• Energy Research

Abstract Severe accident analysis of a reactor is an important aspect for evaluation of source term. This in turn helps in emergency planning and severe accident management (SAM). Analyses have been carried out for VVER-1000 (V320) reactor following LOCA along with station blackout (SBO) to generate information on these aspects. Availability and unavailability of hydro-accumulators (HAs) are also considered for this study. Integral code ASTEC V1.3 (jointly developed by IRSN, France, and GRS, Germany) is used for analysing the transients. The predictions of different severe accident parameters like vessel rupture time, hydrogen and corium production and radioactivity release to containment have been compared for a spectrum of break sizes to provide information for probabilistic safety analysis (PSA) level-2 and severe accident management (SAM) guidelines.

Abstract In the work is presented the investigation of seismicity of the region where the Kozloduy NPP site is located. This paper intends to present an overview of issues concerning earthquake assessments and relevant activities at the Bulgarian nuclear power plant site – Kozloduy NPP. The overview, based on all the available literature sources, includes: the seismicity of the region where the NPP site is located, the seismic design bases – reassessment and current situation, and seismic hazard assessment for the Kozloduy NPP site as well as the experience of INRNE experts on these topics. Furthermore, this paper describes the basic requirements related to the seismic hazard assessment, specified by the IAEA guidances and the Bulgarian regulations, presently in force. The presented work can be used in further work concerning extending of safety assessment of NPP against the availability of such external events as earthquakes for demonstration of safety operation of NPP.

Abstract The aim of the paper is to present the results and insights gained from the analysis of the postulated severe accident scenario for the Nuclear Power Plant (NPP) equipped with a VVER-1000, focusing on the uncertainty assessment in the prediction of selected Fission Products (FPs). To this end, the simulation of the Large Break Loss of Coolant Accident (LB LOCA) along with a Station Blackout (SBO) scenario was carried out. The capabilities of the ASTECv2.2.0.1 [1] severe accident (SA) code along with the uncertainty quantification tool – SUNSET [2] in the coupled mode was used. The analysis was carried out for the eight input variables and ten output variables. The uncertainty quantification method incorporated within the SUNSET statistical tool was used in order to conduct the present analysis. Based on the Wilks’ formula, the number of ASTEC code runs was determined. The main statistical data are obtained for the selected fission products as figures of merit. The main steps of the analysis are outlined. Furthermore, a brief overview of the SUNSET/ASTECv2.2.0.1 coupling procedures [3] is also presented and discussed. This work demonstrates the application of the uncertainty quantification methodology in the analyses of the severe accidents in the Nuclear Power Plants and provides the specific recommendations of its use. The main outcomes from the assessment provided will be useful and may be a base for the future analyses related to the evaluation of the safety of the VVER-1000.

Abstract This paper presents the thermal-hydraulic investigation of spent fuel behavior during its transferring from reactor vessel through refueling cavity (RC) to spent fuel pool (SFP) in case of dry out for VVER440/V230 units 3 and 4 at Kozloduy NPP. The fuel transfer canal connects the refueling cavity and spent fuel pool and this way set up a command pool. The presented analysis has been performed up to the moment of fuel heat up in case of spent fuel pool (SFP) dry out during the first stage of refueling activities. The main feature during this stage is: the maximum decay power of “new” spent fuel; the “new” spent fuel is still in the reactor vessel; the fuel transfer canal connects the refueling cavity and spent fuel pool. In this way the coolant have maximum volume and the spent fuel with maximum decay power is still in the reactor vessel. The “old” spent fuel in SFP has significantly low decay power. The main purpose of this analysis is to estimate the time for dry out of SFP, heat up of spent fuel and time for recovery actions from the operators. In the performed analysis are defined the following stages during the accident. ∘ Termination of natural circulation after decreasing of water level in reactor vessel below the hot nozzles. ∘ Beginning of coolant heat up in the reactor core. ∘ Reaching the temperature of saturation at the outlet of the assembly. ∘ Startup of the reactor core uncover. ∘ Loss of critical safety functions. The analysis has been performed with the thermal-hydraulic computer code RELAP5/MOD3.2. The RELAP5/MOD3.2 model for Kozloduy NPP VVER-440 have been developed and validated at the Institute for Nuclear Research and Nuclear Energy – Bulgarian Academy of Sciences (INRNE-BAS) Sofia.

Abstract 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 923.15 K. • 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.

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

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 hy- draulics 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 VVER- 1000 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.

Abstract This paper describes validation of a computer model that has been developed for VVER 440 Nuclear Power Plant (NPP) for use with RELAP5/MOD 3.2 computer code in the analysis of the following transient: “Control rod assembly drops to fully inserted position”. This validation is a process that compares the analytical results obtained by RELAP5/MOD3.2 model of the VVER 440 with experimental transient data received from Kozloduy NPP, Unit #2. The model of VVER 440 was developed at the the Institute for Nuclear Research and Nuclear Energy for analyses of operational occurrences, abnormal events and design basis scenarios. It will provide a significant analytical capability for the Bulgarian technical specialists located at the Kozloduy NPP. The criteria used in selecting transient are: importance to safety, availability and suitability of data followed by suitability for RELAP5 code validation. The comparisons between the RELAP5 results and the test data indicate good agreement.

Abstract 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.

The Management and Uncertainties of Severe Accidents (MUSA) project, funded in HORIZON 2020 and coordinated by CIEMAT (Spain), aims at consolidating a harmonized approach for the analysis of uncertainties and sensitivities associated with Severe Accidents (SAs) focusing on Source Term (ST). In this framework, the objectives of the Innovative Management of Spent Fuel Pool Accidents (IMSFP ??? WP6), led by IRSN (France), are to quantify and rank the uncertainties affecting accident analyses in a Spent Fuel Pool (SFP), to review existing and contemplated SA management measures and systems and to assess their possible benefits in terms of reduction of radiological consequences. To quantify the propagation of the uncertainties of the input parameters to the output uncertainties of severe accident codes (ASTEC, MELCOR, RELAP/SCDAP), a diverse set of uncertainty quantification (UQ) tools (DAKOTA, RAVEN, SUNSET, SUSA) are used. The statistical framework used by the different UQ-tools is similar e.g. pure random (Monte Carlo) and Latin hypercube sampling (LHS). Fourteen partners from three different world regions are involved in the WP6 activities. The target of this paper is to describe the achievements during the first three years of the project. In a first part, a description is given of the SFP accidental scenario, of the key target variables and radionuclides chosen as ST Figures of Merit (FoM) and of the identified uncertainty sources in models and input parameters. A key element when defining the SFP scenario has been the consideration (or not) of the reactor building, as it is expected to significantly affect analyses. In a second part, the first insights coming out from the calculation phase of the project are presented. The review of existing SA management measures is also exposed, as well as systems whose benefits will be assessed in the second phase of the project. Finally, challenges that arise from such an exercise are discussed, as well as major difficulties found when applying UQ methodologies to SFP scenarios and solutions adopted