Beyond Design Basis Accident Research Articles

Analysis of the hydrogen safety of the NPP with a VVER-1200/V-491 reactor in a severe beyond design based accident

The results of a computational analysis of possible modes of combustion of a hydrogen-air mixture in containment during the evolution of a severe beyond design basis accident (BDBA) with the COCOSYS code are presented. The containment of the NPP with a VVER-1200/V-491 reactor was selected as the object of study. BDBA with loss of coolant occurs due to a break of the injection pipeline of the pressurizer system (LOCA DN179) and the simultaneous failure of all active channels of the emergency core cooling system (ECCS). The calculated parameters of the thermodynamic state (pressure and temperature) of the gas mixture are given, and the values of the concentration distribution of hydrogen inside the containment are presented. After calculating the hydrogen distribution and mixing in all compartments of the containment, an evaluation was made of the flammability of the mixture and the potential for self-ignition in the containment by using a three-component Shapiro-Moffette diagram. It was concluded that in BDBA (LOCA DN179 with failure of active ECCS) detonation of the hydrogen-containing mixture is excluded, and deflagration is possible only in the emergency compartment of steam generators with a pipeline break. Thus, the hydrogen risk mitigation has been achieved in accordance with the standards established by the Belarusian regulator, provided that the localization safety systems are operational in the event of hydrogen deflagration. And the efficiency of the hydrogen removal system from the containment using catalytic recombination is considered sufficient for the considered BDBA.

Large pipe break opening time in pressurized heavy water reactors

In the frame of the licensing process of the AtuchaII PHWR (pressurized heavy water reactor) the BEPU (Best-Estimate Plus Uncertainty) approach has been selected for issuing Chapter 15 of FSAR (Final Safety Analysis Report) dealing with accident analysis. In case of a LOCA (Loss of Coolant Accident) the assumption of instantaneous opening time is too conservative, especially for this kind of reactor with a positive coolant void reactivity coefficient.Analysis was performed to assess the ability of Atucha II nuclear power plant’s fast boron-injection system to prevent overheating of the fuel elements in the event of a Beyond Design Basis Accident (BDBA). The goal of the Analysis was to develop a Dynamic Break Opening which was used as the input for the LOCA analyses. The results show that the fast boron injection can effectively shut down the reactor, thereby preventing the overheating of the fuel elements.

Transient Analysis of LBLOCA with Subsequent Loss of Offsite Power and Loss of All AC Power in VVER-1200 Reactor

The generation III+ VVER-1200 (AES-2006) reactor is equipped with improved safety features with component redundancy compared to the previous models. This paper is concerned with the performance of active and passive safety features during the transient of a major beyond design basis accident (BDBA) scenario of the reactor. The study is done considering large break loss of coolant accident (LB LOCA) with loss of offsite power (LOOP) and LB LOCA with loss of all AC power (LOACP) sources namely station blackout (SBO) using VVER-1200 NPP simulator. It was found that the reactor is capable of mitigating such BDBA scenarios by the effectiveness of the active and passive safety systems. Moreover, LB LOCA with LOACP is observed to be more severe compared to LBLOCA with LOOP. This comparative analysis is compatible with IAEA safety guidelines.

Recent Progress on Creep Properties of ODS FeCrAl Alloys for Advanced Reactors.

In order to meet the growing energy demand, more environmentally friendly and efficient GEN-IV reactors have emerged. Additionally, nuclear structural materials need larger more safety margins for accident scenarios as a result of the Fukushima accident. In order to extend the failure time and lessen the effect of accidents, this design method for accident-tolerant fuel materials calls for cladding materials to maintain good corrosion resistance and mechanical properties during a beyond design basis accident (BDBA). Accidents involving nuclear reactors would undoubtedly result in higher temperatures, which would make it much harder for materials to withstand corrosion. Oxide dispersion strengthened (ODS) FeCrAl alloys have shown promise as candidate materials because of their extraordinarily slow reaction rates under high-temperature steam. However, the addition of the Al element renders the alloy's high-temperature mechanical properties insufficient. In particular, studies on the alloy's creep properties are extremely rare, despite the fact that the creep properties are crucial in the real service environment. Therefore, this paper focuses on the creep properties of ODS FeCrAl alloy, summarizes and analyzes the research results of this material, and provides a reference for future research and applications.

Numerical study of the effect of geometry on the behaviour of internally heated melt pools for in-vessel melt retention

In-vessel melt retention is a vital safety design that is incorporated in some nuclear reactors to protect against beyond-design basis accidents. This entails the retention of the molten fuel, colloquially known as corium, to be retained within the reactor pressure vessel (RPV) for extended periods of time without any release of radioactivity to the surroundings.Multiple experiments have been done to better understand this phenomenon, such as the COPRA experiment, which is simulated in this study. The study aims to analyse the COPRA experiment’s forced convection cooling (via water) while also simulating cooling via natural convection. After comparing this to the original COPRA experiment’s data, the study showcases the effect of geometry variation in such experiments by testing different RPV geometry sizes.Our results can better inform the design of future severe accident experiments while taking into account the historical aspects of severe accident research, as the popularity of Small Modular Reactors (SMRs) increases. Our analysis of the variation in geometry provides insights for RPVs that are significantly smaller than the traditional pressurised water reactors’ RPV size. The difference in cooling effect and crust formation is presented for this variation of vessel size. The difference is not significant but is noteworthy for the design of future experiments and reactors.

Стан наукової проблеми щодо дослідження характеристик продуктів взаємодії розплаву ядерного палива з бетоном

Стан наукової проблеми щодо дослідження характеристик продуктів взаємодії розплаву ядерного палива з бетоном

Analysis of BDBA sequences in a generic IRIS reactor using ASTEC code

Integral Severe Accident (SA) codes are aimed at providing an exhaustive coverage of all the main phenomena taking place in a core melt accident. Today, these deterministic codes have reached a high level of maturity for the simulation of operating reactors and the nuclear technical community is starting to extend their applicability to advanced reactor designs, as Small Modular Reactor (SMR). In the framework of the NUGENIA TA-2 ASCOM (ASTEC COMmunity) collaborative project, a generic input-deck based on the IRIS design has been developed for the ASTEC code. The generic SMR ASTEC model has been already proved able to simulate the main thermal–hydraulic phenomena driving the passive mitigation of a SBLOCA in Design Basis Accident (DBA) conditions. The same initiator event, regarding the guillotine break of a Direct Vessel Injection (DVI) line, will be assumed for the simulation of beyond-design scenarios by considering the unavailability of selected passive safety systems. The results of the ASTEC simulations of four Beyond Design Basis Accidents (BDBAs) (study carried out with ASTEC V2.2, IRSN all rights reserved, [2021]) will be analyzed and discussed against the reference DBA sequence in the present paper. This study is aimed at proving the first insights about the capability of the ASTEC model of a generic IRIS reactor to be used in BDBA and in SA analyses, if significant core degradation takes place. In addition, it characterizes the role played by each safety system in SMR passive mitigation strategy and give the possibility to characterize the phenomenologies specific of SMR designs.

Analisis Skenario Rencana Kontinjensi Nuklir Tangerang Selatan Menggunakan JRODOS

The scenario of the radionuclide release in the South Tangerang city nuclear contingency plan (2013) uses the GA Siwabessy worst accident scenario, namely the Beyond Design Basis Accident (BDBA). The technique uses a predetermined wind direction and wind speed. However, problems in the field are wind direction constantly changing and may not predict the time of an accident in advance. The objective of this paper is to conduct an initial analysis of the scenario of the radionuclide release in the South Tangerang city nuclear contingency plan using JRODOS decision support software. The study compared the design of the radionuclide release written in the South Tangerang city nuclear contingency plan with the scenario of the radionuclide release from the JRODOS prognosis results. The study used a mixed method and a descriptive and analytical approach. In general, the results show that the prognosis of the radionuclides released by using JRODOS to some extent is similar to the scenario of the radionuclide release of the South Tangerang city nuclear contingency plan, both resulted in an affected area that reaches up to a 5 km radius from the GA Siwabessy reactor. The differences are in the specific affected areas due to different weather data, specifically, wind direction data. According to the condition and parameter used in the study, as discussed in the method section, the results show that the villages of Muncul, Kademangan, and Setu have higher affected risk rather than other towns. Nevertheless, for a more comprehensive analysis, further study is needed by using more event-time scenarios and by selecting the event-time within the range of the South Tangerang city nuclear contingency plan scenario’s event-time. Keywords: GA. Siwabessy reactor, nuclear contingency plan, radionuclide release scenario

Efficient CFD modelling of bulk condensation, fog transport and re-evaporation for application to containment scale

Formation of fog due to bulk condensation during a Beyond Design Basis Accident (BDBA) in a water-cooled nuclear reactor is of high safety relevance with regard to the flammability of the containment atmosphere and its interaction with several other safety relevant phenomena such as aerosol transport. Bulk condensation of steam in the presence of non-condensable gases is modelled using ‘return to saturation in constant timescale’ method in single-phase system as mass sink. The fog droplets formed as a result of condensation is transported as passive scalar in an Eulerian frame by considering the effects of convection by the gas mixture, turbulent and Brownian diffusion and droplet drift due to gravitation, drag and inertial forces. The model formulation is based on the assumption that fog volume fraction is fairly low (<1%) so that droplet–droplet interactions and the effect of droplets on the gas flow are negligible. The present work considers only spherical fog droplets of fixed diameter and evolution of droplet by coalescence and breakup is neglected. The effect of different droplet sizes is investigated in the validation cases to understand its consequence on the results. It is also assumed that the fog droplets are always in thermal equilibrium with the surrounding gas mixture. The re-evaporation of the transported fog droplets in regions where gas temperature is higher than saturation temperature is also incorporated into the solver. The model was numerically implemented to the containmentFOAM CFD package based on OpenFOAM. The verification of the phase change(condensation & evaporation) model by simulating the mixing of a cold and hot air–steam–fog mixture gave good agreement with Mollier diagram theory in terms of both final mixture temperature and quantity of steam and fog. The droplets drift transport model was validated by aerosol flow through a bent pipe investigating the effect of droplet Stokes number on deposition efficiency. The performance of the bulk condensation model was further validated and assessed by including the interaction with other relevant containment phenomena (e.g., wall condensation, multispecies transport) on SETCOM experimental facility (Forschungszentrum Jülich, Germany) and a technical scale THAI experiment (Becker Technologies, Germany). The results indicated that the inclusion of bulk condensation model improved the predictive capabilities of containmentFOAM, i.e., the consistency of simulation and measurements for these experiments.

Prediction of radioactivity releases for a Long-Term Station Blackout event in the VVER-1200 nuclear reactor of Bangladesh

Consequences of an anticipated Beyond Design Basis Accident (BDBA) Long-Term Station Blackout (LTSBO) event with complete loss of grid power in the VVER-1200 reactor of Rooppur Nuclear Power Plant (NPP) of Unit-1 are assessed using the RASCAL 4.3 code. This study estimated the released radionuclides, received public radiological dose, and ground surface concentration considering 3 accident scenarios of International Nuclear and Radiological Event Scale (INES) level 7 and two meteorological conditions. Atmospheric transport, dispersion, and deposition processes of released radionuclides are simulated using a straight-line trajectory Gaussian plume model for short distances and a Gaussian puff model for long distances. Total Effective Dose Equivalent (TEDE) to the public within 40 km and radionuclides contribution for three-dose pathways of inhalation, cloudshine, and groundshine owing to airborne releases are evaluated considering with and without passive safety Emergency Core Cooling System (ECCS) in dry (winter) and wet (monsoon) seasons. Source term and their release rates are varied with the functional duration of passive safety ECCS. In three accident scenarios, the TEDE of 10 mSv and above are confined to 8 km and 2 km for the wet and dry seasons, respectively in the downwind direction. The groundshine dose is the most dominating in the wet season while the inhalation dose is in the dry season. Total received doses and surface concentration in the wet season near the plant are higher than those in the dry season due to the deposition effect of rain on the radioactive substances.

Verification of GOTHIC Multivolume Containment Model during NPP Krško DBA LOCA

New containment multivolume model of NPP Krsko for GOTHIC code is developed. It is based on plant drawings and other available data. It is supported by developed SketchUp 3D containment model. The model is subdivided in volumes following physical boundaries and clearly defined flow paths. All important concrete heat structures are taken into account. Metal heat structures are based on plant’s SAR Chapter 6 licensing model. RCFC (Reactor Containment Fan Cooler) units are explicitly modelled as well as all main ventilation ducts. The model includes two trains of containment spray system. PARs (Passive Autocatalytic Recombiner) and PCFV (Passive Containment Filter Venting) filters added during plant safety upgrade project are part of the model too. It was intention to use model for both DBA (Design Basis Accident) and for DEC (Design Extended Conditions) and BDBA (Beyond Design Basis Accident) calculations. Based on the same discretization and data, and on experience acquired during GOTHIC model development and use, containment models for MELCOR and MAAP integral codes are developed too. As part of initial verification of the GOTHIC model containment DBA LOCA calculation is performed using SAR MER (Mass and Energy Release) data. The influence of different break positions on peak containment atmosphere pressure and temperature was studied. The results were compared against results obtained in single volume containment licensing model. Beside local effects due to different containment subdivision similar results are obtained when comparing containment dome from multivolume and the single volume in licensing model. Special attention was paid to distribution of water in lower part of the containment during recirculation phase. In this case much more valuable information are obtained in multivolume model with explicit volumes for main sump, recirculation sump and sump pit. Another point of interest was influence of containment spray duration on long term pressure and temperature behaviour. The intention was to study consistency of assumed different spray operation times used in safety analyses, EQ analyses and SAMGs and related consequences for plant operation during DBA LOCA.

Hydrogen explosion mitigation in DEMO vacuum vessel pressure suppression system using passive recombiners

An important issue for the Fusion Reactors is the hydrogen explosion hazard assessment. For this reason, in the Work Package of Safety Analyses and Environment (WPSAE) of the EUROFusion consortium, a task was established to identify the potential for hydrogen production in DEMO vessel in accident situations and to investigate the possible solutions which can prevent the risk of hydrogen and dust explosion. The hydrogen in the Vacuum Vessel (VV) can lead to combustion progressing in deflagration and detonation. Besides, the tungsten dust could enhance the effects of H2 reaction as demonstrated by experiments, however, in this first set of accidents, only the H2 generated by the chemical reaction is accounted for.For these reasons, the solutions to avoid or limit the H2 production and/or to control safely the risk of H2 accumulation and explosion need to be screened starting from the solutions adopted in the fission technology, and later on, adapted and assessed in the Pressure Suppression System (PSS) foreseen in DEMO. One promising idea seems to install passive catalytic recombiners in the PSS.The aim of this paper is firstly to set a range of operating parameters for the PAR functioning and intervention suitable in DEMO PSS configuration. Secondly to verify if the present criteria and parameters work efficiently in mitigating the H2 accumulation a PSS in accidental conditions. To test the theoretical effectiveness of the PAR intervention, two reference accidents have been analyzed: an in-VV Loss Of Coolant Accident (LOCA) in the frame of the Design Basis Accident (DBA) and a Loss Of Flow Accident (LOFA) without active plasma shutdown in the context of the Beyond Design Basis Accidents (BDBA).The results seem promising if the performances of the PAR will be confirmed in sub atmospheric and saturated conditions.

Development of the reactor core isolation cooling system model for the extended station black-out accident

To accurately predict the behavior of the Reactor Core Isolation Cooling System (RCIC) in the context of the FLEX strategies under the extended station black-out accident (SBO), we present a mechanistic model of the RCIC in this work. The control volume formulation of angular momentum conservation is employed to model the Terry turbine in the RCIC system. The two-phase nozzle flow model developed in the previous work is utilized to simulate the convergent-divergent nozzles in the Terry turbine. A nozzle flow table is generated to cover the operation range of the RCIC during beyond design basis accidents (BDBAs). The mechanistic RCIC turbine-pump model and the nozzle flow table are incorporated in a TRACE BWR model. Simulations of short-term and long-term SBO transients are performed to demonstrate dynamic RCIC responses. A system analysis code with the proposed mechanistic model of the RCIC system will have a better prediction of system safety response during SBO accidents. With the addition of such more realistic models, the system code will be capable of providing expanded understanding of safety margin, which can help help to form the technical basis for the successful implementation of the FLEX and SAMG measures under Extended Loss of AC Power (ELAP) conditions for BWR or PWR reactor designs.

Fuel for VVER and PWR: Current Status and Prospects

The Rosatom State Corporation activity in the nuclear fuel market requires the development of new designs and modifications meeting the current requirements of safety and operational reliability. The history of the development of a competitive fuel meeting heightened requirements of economic efficiency, safety, and operational reliability for NPP with VVER and PWR is presented. Ways to improve the Russian fuel for homegrown and foreign reactors are described. The development of fuel resistant to beyond-design basis accidents is considered.

Accident tolerant fuels (FeCrAl Cladding & Coating) performance analysis in Boiling Water Reactor (BWR) by the MELCOR 1.8.6 UDGC

Since the accident at Fukushima, one major goal of reactor safety research has been the development of more accident tolerant technologies that can mitigate or delay fuel degradation during a Beyond Design Basis Accident (BDBA). One major effort has been focused on increasing the capability of the fuel to be more tolerant of damage during an accident, i.e., Accident Tolerant Fuel (ATF) materials. In this work, we present the development of a generic BWR plant model, the modification of MELCOR to model ATF materials and the use of ATF materials (specifically FeCrAl alloy) as a coating on Zircaloy cladding or as a substitute material for cladding and fuel assembly canister material and its effect on severe accident progression, specifically, a Station Blackout accident. The analysis indicates that significant fuel degradation via fuel heat-up, clad oxidation, and hydrogen generation was delayed up to an hour if FeCrAl alloy was used as a clad and canister material. And, combined with the passive safety systems (i.e., the Reactor Core Isolation Cooling system, RCIC), the extended operation of these systems delayed fuel degradation further. However, an adverse effect should be emphasized for the monolithic FeCrAl design—it generated more hydrogen than the designs based on the Zircaloy due to the high reaction rate at a high temperature of FeCrAl. The design of the FeCrAl-coated-Zircaloy avoids this defect. Therefore, it is a promising choice to combine some of the beneficial traits of both materials.

Integrity Assessment of RBMK Physical Safety Barriers during Prolonged Station Power-Unit De-Energization

Two successive stages of the control of a beyond-design basis accident with de-energization of an RBMK reactor together with multiple equipment failures are analyzed: depressurization of the circulation loop and restoration of reactor cooldown with cold water as the reactor subsequently warms up. The factors affecting the pressure surge in the circulation loop on restoration of reactor cooling as the reactor heats up are examined. It is shown that the sequence of control actions and the employed equipment are effective from the standpoint of the possibility of overpressurization of the channel tubes and the reactor space. The state of the channel tubes and the reactor space in the presence of arising deformation loads was analyzed using an experimentally determined criterion for the acceptability (integrity) of a channel tube.

Design and development of a comprehensive program for the assessment and analysis of environmental effects due to the release of radioactive materials from the stack of nuclear installations: A case study in Bushehr nuclear power plant

The current investigation presents a comprehensive program called KIANA to assess and analyze the environmental effects due to the release of radioactive materials from the stack of nuclear installations. Bushehr Nuclear Power Plant Unit One (BNPP-1) is modeled using the KIANA software and the release of radioactive materials from the BNPP-1 stack on the members of the public through the airborne pathway is evaluated during normal operation as well as in the accident conditions including Design Basis Accidents (DBA) and Beyond Design Basis Accidents (BDBA). To verify and validate the KIANA software, the results of the KIANA are compared with the results of DOZA_M, ESTE, PC-CREAM 98, RECASS Express, and the values of the total effective dose monitored by the local detectors of BNPP-1. The KIANA is developed based on the Gaussian diffusion model written using C# programming language. In the current research, the total effective dose received by a member of the public due to the radioactive plume passage through the airborne pathway is calculated in the normal condition of BNPP-1. Moreover, the total effective dose in the case of the primary-to-secondary leakage inside the steam generators, the total effective dose in the case of Large Break Loss of Coolant Accident (LBLOCA) and Small Break Loss of Coolant Accident (SBLOCA), and the equivalent dose of the thyroid gland tissue for an infant group (1–8 years old) in the case of LBLOCA in DBA conditions are evaluated. Finally, the absorbed dose of the whole body of adults at the initial stage after a BDBA, absorbed dose of the thyroid gland tissue for an infant group (1–8 years old) at the initial stage after a BDBA, and the total effective dose in the first year after the accident in the case of a BDBA are assessed. The results of the KIANA software indicate a good agreement with the results of the DOZA_M, ESTE, PC-CREAM 98, RECASS Express computer programs, and the values of the total effective dose monitored by the local detectors of BNPP-1. The developed software has the potential of calculations of concentration and radionuclide dose received through all the exposure pathways such as airborne, foodstuff, marine, soil, animals, and vegetation without any restriction in normal and accident conditions, simultaneously.

THE VALIDATION OF COMPREHENSIVE SOFTWARE FOR THE ASSESSMENT OF RADIOLOGICAL IMPACTS ON HUMANS THROUGH BENCHMARK ANALYSIS WITH RADIOLOGICAL CODES.

In this paper, a novel domestic code called KIANA was developed for the assessment of radiological impacts on the population in normal and accident conditions including design basis accident (DBA) and beyond DBA (BDBA) for the nuclear power plants. The validation process of the KIANA code was performed using the results of the DOZA_M radiological code, whose results are presented in the Final Safety Analysis Report (FSAR) of the Bushehr Nuclear Power Plant Unit One (BNPP-1). The calculations of KIANA are performed based on the Gaussian diffusion model. The developed KIANA code has the potential of calculating the concentration and radionuclide doses due to the pathways such as airborne, foodstuff, marine (both one and two boxes models), soils, animals, vegetation (with and without tritium) and other pathways without any restriction. In the current research, the individual dose from a cloud to the member of the public in the region of BNPP-1 in normal condition was calculated. Moreover, the total effective dose to the member of the public from the primary to the secondary leakage inside steam generators, large break loss-of-coolant accident (LBLOCA) and small break loss-of-coolant accident was calculated. Thyroid gland equivalent dose for the infant (1-8 years) in the case of LBLOCA at the BNPP in DBA conditions was also evaluated. Finally, the prevented dose at the initial stage for the whole body of adults after BDBA, prevented dose at the initial stage for the thyroid gland of children after BDBA and the effective dose during the first year after the accident (external body irradiation from presence in the area) in the case of BDBA are assessed. The KIANA simulation results showed a good agreement with the FSAR data of BNPP.

Preliminary sensitivity analysis for an ex-vessel LOCA without plasma shutdown for the EU DEMO WCLL blanket concept

Abstract In this early development phase of the DEMO design the uncertainty affecting many operational and design parameters can modify main outcomes of accident scenario aiming at studying the critical conditions for the vacuum vessel and the contiguous containment volumes. The aim of this paper is to perform a preliminary sensitivity analysis of an accident progression predicted by MELCOR code considering selected parameters as a figure of merit to predict possible code outcomes. The uncertainty band will be evaluated through sensitivity analyses programmed, collected and statistically manipulated through RAVEN software tool. MELCOR and RAVEN have been internally coupled through a new Python code interface developed by Sapienza University of Rome, to perform sensitivity and uncertainty quantification analyses during severe accident transient. The Beyond Design Basis Accident (BDBA) analysis of an ex-vessel loss of coolant accident (LOCA) for the water-cooled lithium lead (WCLL) blanket concept has been simulated with the fusion version of MELCOR 1.8.6 code. The postulated initiating event (PIE) is a double-ended break in the first wall (FW) cooling system distributor ring, with simultaneous failure of the plasma shutdown system. An in-vessel breach of the coolant system occurs because of FW failure, with consequent unmitigated plasma shutdown transient. Sensitivity analysis results have shown that the FW temperature at which plasma in-vessel breach occurs is strongly correlated with the mass of hydrogen produced. The same parameter has also an impact on the overall accident scenario, such as the trigger of VVPSS rupture disks and thus source term mobilization.

Characterization of Kanthal APMT and T91 oxidation at beyond design-basis accident temperatures

Limited information is available on the oxidation mechanism of accident tolerant claddings (ATC) Kanthal APMT and T91 at the onset of beyond design-basis accident (BDBA) conditions. We characterized the surface of these ATC alloys after steam and air exposure at 1200 °C for 2 h, defining the oxidation mechanism. Thickness and composition were analyzed with microscopy, Raman spectroscopy, and synchrotron diffraction. Our results demonstrate that APMT forms a compact and homogeneous α-Al2O3 layer when exposed to air or steam. T91 forms a heterogeneous porous layer, containing a mixture of Cr- and Fe-based oxides, whose composition changes with the exposure environment.