European Pressurized Reactor Research Articles

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.

Brand New or More of the Same Nuclear? (De)Constructing the Economic Promise of the European Pressurised Reactor in France and the UK

ABSTRACT Technological innovation needs construction of promises and expectations to mobilise resources and supportive networks, yet exaggerated promises risk leading to disappointment and undermining this very support. Drawing on an analysis of secondary literature and press articles, the concepts of hype cycle and Regimes of Economics of Techno-scientific Promises (ETP) are applied to examine the construction of the largely failed promise of the European Pressurised Reactor (EPR), designed to spearhead a French-led ‘nuclear renaissance’ in the 1990s. The debates on the EPR economics in France and the UK illustrate the country-specific features that condition the ability of an incremental in-between innovation, in an archetypically ‘modernist’ field of technology, to survive in today’s ‘presentist’ era of shrinking timeframes. The phase of disillusionment depicted in the hype cycle can better be described as two country-specific processes whereby the initial promise was continuously modified and requalified in order to maintain its legitimacy and credibility. As an incremental innovation, the EPR continues to struggle between the contrasting needs of demonstrating radical novelty and experience-based continuity. This tension is accentuated by the country-specific legacies and imaginaries, including the historically shaped ideological trust in the state and the market.

Implications of different actinide recycling options for EPR reactors on resource utilization and resultant high level waste disposal

The objective of this study is to investigate actinide recycling options that are applicable in thermal reactors with existing fuel production technologies and to assess impacts on the resources needed and resultant high level wastes. European Pressurized Reactor (EPR) design is considered as reference thermal reactor and fuel cycles with different Transuranics (TRU) recycling approaches are examined in order to determine which of the options is most effective in reducing the load of spent fuel disposal and in better use of resources. In the study, the comparison of recycling scenarios in terms of resource utilization and waste stream characteristics is based on production of 1 TWh electricity at the equilibrium fuel cycles. Comparative study results show that, among the various options investigated, fuel cycle with monorecycling of americium has least natural uranium requirement and disposal burden and can be applied as a back-end fuel cycle strategy for thermal reactors to reduce the rate of increase of TRU inventory until the development of advanced technologies for the efficient transmutation of TRU.

The greenhouse gas emissions of nuclear energy – Life cycle assessment of a European pressurised reactor

Nuclear energy contributes ~10% of the global electricity generation and different views exist on its carbon-intensity and sustainability. Context is crucial to determine the sustainability of new nuclear power generators, making the existence of a global answer to the unresolved question unlikely. This study aims to establish the life-cycle greenhouse gas emissions associated with nuclear energy in Europe given ongoing construction of nuclear generators. Due to the high uncertainty and complexity that characterise construction and operation of nuclear generators, we adopt a multi-method, scenario-based approach. The three methods used are: process-based, input-output, and hybrid life cycle assessment. Scenarios account for different total energy outputs over the life cycle of the nuclear generator, different end of life options, and different sectoral allocations of costs in the input-output calculus. Results for the process-based, input-output, and hybrid methods range between 16.55–17.69, 18.82–35.15, and 24.61–32.74 gCO2e/kWh, respectively. These are either well above or at the upper end of the range of possibilities (5 to 22 gCO2e/kWh) stated in a report for the UK’s Committee on Climate Change, and significantly higher than the median value of 12 gCO2e/kWh presented by the Intergovernmental Panel on Climate Change. They are also higher than the values acknowledged by the nuclear industry. Given the severe potential lock-in effects of today’s energy choices for future generations, this research questions the role of nuclear energy to meet the UN Sustainable Development Goals and calls for further scrutiny on its sustainability and environmental viability.

Validation of NECP-Bamboo with BEAVRS and AP1000 measurements

PWR-core fuel management code system NECP-Bamboo has been validated by about one hundred commercial PWR-core cycles including both the Second and Third Generations such as Chinese Nuclear Power CNP300 and CNP600, Chinese Pressurized Reactor CPR1000 and ACPR1000, France Framatome M310 and European Pressurized Reactor (EPR), USA Westinghouse AP1000, Chinese CAP1400 and Hualong-1, etc. Among those, the last four belong to the Third Generation PWRs. Besides, BEAVRS is the first worldwide well-known international benchmark that provides a complete two-cycle description of a commercial PWR-core, including detailed operating data and measurements, while the Unit 1 of Sanmen Nuclear Power Plant (NPP) in mainland China is the worldwide first AP1000 core. Thus, in this paper, measurements of these two PWR-cores were chosen to validate the code NECP-Bamboo, including critical boron concentration (CBC), control rod worth, temperature coefficient, detector response and boron letdown curve. Satisfying agreements have been found between the simulation with NECP-Bamboo and the corresponding measurements.

Investigative Study of Neutronic Safety Parameters of HPR and EPR Using the MCNP Code

As part of technology assessment of proposed commercial nuclear power reactor technologies for Ghana’s Nuclear Power Programme, the neutronic safety parameters of the European Pressurized Reactor (EPR) and High Temperature Pressurized Reactor (HPR) reactor technologies are theoretically analyzed and compared. The MCNP neutronic code was employed as a computational tool to analyze the reactivity temperature coefficients, moderator void coefficient, criticality, and neutron behavior at various operating conditions. The HPR, which is still under construction and under theoretical safety analysis, showed good inherent safety features comparable to the already existing EPR technology.

Neutronic safety analysis of proposed reactor technologies for Ghana’s nuclear power plant using the MCNP code

In pursuance of sufficient, stable and clean energy to solve the ever-looming power crisis in Ghana, the Nuclear Power Institute of the Ghana Atomic Energy Commission has on the agenda to advise the government on the nuclear power to include in the country's energy mix. After consideration of several proposed nuclear reactor technologies, the Nuclear Power Institute considered a high pressure reactor or vodo-vodyanoi energetichesky reactor as the nuclear power technologies for Ghana's first nuclear power plant. As part of technology assessments, neutronic safety parameters of both reactors are investigated. The MCNP neutronic code was employed as a computational tool to analyze the reactivity temperature coefficients, moderator void coefficient, criticality and neutron behavior at various operating conditions. The high pressure reactor which is still under construction and theoretical safety analysis, showed good inherent safety features which are comparable to the already existing European pressurized reactor technology.

Investigation of fuel cycles containing Generation IV reactors and VVER-1200 reactors

Abstract Gen-IV fast reactors are envisaged to operate in closed fuel cycles due to their ability to breed their fuel from fertile feed and burn minor actinides produced by themselves or thermal reactors in the nuclear park. The optimization of such fuel cycle strategies requires detailed models, capable of simulating the transition from initial state to equilibrium. A fast and flexible burn-up scheme based on polynomial fitting of the one-group cross-sections, FITXS was used to develop burn-up models for the Gen-IV Gas-cooled Fast Reactor (GFR), Lead-cooled Fast Reactor (LFR) and Sodium-cooled Fast Reactor (SFR), as well as a European Pressurized Reactor (EPR) and a VVER-1200 MOX fuel assembly. The burn-up models were integrated in a closed fuel cycle model containing Gen-IV LFR and MOX fueled VVER-1200 reactors, and different scenarios were investigated and compared concerning the reduction of transuranium inventories and the stabilization of the plutonium inventory. Results show that the LFR is capable of burning minor actinides from spent VVER-440 fuel and that transuranium inventories can be stabilized or reduced with a mixed fleet of LFR and MOX fueled VVER-1200 reactors.

Neutronic design of homogeneous thorium/uranium fuel for 24 month fuel cycles in the European pressurized reactor using MCNP6

Abstract A new approach to designing thorium based fuels for a European pressurized reactor (EPR) core is introduced by considering a three dimensional full core MCNP6 model. Existing information on designing thorium/uranium fuel for Generation II pressurized water reactors (PWRs) have been analyzed and the new fuel was proposed for the EPR (Gen III+). The newly designed fuel assembly (Th-B1) can reach 24-month fuel cycles without altering the geometry or disregarding any design limits. The Th-B1 was developed by reducing and ultimately removing Gd2O3 and slightly reducing the overall initial fissile content compared to the uranium only EPR. The same overall initial fissile content of the U-EPR was maintained, with the exception for fuel pin sections where pure ThO2 replaced the UO2 and Gd2O3 containing sections of the original U-EPR design. Results showed that there is no need to increase the initial fissile content or soluble boron enrichment as predicted by previous studies. Both xenon and samarium fission product yields for the Th-B1 are lower than for the U-B1, which contributes to a higher value for k ∞ and better breeding. Results also showed that the total fissile content for the Th-B1 shows a smaller decrease with time, which implies a smaller decrease for k ∞ with time and subsequently an extension of the fuel cycle. The innovative optimized fuel design of the EPR (five axial zones), enabled fuel enrichment savings and longer fuel cycles for Th-EPR. Not only does one save on fuel costs by keeping the overall initial fissile content similar, but this also results in reactor properties that are similar to those of the U-EPR and no serious changes need to be made. The only concerns are a lower delayed neutron fraction and boron worth for the Th-B1 at EOL. However, not all assemblies in a full core will be at the end of their life and if a three batch reloading scheme is assumed, the boron worth and delayed neutron fraction will most likely be within acceptable limits. Other differences between the Th-EPR and U-EPR such as the axial power profile and difference in boron letdown curve can be accounted for by amending the operating procedures during the design phase.

Developing a Full-Core MCNP6 and RELAP5 Model of the European Pressurized Reactor Using NWURCS

The European pressurized reactor (EPR) is classified as a Generation III+ reactor. It differs from a conventional pressurized water reactor in many aspects, one of which is the core design. This evolutionary reactor lends itself to new fuel designs, such as thorium-based fuels. To perform new design calculations, a base case model needs to be established because the detailed models that are currently available are either proprietary or regulated. This paper therefore presents such a model based on the Monte Carlo method. This method is a valuable component of reactor neutronic calculations because geometry and materials can be accurately modeled.We modeled a full core of the EPR using MCNP6, in which the individual fuel pin geometry and material definitions were used together with radial and axial temperature characterization based on fuel assemblies considered as nodes. Data for both the neutronic and thermal-hydraulic models were mainly obtained from the U.S. EPR Final Safety Analysis Report (FSAR) [Rev. 5, AREVA (2013)].The neutronic and some thermal-hydraulic results were compared with data from the EPR FSAR. The following core neutronic parameters compared well with the FSAR data: the boron worth, axial flux distribution, neutron flux spectrum, reactivity coefficients, and control rod worth. However, the delayed neutron fraction showed a somewhat larger difference compared to the FSAR. Given this verification with the FSAR, confidence in the MCNP6 EPR model was therefore established. The model that we have developed serves as the basis for the follow-on study of introducing thorium in the EPR core.

Influence of accident tolerant fuel cladding material on the European pressurised reactor core neutronic characteristics

Research on accident tolerant fuel became an important research area to improve safety of the nuclear reactors, especially following the Fukushima disaster. Numerous studies have demonstrated that silicon carbide is a promising candidate material that could potentially replace the zirconium cladding in LWR reactors, which is currently under active research by multiple research and industry organizations. Many of material testing of silicon carbide were already performed and the results are available. Also molybdenum is considered as fuel cladding. This paper is focused on the potential of using silicon carbide and molybdenum as a fuel cladding in the European Pressurized Reactor (EPR) core from the neutronic point of view.

Assessment of the Anticipated Environmental Footprint of Future Nuclear Energy Systems. Evidence of the Beneficial Effect of Extensive Recycling

In this early 21st century, our societies have to face a tremendous and increasing energy need while mitigating the global climate change and preserving the environment. Addressing this challenge requires an energy transition from the current fossil energy-based system to a carbon-free energy production system, based on a relevant energy mix combining renewables and nuclear energy. However, such an energy transition will only occur if it is accepted by the population. Powerful and reliable tools, such as life cycle assessments (LCA), aiming at assessing the respective merits of the different energy mix for most of the environmental impact indicators are therefore mandatory for supporting a risk-informed decision-process at the societal level. Before studying the deployment of a given energy mix, a prerequisite is to perform LCAs on each of the components of the mix. This paper addresses two potential nuclear energy components: a nuclear fuel cycle based on the Generation III European Pressurized Reactors (EPR) and a nuclear fuel cycle based on the Generation IV Sodium Fast Reactors (SFR). The basis of this study relies on the previous work done on the current French nuclear fuel cycle using the bespoke NELCAS tool specifically developed for studying nuclear fuel cycle environmental impacts. Our study highlights that the EPR already brings a limited improvement to the current fuel cycle thanks to a higher efficiency of the energy transformation and a higher burn-up of the nuclear fuel (−20% on most of the chosen indicators) whereas the introduction of the GEN IV fast reactors will bring a significant breakthrough by suppressing the current front-end of the fuel cycle thanks to the use of depleted uranium instead of natural enriched uranium (this leads to a decrease of the impact from 17% on water consumption and withdrawal and up to 96% on SOx emissions). The specific case of the radioactive waste is also studied, showing that only the partitioning and transmutation of the americium in the blanket fuel of the SFR can reduce the footprint of the geological disposal (saving up to a factor of 7 on the total repository volume). Having now at disposition five models (open fuel cycle, current French twice through fuel cycle, EPR twice through fuel cycle, multi-recycling SFR fuel cycle and at a longer term, multi-recycling SFR fuel cycle including americium transmutation), it is possible to model the environmental impact of any fuel cycle combining these technologies. In the next step, these models will be combined with those of other carbon-free energies (wind, solar, biomass…) in order to estimate the environmental impact of future energy mixes and also to analyze the impact on the way these scenarios are deployed (transition pathways).

Research and application of piping inside grinding robots in nuclear power plant

Technical installations especially that are subject to special safety criteria should always be kept at the best possible level of operation safety. This applies in particular to their innumerable pipeline systems which require a high level of reliability and safety, such as those in nuclear power plants.Taishan EPR (European Pressurized Reactor) nuclear power plant, which is in Guangdong province in China, has been the first one which succeeds to finish cold function test (CFT) among all the third generation nuclear power stations around the world.Before CFT, a lot of time has been spent to finish the piping inside grinding work, which is supposed to be the first case of nuclear pipe line inside grinding in China, according to RCC-M code and technical specification, to relieve the weld structure from mechanical constrains, specifically high fatigue stress.Inside grinding robots are custom made, from DN100 to DN500, including 4 types of robots, to carry out the inside grinding work in pipes instead of human beings when it is impossible for workers to achieve.

Modular construction mechanics of a European pressurized reactor steel containment liner

A European pressurized reactor (EPR) steel containment liner structure is comprised of the cylinder part and the dome part. An introduction of the steel liner structure is presented, followed by studies on the key mechanical features of the construction process using a refined finite element method. The steel liner was divided into several modules and then assembled during construction. Firstly, the equipment structure used to hoist the liner module was optimized, the lifting lug was analyzed using a multi-scale finite element model; the wind speed limit during lifting was also studied. Subsequently, the effect of internal forces during assembly between the liner modules, the lateral pressure of fresh concrete, the non-uniform temperature load, and the wind load on the cylinder module was analyzed. According to the time-varying structural performance during continuous concrete pouring and the hardening construction, an “overlapping element and birth-death element” technique was adopted to analyze the deformation and stress of the long-span steel dome liner. In addition, the stability-bearing capacities of the dome structure during construction were also studied, which took into consideration the effect of the initial geometrical imperfections and the elasto-plasticity of the material. This study presents a reference in terms of the mechanics of the construction scheme and the safety of such a type of structure.

FITXS: A fast and flexible burn-up scheme based on the fitting of one-group cross-sections

The closure of the nuclear fuel cycle is currently envisaged with the deployment of Generation IV fast reactors which are able to generate their fuel from fertile 238U or 232Th, and also burn minor actinides arising from legacy wastes and thermal reactors in the nuclear park. The optimization of such fuel cycle strategies requires detailed models, capable of simulating the transition from initial state to equilibrium. Due to the high computational cost of detailed burn-up calculations, most scenario codes use parametrized few group cross-sections to calculate fuel depletion in the reactors. This paper presents a fast and flexible burn-up scheme called FITXS which we applied on the GFR2400 Gas-cooled Fast Reactor and a European Pressurized Reactor MOX fuel assembly. Based on the fitting of one-group cross-sections as functions of the detailed fuel composition, the developed models are able to calculate spent fuel compositions with high accuracy for a wide range of initial compositions in less than one second computational time. In order to demonstrate applicability, the models were integrated into a fuel cycle model containing GFR2400 and European Pressurized Reactors, as well as conventional Light Water Reactors, and the fuel utilization and transmutational properties of the system were analyzed. It was found that the GFR2400 can operate in a closed fuel cycle, with positive breeding gain in the equilibrium. Increased minor actinide feed improves breeding in the GFR2400, but the accumulation of plutonium can be prevented by recycling the excess plutonium as MOX fuel in thermal reactors.

Coupled one-dimensional and CFD models for the simulation of steam generators

A computational procedure suitable for modeling both vertical and horizontal steam generators is presented. The procedure consists of a one-dimensional model for the whole steam generator and a CFD model for the secondary side. The three-dimensional two-phase CFD simulation of the secondary side utilizes the porous medium approximation and the mixture k-ε turbulence model. One-way coupling between the models is used, in which the surface temperatures of the heat transfer tubes are considered as a boundary condition for the CFD simulation. A stationary state and two pressure transients, a pressure rise and a pressure reduction, for the PWR PACTEL steam generator are calculated and the internal circulation on the secondary side is analyzed. PWR PACTEL is a test facility, located at Lappeenranta University of technology, in Finland, which was constructed for research activities associated with the European Pressurized Reactor (EPR).

Benchmark description of a stylized three-dimensional European Pressurized Reactor (EPR) problem

This paper describes a whole-core, three-dimensional, code-to-code benchmark problem based on the European Pressurized Reactor (EPR) developed by AREVA NP. The core specifications were taken directly from the Final Safety Analysis Report (FSAR) submitted to the Nuclear Regulatory Commission (NRC) and the reactor core was modeled in a stylized manner while maintaining full heterogeneity at the pin and assembly level. Three different fresh core control rod configurations are presented in this paper – All Rods Out (ARO), Some Rods In (SRI), and All Rods In (ARI). Detailed Monte Carlo results are presented and include core eigenvalues, pin fission densities, and assembly averaged fission distributions. The benchmark results together with the provided macroscopic cross section libraries are useful for testing the numerical accuracy of transport solution methods for criticality analysis. The benchmark description includes the material atom densities and other relevant parameters for cross section studies.

Thermal-hydraulic calculations for a fuel assembly in a European Pressurized Reactor using the RELAP5 code

Abstract The main object of interest was a typical fuel assembly, which constitutes a core of the nuclear reactor. The aim of the paper is to describe the phenomena and calculate thermal-hydraulic characteristic parameters in the fuel assembly for a European Pressurized Reactor (EPR). To perform thermal-hydraulic calculations, the RELAP5 code was used. This code allows to simulate steady and transient states for reactor applications. It is also an appropriate calculation tool in the event of a loss-of-coolant accident in light water reactors. The fuel assembly model with nodalization in the RELAP5 (Reactor Excursion and Leak Analysis Program) code was presented. The calculations of two steady states for the fuel assembly were performed: the nominal steady-state conditions and the coolant flow rate decreased to 60% of the nominal EPR flow rate. The calculation for one transient state for a linearly decreasing flow rate of coolant was simulated until a new level was stabilized and SCRAM occurred. To check the correctness of the obtained results, the authors compared them against the reactor technical documentation available in the bibliography. The obtained results concerning steady states nearly match the design data. The hypothetical transient showed the importance of the need for correct cooling in the reactor during occurrences exceeding normal operation. The performed analysis indicated consequences of the coolant flow rate limitations during the reactor operation.

Schutzzielorientiertes Design der Sicherheitsleittechnik

Safety instrumentation and control in nuclear power plants has to be systematically designed and revisable in order to meet the highest safety requirements. A safety objective oriented design process is presented in combination with the defence in depth philosophy. The approach is applicable for the construction of new plants like the European Pressurized Reactor (EPR) and for the comprehensive refurbishment of existing plants. Its success has been demonstrated in several domestic and international I&C projects.

Schutzzielorientiertes Design der Sicherheitsleittechnik

Safety instrumentation and control in nuclear power plants has to be systematically designed and revisable in order to meet the highest safety requirements. A safety objective oriented design process is presented in combination with the defence in depth philosophy. The approach is applicable for the construction of new plants like the European Pressurized Reactor (EPR) and for the comprehensive refurbishment of existing plants. Its success has been demonstrated in several domestic and international I&C projects.