Design Extension Conditions Research Articles

Sensitivity Analysis for Design Extension Conditions of a VVER-1200 Power Unit

Sensitivity Analysis for Design Extension Conditions of a VVER-1200 Power Unit

Safety Analysis of the Fourth Unit of Balakovo NPP under the Design Extension Conditions

Safety Analysis of the Fourth Unit of Balakovo NPP under the Design Extension Conditions

Prolonged Operation of Hydro-Accumulators in VVER-1000/V446 Considering Flow Restriction Components

Hydro-accumulators, as a passive part of the emergency core cooling system, have been upgraded in different types of VVER-1000s. The number and performance of these systems are modified in newer types of VVER-1000s to operate at different pressures and flow rates. The VVER-1000/446, as a distinguished design from the old generation (VVER-1000/320), utilizes two stages of hydro-accumulators. The second-stage accumulators (SSAs) are dedicated to improving the performance of the VVER-1000/446 under design extension conditions (DECs) and delaying the onset of core damage. During the upgrading of the VVER-1000/446, some orifices have been embedded as flow restriction components (FRCs) on the injection line of the SSAs to increase the water injection time during DECs. This paper aims to assess the performance of the SSAs according to their orifice diameter changes during median-break (MB)– and large-break (LB)–loss-of-coolant accidents (LOCAs). Several cases of FRC arrangements were investigated to demonstrate the core damage times during MB-LOCA (200-mm break) and LB-LOCA (400-mm break) scenarios. The results show that although the current size of the installed FRCs on the VVER-1000/446 is quite suitable for the MB-LOCA, it is not a proper choice for the LB-LOCA. Therefore, this demonstrates that the core damage time is highly dependent on the FRC size arrangement. This paper indicates an arrangement of FRCs that can modify the plant’s response in both MB and LB LOCAs without active injections. Therefore, as an additional finding, the sequential arrangement of FRCs can increase plant resilience against a broad range of LOCAs without operating active emergency cooling systems.

Evaluations of radionuclide activity releases into environment during loss of coolant accidents using the ASTEC code in pressurized water reactors within design basis and design extension conditions

The work described in this paper was carried out within the R2CA (Reduction of Radiological Consequences of design basis and extension Accidents) project, funded in HORIZON 2020 and coordinated by IRSN (France). An increase of the level of Nuclear Power Plant (NPP) safety by consolidated and more realistic evaluations of the Radiological Consequences (RC) of Design Basis Accidents (DBA) and a strengthening of the assessments of the NPP safety levels by considering accidental situations more severe than those integrated in plant designs (i.e belonging to Design Extension Conditions domain) were the two main motivations behind this project.More specifically, the project aims at consolidating and/or refining the assessments of the radiological consequences of explicit accidental scenarios within Design Basis Accidents (DBA) and Design Extension Conditions (DEC-A conditions without significant fuel degradation) in Light Water Reactors (LWR) through the improvements of existing code predictability; the upgrading of calculation chains and methodologies; the development/refinements of models. Within the Work Package 2 of the project, coordinated by TRACTEBEL and dedicated to calculation methodologies, existing methodologies or calculation chains and simulation tools have been applied to run a first batch of calculations dealing with different reactor types: PWR (Pressurized Water Reactor), BWR (Boiling Water Reactor), VVER (Water-Water Power Reactor) and EPR (European Pressurized Reactor). Loss Of Coolant (LOCA) and Steam Generator Tube Rupture (SGTR) accidents have been selected for the exercise and bounding scenarios of the DBA and DEC-A domains have been analysed. The results of this first set of calculations will be used as a reference to quantify the gains obtained by the updated methodologies/simulation tools developed within the project.This paper describes the results of the first batch of calculations, performed with the ASTEC integral code, simulating LOCA scenarios (DBA and DEC-A categories) in a PWR 900 MWe with a focus on the predicted number of failed fuel rods and Source Term (ST) in the environment governing the RC. Limitations of the used approach are outlined, as well as the needs for further upgrading the calculation chains are proposed in the light of the improvement that was planned within the project in Work Package 3 (WP3: LOCA – Loss of Cooling Accidents), coordinated by IRSN and dedicated to the improvement of code models dealing with LOCA scenarios.

Uncertainty calculation of thermal hydraulic simulation of DEC-A accident scenario at VVER-1000 NPP using CIAU method

All thermal hydraulic best estimate simulations of transient scenarios at Nuclear Power Plants (NPPs) are subject to uncertainties. One method to evaluate those uncertainties is implemented by the ”Code with the capability of Internal Assessment of Uncertainty” (CIAU-Code), which calculates the uncertainty for the parameters primary system mass, primary system pressure and hot rod temperature by accuracy extrapolation.The present work now presents an extended use of the CIAU method, by deriving the uncertainty of the parameter iodine-131 release to the environment. For the analysis, a steam generator hot header break with the assumption of a stuck open safety relief valve (BRU-A) at the first opening at the VVER-1000 reactor was selected. Such transient constitutes a design extension condition A (DEC-A), characterized by multiple failures of safety systems but with the reactor core remaining intact. The thermal hydraulic best estimate simulation was conducted with RELAP5-3D. Additionally, the Iodine Spiking (IS) phenomenon for the examined accident scenario was simulated using two different empirical IS models (NRC IS Model and BOKU University IS Model). To allow the application of the CIAU method for the IS calculation a theoretical derivation is provided.The results show that for the selected DEC-A scenario a high uncertainty range has to be expected. The analysis allows a conservative estimation of the risk expected in the event of the examined transient sequence.

Safety evaluation of Multiple Steam Generator Tube rupture accident using the best estimate plus uncertainty approach

Following the disaster in Fukushima Nuclear Power Plant (NPP) in 2011, the awareness of securing the safety of NPP under extreme events has been raised. For this purpose, the concept of Design Extension Conditions (DECs) was introduced, to enhance the plant’s capability to withstand events that are more severe than Design Basis Accidents (DBA), such as Multiple Steam Generator Tube Rupture (MSGTR) accident, given its rapid progression and significant potential for large release of radioactive materials to the environment. It is worth noting that MSGTR has never occurred during the commercial operation of nuclear power plant, therefore the knowledge of consequences of the event and potential mitigation strategy is limited. In this study, a postulated MSGTR (simultaneous rupture of 5 U-tubes in a single SG) accident is analysed using the best-estimate thermal hydraulics code RELAP5/SCDAPSIM/MOD3.4. The main focus of this work is evaluation of appropriate operator mitigation actions under various inherent uncertainties. To this end, the best estimate plus uncertainty (BEPU) approach is adopted, whereby RELAP5/SCDAPSIM/MOD3.4 was coupled to the statistical analysis software, DAKOTA, to conduct the uncertainty quantification (UQ). Using BEPU analysis, an ensemble of system responses under various epistemic and aleatory uncertainties is obtained and used to validate the efficacy of operator actions in bringing the plant to a safe shutdown condition.

Safety margin quantification by integrating probabilistic and deterministic safety assessments — Application to design extension conditions

Various safety analysis methods have been used over the years to quantify safety margins: at the start by Deterministic Safety Analysis (DSA) with conservative models or best estimate models and “bounding” assumptions; then by the Best Estimate Plus Uncertainty (BEPU) approach which takes into account uncertainties related to physical and modelling parameters; and recently by the Extended BEPU (EBEPU) approach with the intention of incorporating information from Probabilistic Safety Assessment (PSA).By combining probabilistic and deterministic safety assessments while taking into consideration a wide variety of uncertainties, the current work aims at the development of an Integrated Safety Margin Quantification (ISMQ) technique to extend the scope of existing approaches. Multiple aspects of the “safety margin” can be handled by this ISMQ methodology, including pertinent Initiating Events (IE) and sequences, the distance between best-estimate load and safety limit, and the likelihood of exceeding the safety limit.Together with the suggested methodology, an industrial application on Design Extension Conditions (DEC) of a typical Generation II 1000 MWe Pressurized Water Reactor (PWR) is presented to illustrate the approach and the anticipated outcomes. This ISMQ methodology is also applicable for safety analyses of other reactor designs including Small Modular Reactors (SMRs).

Contribution of IAEA coordinated research projects to light water reactors advanced technology fuel testing and simulation

The Fukushima-Daiichi accident in 2011 highlighted the pressing need for enhanced design and safety analysis of nuclear fuels, especially under accident conditions in nuclear power plants (NPPs). To address this need, the international nuclear fuel community has developed several concepts of Accident Tolerant and Advanced Technology Fuels (ATFs) for light water reactors. The International Atomic Energy Agency (IAEA), at the request of its Member States, has initiated a series of three Coordinated Research Projects (CRPs) to aid in the development and testing of these ATFs. These projects, known as FUMAC, ACTOF, and ATF-TS, focus on the experimentation and simulation of ATFs under various accident conditions, including design basis accidents and design extension conditions. This paper provides a comprehensive overview of the main activities, as well as the achieved or anticipated results of these three CRPs. The insights and advancements gained from these projects will enable IAEA Member States to further refine nuclear fission technology, thereby contributing to efforts aimed at mitigating climate change.

Implementation of the design extension conditions concept in the Czech Republic regulatory framework and NPPs

The paper describes evolution of the concept of design extension condition (DEC) and its implementation in the Czech Republic. The concept of design extension condition had been introduced in the early 1990ies and later was accelerated after the Fukushima accident in 2011. Today, the concept is elaborated in the EUR document, in the WENRA Reference Levels, and also in the IAEA Safety Standards. A number of countries are implementing the DEC concept into the national regulatory framework, safety analyses reports, plant design modifications, and the accident management procedures. Typically, the DEC events are divided into two groups: the DEC without severe fuel damage (DEC-A) and the DEC with severe fuel damage (DEC-B). Information on the implementation of the DEC concept in the Czech Republic, an overview of relevant plant modifications, and examples of safety assessment are given in the paper.

Iodine source term assessment as result of iodine spiking and mass transfer phenomena during a SGTR transient using MELCOR 2.2 and CATHARE 2 codes

The European project denominated Reduction of Radiological Consequences of design basis and design extension Accidents (R2CA) was launched in September 2019, with a very broad participation: 11 countries, 17 participating organisations, including international organisations, utilities, regulators, technical support organisations, researchers and developers and was coordinated by IRSN. The main goal of the project was to assess the conservatisms in the radiological releases calculations in nuclear power plant (NPP) studies.The work here presented is focused on developing new calculation methodologies and updating computer code models to carry out more detailed assessments of source terms resulting from a steam generator tube rupture (SGTR) accident in the Design Extension conditions A (DEC-A) domain. For this purpose, participants in Work Package 2 (WP2) of R2CA developed and implemented simplified models in the Severe Accident (SA) and Thermal-Hydraulic (TH) codes that take into account iodine spiking and iodine transport phenomena within primary and secondary circuit. These methodologies will not only provide a better estimation of the radiological consequences, but should also serve to improve accident management procedures, innovative instrumentation development and early detection tools.The new approaches are tested in a SGTR + Steam Line Break Outside Containment scenario without significant fuel degradation in a three-loop Western 1000 MWe PWR. During the transient, operators are also assumed to implement emergency operating procedures (EOPs) in line with Westinghouse EOPs to limit the release of radioactivity and control the evolution of plant parameters.The calculations are performed by Tractebel and CIEMAT with the American SA tool MELCOR and Bel-V with the French TH code CATHARE.The results highlighted some limitations of implemented models in predicting iodine behaviour as well as small discrepancies in the TH evolution of the transient, both of which were analysed and discussed in detail.

Total Loss of Feedwater Analysis of Pressurized Water Reactor Using RELAP5

Abstract In Europe, the design extension conditions (DEC) were introduced after the Fukushima Dai-ichi accident as preferred method for giving due consideration to the complex sequences and severe accidents without including them in the design basis conditions. The objective of the study is to determine available elapsed time before core uncovery and needed DEC safety features for total loss of all feedwater (TLOFW) in a two-loop pressurized water reactor. RELAP5/MOD3.3 computer code has been used for calculations. The initiating event for TLOFW is multiple failures in which, in addition to the loss of main feedwater, the auxiliary feedwater is also lost. The scenarios without DEC safety features and the scenarios with DEC safety feature assumed have been simulated. The results showed that after TLOFW event initiation, it is very important to trip the reactor as soon as possible. In case of loss of offsite power, the reactor coolant pumps stop, and the reactor very quickly trips on low reactor coolant pump flow. When normal operation systems are assumed, the reactor trip occurs on low-low steam generator narrow level few tens of seconds after accident initiation, resulting in less time available before core uncovery occurence. The results for TLOFW scenarios with normal operation systems and DEC safety features assumed demonstrated that secondary side bleed and feed can prevent core uncovery in case when no operator actions are credited in the first 30 min of event. When primary side bleed and feed is used, less time is available for operator actions.

Methods for the radioactive release estimation under DEC-A conditions

After the Fukushima Daiichi Nuclear Power Plant accident, the importance was raised to strengthen the assessment of the safety level of Nuclear Power Plants (NPP) by considering situations more severe than those integrated during the design of the plants. Design Extension Conditions (DEC) term was introduced by IAEA and WENRA. One of the most important subjects in the analysis of these conditions is the evaluation of radiological consequences.The Reduction of Radiological Accident Consequences (R2CA) collaborative project started in 2019 in the frame of the Horizon-2020 Program of the European Commission. The project addresses a broad scope of LWR designs (Gen II, III and III +) through the analyses of bounding scenarios of Loss Of Coolant Accidents (LOCA) and Steam Generator Tube Rupture (SGTR) transients and explores DBA and DEC-A conditions. The Lithuanian Energy Institute is taking part in this project as well.The part of work provided in the frame of R2CA project is presented in this article. The generic BWR-4 type reactor with MARK-I containment was analyzed in the case of DEC-A conditions. In this analysis the main thermal hydraulic processes, fission product release and its transport from the broken loop through containment to the environment were investigated. The number of ruptured fuel assemblies is the key parameter determining the fission product release from the core. The methodology which would help more precisely evaluate the number of failed fuel assemblies is presented in this work. The methodology consists of the application of integral severe accident code ASTEC and fuel performance code TRANSURANUS. Analysis using ASTEC code showed the importance of the reactor core nodalisation scheme. Taking iterative calculations, the relative power to which fuel assemblies remain intact at the selected DEC-A condition scenario was obtained. These results were verified with TRANSURANUS code calculations considering the uncertainties of the most relevant parameters. Based on TRANSURANUS calculation results, the ASTEC core nodalisation was improved. Improved ASTEC model nodalisation allowed to calculate more realistic rates of fission products releases from the fuel to the coolant, from the coolant to the containment and as well from the containment to the environment.

Assessment of CATHARE code against DEC-A upper head SBLOCA experiments

Design Extension Conditions (DEC)-A assessments of the operating nuclear power plants are generally considered for the purpose of getting additional safety demonstrations of their capability to undergo conditions that are generally more severe than DBAs by features implemented in the design and accident management measures. The pursued methodology is generally based upon Best Estimate approaches aiming at verifying that the safety limits in terms of integrity of the barriers against eventual large or early releases of radioactive material are fulfilled. These aspects are nowadays being experimentally and analytically addressed within the OECD/NEA experimental projects like the ATLAS and PKL series where a set of DEC-A experiments are considered. In this paper, experiments related to SBLOCA at the vessel upper head of the pressurized vessel of ATLAS and PKL are analytically assessed using the CATHARE code. These experiments includes issues related to common cause failure of the safety injection system and operator actions for preventing core excessive overheating. It is shown that, on the one hand, the safety features embedded in the design together with the operator actions are capable to prevent the progression towards a severe accident state and on the other hand, the code prediction capabilities for such scenario are generally good but still to be enhanced.

Experimental simulation of selected design extension condition scenarios without core meltdown in the CODEX facility

Two DEC-A scenarios were simulated with electrically heated 7-rod bundles. The test parameters were selected on the basis of reference scenarios from VVER-440 safety analyses.The CODEX-SBO experiment simulated a station blackout event in two separated phases. In the first phase the arrival of cold water from hydroaccumulators stopped the further heat-up of the fuel rods. In the second phase the heat-up continued up to 1900 °C and the test was terminated with the injection of emergency coolant from the recovered active injection system.In the CODEX-NITRO experiment it was supposed that after emptying of hydroaccumulators nitrogen enters the primary coolant and the test was focused on the interaction of hot fuel rods with nitrogen. The presence of nitrogen at high temperature (up to 1700 °C) resulted in the acceleration of chemical reactions, including nitriding of zirconium components.

Backfitting of Dukovany NPP for design extension conditions

The contribution deals with the backfitting of Dukovany NPP for design extension conditions with emphasis on severe accidents. This NPP, located in Czech Republic, is equipped with 4 units of VVER-440/V213 reactors. During the last decades significant attention was paid to backfitting of these units. New safety systems aimed at mitigating the consequences of severe accidents have been designed and have been (are being) implemented. At first, special features of VVER-440/V213 reactor and containment equipped with bubbler condenser pressure suppression system are briefly described. The systematic process of plant backfitting for design extension condition is summarized further. Attention is paid to description of crucial severe accident management strategies that have been adopted and design and implementation of new safety systems that are necessary to fulfil these strategies. Than the requirements of Czech regulatory body (SUJB) set for practical elimination of early or large radioactive releases are discussed. Finally, the performances of severe accident management strategies are illustrated using the specific case of blackout scenario. For this purpose, an analysis was performed using ASTEC code developed by the French IRSN. The results of analysis are briefly described and discussed further. From the obtained results it follows that the high-pressure melting scenario is reliably avoided, corium is localised and stabilised inside the reactor vessel and the hydrogen in containment atmosphere is safely controlled by passive autocatalytic recombiners. Thus, stable plant state with preserved containment integrity is reached. Source term that has been estimated under these conditions indicates that the predicted releases are well below the criteria set by SUJB for early or large radioactive releases, and therefore long-term contamination of the environment is avoided.

DEC safety evaluations and challenges in Belgium

After the Fukushima-Daiichi accident, the need to consider safety analyses under design extension conditions of the nuclear power installations has largely increased. The objective is to reassess and investigate the possibilities of the current nuclear installations to withstand accidental conditions that were not originally covered by the design basis accident. In Belgium, new regulation introduces the Design Extension Conditions (DEC) in the requirements for the design and operation of nuclear power plants. The purpose is to assess the safety of the current operating nuclear power installations and to identify design improvements, procedural measures and additional practicable provisions or mitigation means that could be used to prevent severe accidents. This paper provides a brief overview of the issues and challenges related to DEC safety evaluations in Belgium for the Licensee and also for the Safety Authority.

A commensurate industry approach to DEC A safety analysis

The new revision of the WENRA reference level 2014 (WENRA, 2014a) and the IAEA safety standard (International Atomic Energy Agency, 2016) require that accidents under design extension conditions without significant fuel degradation (DEC A) be identified and safety analysis be performed. To that aim, Tractebel has developed a practical method to concretely address the high level requirements of WENRA RL 2014 and IAEA SSR 2/1 and SSG2 guidelines. For the selection of DEC A sequences, a global methodology using PSA, internal and external hazard as well Combination of Events screening and characterization methods has been derived. For the analysis of DEC A, a graded approach has been derived. It aims at optimizing engineering effort for uncertainties treatment evaluation as well as decision making process regarding potential plant improvements resulting from the DEC A safety analyses. Some typical results of the applications are presented as example.

Safety analysis of representative accidental transients for a 100 MWe small modular LBE-cooled reactor

Nuclear energy plays an important role in the green low-carbon energy. The lead-based reactor is one of the most popular small modular reactor types. The pre-conceptual design of a 100 MW(e) small modular lead-bismuth eutectic cooled reactor was developed by Shanghai Nuclear Engineering Research & Design Institute CO., LTD. to meet the market demand of advanced small-scale nuclear plant. Representative accidents were analyzed for the 100 MW(e) small modular lead-bismuth reactor using ATHLET/MOD 3. The result shows that under the design basis conditions, the performance of the reactor can meet the safety criteria with the protection system working normally. The coolant solidification in the primary circuit under these transients should be prevented. Under the design extension condition, the highest cladding temperature is much lower than the melting point. In the unprotected reactivity insertion condition, there may be a short time of local melting of the fuel, which also meet the acceptance criteria. The safety analysis demonstrated the passive safety of the design with the negative temperature coefficient, strong natural circulation capability, and the passive residual heat removal system.

DEMO Divertor Cassette and Plasma facing Unit in Vessel Loss-of-Coolant Accident

As part of the pre-conceptual design activities for the European DEMOnstration plant, a carefully selected set of safety analyses have been performed to assess plant integrated performance and the capability to achieve expected targets while keeping it in a safe operation domain. The DEMO divertor is the in-vessel component in charge of exhausting the major part of the plasma ions’ thermal power in a region far from the plasma core to control plasma pollution. The divertor system accomplishes this goal by means of assemblies of cassette and target plasma facing units modules, respectively cooled with two independentheat-transfer systems. A deterministic assessment of a divertor in-vessel Loss-of-Coolant Accident is here considered. Both Design Basis Accident case simulating the rupture of an in-vessel pipe for the divertor cassette cooling loop, and a Design Extension Conditions accident case considering the additional rupture of an independent divertor target cooling loop are assessed. The plant response to such accidents is investigated, a comparison of the transient evolution in the two cases is provided, and design robustness with respect to safety objectives is discussed.

Finite Element Analysis and simulation of missile impact on Nuclear Reactor Containment Structure

Nuclear Powerplants are planned and designed to withstand high internal, external pressures and impact loads. International Atomic Energy Agency recommends Design Extension Condition (Impact of Missile or Aircraft) that is a mandatory condition to be fulfilled by Containment structure. In this Finite Element Analytical study, wall joint section of containment structure is modelled and analysed for Missile body impact, and it is observed that, the surface induced pressure in wall section is dynamic in nature, as it varies with respect to time and load. By obtained 3D simulation and contour pattern on PCC panel, it is found that the induced stress is bending pressure.