the radioactive aerosol and their concentrations, the data captured for the aerosol in the melting facility on the Kozloduy NPP premises in Bulgaria were applied to estimate workers' internal exposure. As a result, each worker was found not to have received more than 20mSv/yr, which is the maximum annual permissible dose for workers.

Calculation of decay heat rate in VVER-1000 spent fuel pool including uncertainties

This paper presents a decay heat rate analysis of the VVER-1000 (unit 6 in “Kozloduy NPP” Plc.) spent fuel pool (SFP) using the method outlined by the United States (US) Nuclear Regulatory Commission (NRC) Regulatory Guide (RG) 3.54 (Revision 2, 2018), titled “Spent Fuel Heat Generation in an Independent Spent Fuel Storage Installation.” The usage of RG 3.54 for SFP calculations is extremely attractive because it enables fast determination of the decay status of a large number of stored fuel assemblies (FAs) in an SFP and further evaluation of their uncertainties.Two batches of fuel FAs in an SFP were analyzed in the study: The first batch consisted of 402 FAs that were stored for 321 days of reactor power operation from 2014 to 2015. The second batch consisted of 355 FAs that were stored for 326 days of reactor power operation from 2018 to 2019. The total decay heat rate and the corresponding decay heat rates of the FA that was placed in each of three VVER-1000 SFP sections were evaluated. Additionally, the propagation of the input uncertainties introduced by the initial fuel mass, burnup, and operation time on full power effective days was estimated.The applied method was derived from the uncertainty method, which was previously introduced by Gesellschaft für Anlagen-und Reaktorsicherheit (GRS), for free from deliberate pessimism safety analyses (named best estimate). The variation in the selected input parameters produced decay heat generation uncertainty as follows: ±73.2 kW (2014)/±59.1 kW (2018) at the beginning of each batch storage, and correspondingly, ±12.0 kW (2015)/±10.7 kW (2019) at the end of each batch storage. The results indicate that the method is sufficiently sensitive to the initial mass variation and irradiation uncertainties.The degree of conservatism (via a parameter referred to as a ‘safety factor’), introduced in RG 3.54 to address the numerical and nuclear data uncertainties in pressurized water reactor (PWR) decay heat calculations, was calculated for the stored VVER-1000 FA. An original approach for calculation of the scaling coefficient kcorr is suggested in the paper for estimation of the level of conservatism using the safety factor in RG 3.54 and other sources. The obtained results are reported.The recommendations by RG 3.54 were implemented in an input script for the MAXIMA Computer Algebra System (CAS). The MAXIMA CAS calculations of the VVER-1000 SFP decay heat rate were verified with the corresponding Oak Ridge Isotope Generation (ORIGEN) (SCALE 6.1) data at the beginning and end of the analyzed time periods. The authors determined that the obtained values of kcorr = 0.91–0.93 are useful in the re-evaluation of the RG 3.54 degree of conservatism. Furthermore, the scaled safety factor decay heat generation output was replaced with the calculated upper uncertainty bound. The upper boundary values in the decay heat rate uncertainty range showed flexibility and reduced conservatism in comparison with the upper 10% of the conservative bound of the nominal values.

Corporate Social Responsibility and Socially Responsible Investing Strategies in Transitioning and Emerging Economies. Choosing A Strategic Investor in The Field of Energy.

Abstract: Corporate Social Responsibility (CSR) and socially responsible investment (SRI) are concepts that have entered our everyday life over the last decades. These indicators of the state of the economy and its related components were introduced by developed economies 40 years ago, and are being imposed in the transition and emerging market economies in the 21st century. Initially, large national and international companies introduced this, and in the next phase, medium and small companies also implemented them. At this stage of development, it is unacceptable for companies not to apply CSR and SRI, but there are still quite a number of such companies in Bulgaria and Eastern European countries. The energy sector in the world is one of the leaders in the implementation of CSR and SRI. In Bulgaria and other restructured market economies, energy companies also occupy the top positions in various CSR and SRI implementation charts, such as Kozloduy NPP, Mines Maritza East, Assarel, EVN and others. In particular, some positive examples of the implementation of corporate social responsibility and socially responsible investments in the energy sector in Bulgaria in recent years are given.

Plasma Technology to Recondition Radioactive Waste: Tests with Simulated Bitumen and Concrete in a Plasma Test Facility

The operation and maintenance of nuclear power plants, the non-nuclear fuel cycle, etc generate low-level radioactive waste which, along with the historical radioactive waste from past nuclear activities, needs to be treated and stored, awaiting final disposal. Plasma technology offers a very effective way of treating this waste with a high-volume reduction factor (VRF), free from organics, liquids and moisture, and meets without a doubt the acceptance criteria for safe storage and disposal. By means of a plasma beam of approximately 5000°C, the inorganic materials are melted into a glassy slag, containing most of the radioactive isotopes while the organic material is gasified, oxidized and purified in an off-gas cleaning system. First the paper describes the new full-scale Plasma Melting Facility (PMF) at the Kozloduy Nuclear Power Plant in Bulgaria which was taken into nuclear operation in May 2018. The plant has a capacity of 250 tons per year and the maximum contact dose rates of the incoming waste is 2mSv/h. Different mixtures of radioactive waste packed in 200l drums were successfully treated resulting in a glassy slag free from liquids and organic material with an important volume reduction factor (VRF). The Project was co-financed through a grant by Kozloduy International Decommissiong fund (KIDSF) administated by the EBRD through Bulgarian national funding. Plasma is a suitable technology for treatment of problematic waste or even reconditioning waste so Belgoprocess was contracted to do plasma tests with simulated conditioned waste types. One can do tests on a laboratory scale on smaller samples and torch capacities of e.g. 50kW but Belgoprocess wanted to do more realistic and reliable tests. So Belgoprocess contracted Phoenix Solutions Co who has a full-scope test facility equipped with a 1200kW plasma torch for full-scope treatment of simulated conditioned waste. For a first confidential contract simulated 2001 (55 gallon) bitumen drums were treated. The drums contained different pucks of compacted waste such as rags, used filters, granulates, etc. The pucks were stacked in the 2001 drums and subsequently embedded with bitumen. A total of 6 drums were treated in the plasma facility. For a second contract simulated homogeneous 2001 (55 gallon) concrete drums with on the one hand concentrates and on the other hand spent resins were selected. A total of 6 drums with concrete and spent resins were treated and melted in the plasma testing facility. The paper describes the test facility, volume reduction factor (VRF) of different waste streams and most important parameters.

Simulating the Operation of the Jet-Vortex Condenser Used in the VVER-440 Reactor Plant’s Confinement System

A jet-vortex condenser (JVC) used as part of the confinement system (CS) serves for limiting the pressure in sealed rooms to the maximum admissible value for which the civil structures of NPPs equipped with VVER-440 reactors made according to the V-230 and V-179 projects are designed under the loss of coolant accidents' conditions. Fitting the Kola NPP’s first stage equipped with the VVER-440 reactor (made according to the V-230 project) in combination with taking certain other measures opened the possibility to extend the NPP service life. To substantiate the possibility of extending the service life of NPP power units, their safety will be assessed using an analysis model. The article presents the results from such modeling aimed at analyzing the performance of the jet-vortex condenser that is used as part of the confinement system for limiting the pressure in the sealed rooms down to the maximum admissible value for which the VVER-440 reactor-based NPP civil structures are designed with due regard to loss of coolant accidents (LOCA). The jet-vortex condenser operation model incorporated into the KUPOL-M computer code was verified against the experimental data obtained on the test facility installed at the All-Russian Research Institute for Nuclear Power Plants Operation (VNIIAES). A comparative analysis has shown that the calculated and experimental data are in fairly good agreement with each other. The KUPOL-M code with the jet-vortex condenser model integrated into it was used for calculating the parameters of medium in the system of sealed rooms in modeling the large break LOCA and small break LOCA conditions for the Kola NPP’s first stage, the distinctive feature of which is that it contains a JVC in the confinement area. The calculation results have confirmed the JVC efficiency under the conditions of a wide range of loss of coolant accidents. At present, the JVC has been put into use in the Kola NPP units 1 and 2, in the Novovoronezh NPP units 3 and 4, and in the Kozloduy NPP units 3 and 4 in Bulgaria.

The Bulgarian nuclear sector in transition: Adopted Russian technology, state sovereignty and accession to the EU

Applying the notion of ‘orders of worth’ (L. Boltanski, L. Thévenot), the paper explores the technology & policy challenges the Bulgarian nuclear energy sector faced from the mid-1980s to the early 2000s. Emerging after the 1960s based on Soviet technology and under Soviet political domination, it had to adapt to the new conditions set by the winners of the Cold War and to the pressure of local green movements. Key events included the completion of the second 1000 MW Soviet nuclear reactor at the Kozloduy Nuclear Power Plant (NPP), the termination of work on the Belene NPP (1990–1991), and the painful decommissioning under EU pressure of the four old 440 MW reactors (2004–2006). The paper traces the public debates during these developments, the resources used by the actors to justify their behavior and draws conclusions about the peculiarities of emerging ‘technical democracies’ in Eastern Europe.

Discussion on Practical Elimination of Early or Large Releases for WWER-1000/V320

The paper presents a brief summary of the introduction of the term “practical elimination” as prevention of the conditions that could lead to early or large radioactive releases. The concept of “practical elimination” is defined as part of the Defence in Depth (DiD) of Nuclear Power Plant (NPP) in the International Atomic Energy Agency (IAEA) document INSAG-12 in 1999. But, the special attention to it was paid after the accident in Fukushima NPP in 2011. The mechanisms of the containment failure of reactor WWER-1000/V320 are presented. As an example, the summarized design features and preventing and mitigation measures already implemented at Kozloduy NPP to extend the design basis and beyond design basis envelop are presented. Issues related to external steam explosion are underlined for further study.

Ageing management in NPP. Effectiveness of the methods for control, examination and monitoring in relation to mechanisms of degradation of mechanical properties

This paper describes ageing management activities for Long Term Operation (LTO) of components of Kozloduy NPP EAD. The stages of LTO Project are: Stage 1 - implementing a comprehensive assessment of the actual condition of the equipment; Stage 2 - providing of a Complex Analysis; Stage 3 – licensing of Plant Life Extension (PLEX) for long-term operation (LTO). The main activities are assessment of actual condition of the equipment and building; review of methods for control (non-destructive), examination and monitoring; assessment of effectiveness of methods; necessity of additional control/ examination/ service. The paper describes some mechanisms of degradation of mechanical properties, methods for control and criterions for their effectiveness.

A study of the pressure vessel steel of the WWER-440 unit 1 of the Kozloduy nuclear power plant

A comparison between highly neutron irradiated samples from the region of weld № 4 and low irradiated samples from weld № 1 taken from the pressure vessel of the WWER-440 Unit № 1 of the Kozloduy NPP has been performed. Measurements of the residual activity of samples from the outer surface of the reactor pressure vessel bottom corpus reveal very low activity of 60Co. Insofar as there the base and weld metal appear to be exposed to a very low neutron fluence, the samples from these locations can be considered as practically not affected and may serve as a reference basis for comparison with highly irradiated pressure vessel regions. The Mossbauer parameters isomer shift (IS) and quadrupole splitting (QS) were found to be absolutely irradiation insensitive. A stepwise reduction of the internal hyperfine magnetic field Bhf, each by about 2.6 T, was observed. This can be attributed to the replacement of one or two surrounding iron atoms as first nearest neighbors by non-iron alloying atoms. The Mossbauer experimental line widths for irradiated and non-irradiated samples are practically the same, which is a quite unexpected result. The area fraction ratio for the three main Zeeman sextet subspectra S1:S2:S3 shows very high irradiation sensitivity. For the bottom low irradiated region of the reactor vessel the values are S1:S2:S3 = 50.1:40.0:9.4. After seven years of operation between the pressure vessel annealing in 1989 and the autumn of 1996 when the samples from weld № 4 were taken the ratio changes strongly to S1:S2:S3 = 56.4:34.7:8.5. A possible explanation of this result is that neutron irradiation gives rise to a precipitation process involving predominantly alloying atoms as Ni, Mn, Cr, Mo and V which become mobile and precipitate in the form of carbides and/or P-rich phases and alloying atom aggregates. This “refinement” process lowers the partial area of subspectra S2 and S3 where alloying atoms are involved and leads to a higher area fraction of the pure iron component S1, which is the major experimental result. For a more complete Mossbauer investigation on the processes of generation of structure defects caused by the neutron fluence, a new series of measurements will be performed by using a set of so-called surveillance specimens with different irradiation histories which are available only for the WWER-1000 reactors of the Kozloduy NPP.

Assessment of strontium transport in clay medium by using a reactive transport modelling approach – an HP1 case study model

Modelling of radionuclide transport in the geosphere is a key issue of the evaluation of safety of low- and intermediate-level radioactive waste (LILW) repositories. In Bulgaria, due to the operation of Kozloduy Nuclear Power Plant (NPP), a building of such facility is ongoing. The migration of 90Sr was evaluated by simulation of flow and chemical modelling, performed with coupled numerical code HP1. The adsorption properties of the clays were implemented in the code by surface complexation reactions between strontium and iron oxides. The results were compared with another model performed with computer code HYDRUS 1-D, in which the adsorption properties of the medium were evaluated with the relevant for strontium distribution coefficient, Kd.

Determination of plutonium isotopes in environmental samples with triisooctylamine and alpha spectrometry

An alternative method for determination of Pu isotopes in various environmental samples has been developed. The separation of Pu is attained by two stage solvent extraction with triisooctylamine/xylene in hydrochloric acid and nitric acid [Pu is converted to Pu4+ by the redox system Fe3+/(NO3)−/(SO3)2−] media. The analytical method has been successfully applied for the determination of Pu isotopes in surface air in a 100 km area around the Kozloduy Nuclear Power Plant in Bulgaria and in surface water of the Danube River. The analytical quality was checked by analyzing reference materials from IAEA (Soil-6, IAEA-375, IAEA-384) and NPL (AL-2008, AL-2009, AL-2010, AL-2011).

Study of in-vessel melt retention for VVER-1000/v320 reactor

This study aims to assess the applicability of the in-vessel melt retention (IVMR) strategy with external vessel water cooling to the reactors of VVER-1000/v320 type. IVMR strategy is one of the feasible solutions to mitigate reactor vessel failure and further fission products release to the containment and to the environment outside.The reference power plant for this investigation is VVER-1000/v320 reactor sited at Unit 5 and 6 of Kozloduy NPP. The ASTECv2.0r3 severe accident computer code was used to simulate Large Break LOCA (2×850mm) with full Station blackout (SBO) in VVER-1000/v320 reactor model. The external water cooling of the vessel bottom head during the accident was also simulated by the code.The results from four calculations have been assessed: two integral calculations made with ASTECv2.0r3with and without simulation of external vessel water cooling and two stand-alone calculations made with the ICARE module of ASTECv2.0r3. The heat fluxes from the corium to the vessel and the heat fluxes from the vessel to the outside water have been predicted by the code. There were also accounted periods of maximum heat input from the corium to the vessel steel wall.

Status of Severe Accident Management Guidelines at Kozloduy Nuclear Power Plant

The objective of this paper is to present the current organization of the Emergency Procedures including Emergency Operating Procedures (EOP) and Severe Accident Management Guidelines (SAMG) in Kozloduy Nuclear Power Plant (KNPP) as a function of the severity of the accident conditions. Special attention is paid to SAMG. It is described when the SAMG are used and at which conditions in a transition between the EOPs and the SAMG should be made. The Critical Safety Function Restoration Guidelines and their connections with SAMGs and EOPs are also discussed. The arrangement of SAMG is described in detail, since in the KNPP exist 2 types of SAMGs for Main Control Room (MCR) and for the Accident Management Centre (AMC) and they contain the same strategies, but they are different in format. Both types are symptom oriented procedures, but those for MCR are in 2-column-format with interconnections, whereas those for the AMC are developed in a logical manner and simplified for people, who take decisions. In the paper, they are also discussed the adopted strategies in existing SAMG that should be followed to recover from a damaged core condition and to prevent or mitigate the release of fission products. In the paper, they are also described a number of technical measures for management and mitigation of severe accidents, which are implemented in KNPP before and after the Fukushima accident. Many of them are common for WWER-1000 type of reactors, but some of them are unique and plant specific. This information can be useful for operators of other WWER type reactors or even PWR reactors.

Study of Accident Progression in Unsealed WWER-1000/V320 Reactor during Maintenance

This paper discusses the results obtained during an investigation of WWER-1000 Nuclear Power Plant (NPP) behavior at shutdown reactor during maintenance. For the purpose of the analysis is selected a plant operating state with unsealed primary circuit by removing the MCP head. The reference nuclear power plant is Unit 6 at Kozloduy NPP (KNPP) site. 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 KNPP 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 plant modifications performed in frame of modernization program have been taken into account for the investigated conditions for the unsealed primary circuit. The most specific in this analysis compared to the analyses of NPP accidents at full power is the unavailability of some important safety systems. For the purpose of the present investigation two scenarios have been studied, involving a different number of safety systems with and without operator actions. The selected initiating event and scenarios are used in support of analytical validation of Emergency Operating Procedures (EOP) at low power and they are based on the suggestions of leading KNPP experts and are important in support of analytical validation of EOP at low power.

Analytical validation of operator actions in case of primary to secondary leakage for VVER-1000/V320

This paper presents the results of analytical validation of operator actions in case of “Steam Generator Tube Rupture” (SGTR) 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 purpose of the analyses is to demonstrate the ability to terminate primary to secondary leakage and to indicate an effective strategy for preventing secondary leakage to the environment and in this way to prevent radiological release to the environment. Following depressurization and cooldown of reactor coolant system (RCS) with isolation of the affected steam generator (SG), in these analyses are validated options for post-SGTR cooldown by:•back up filling the ruptured SG;•using letdown system in the affected SG and•by opening Fast Acting Isolation Valve (FAIV) and using Steam Dump Facility to the Condenser (BRU-K).The results of the thermal-hydraulic analyses have been used to assist KNPP specialists in analytical validation of EOPs.The RELAP5/MOD3.2 computer code has been used for 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 thermal-hydraulics code RELAP5/MOD3.2 at the Institute for Nuclear Research and Nuclear Energy – Bulgarian Academy of Sciences (INRNE-BAS).This paper is possible through the participation of leading specialists from KNPP.

Investigation of station blackout scenario in VVER440/v230 with RELAP5 computer code

During the development of symptom-based emergency operating procedures (SB-EOPs) for VVER440/v230 units at Kozloduy Nuclear Power Plant (NPP) a number of analyses have been performed using the RELAP5/MOD3 (Carlson et al., 1990). Some of them investigate the response of VVER440/v230 during the station blackout (SBO).The main purpose of the analyses presented in this paper is to identify the behavior of important VVER440 parameters in case of total station blackout.The RELAP5/MOD3 has been used to simulate the SBO in VVER440 NPP model (Fletcher and Schultz, 1995). This model was developed at the Institute for Nuclear Research and Nuclear Energy, Bulgarian Academy of Sciences (INRNE-BAS), Sofia, for analyses of operational occurrences, abnormal events and design based scenarios. The model provides a significant analytical capability for specialists working in the field of NPP safety.