Lifetime Of Nuclear Power Plant Research Articles

Impacts and challenges for Japanese nuclear operators as a result of legislative change to extend the legal lifetime of nuclear power plants in Japan

The purpose of nuclear regulation is not to impede the sound development of industry's competitiveness. Therefore, the regulations must be coherent, transparent and most of all, logical.On May 31st, 2023, a bill including rules to change the legal lifetime of nuclear power plants was approved and endorsed at the 211th ordinary session of the National Diet in Japan. An important characteristic of this legislative change is that it does not simply grant additional operating years to the original lifetime; rather, it considers the length of time that the plant was under shutdown status due to reasons not attributable to the plant owners and excludes this period from the countdown of the predetermined lifetime, thus extending the original legal lifetime.Three major problems could be listed in the procedures for changing the rules and in the new rules themselves. Firstly, the period during which this matter was discussed was too short to discuss issues raised by the change in detail. Secondly, the definitions of the period to be excluded from the predetermined lifetime are neither scientifically logical nor clear and would likely cause a great deal of confusion for both licensees and regulators when the new rule comes into effect. Thirdly, under this rule, the legal lifetime of a plant which performance is worse would be longer than that of the one with better operating performance. In particular, the second and the third problems deviate greatly from the purpose of nuclear regulation, which is not to impede the sound development of industry's competitiveness. The author strongly urges all parties involved to reconsider their decisions even after the bill has got an official approval at the National Diet.

Degradation of Concrete Structures in Nuclear Power Plants: A Review of the Major Causes and Possible Preventive Measures

Concrete, an integral part of a nuclear power plant (NPP), experiences degradation during their operational lifetime of the plant. In this review, the major causes of concrete degradation are extensively discussed including mechanisms that are specific to NPPs. The damage mechanism could be chemical or physical. The major causes of chemical degradation include alkali–aggregate reactions, leaching, sulfate attack, bases and acids attack, and carbonation. Physical degradation is a consequence of both environmental and mechanical factors combined. These factors are mainly elevated temperature, radiation, abrasion and erosion, salt crystallization, freeze–thaw distortions, fatigue and vibration. Additionally, steel reinforcements, prestressing steels, liner plates, and structural steel also experience degradation. The prospective areas in the structural components of the NPP where the degradation could occur are mentioned and the effective solutions to the causes of degradation are highlighted. These solutions are designed to enhance the physical and chemical characteristics of concrete. Some of the major recommendations include addition of mineral substitutes, use of low water-to-cement ratio as well as low water-to-binder ratio, use of low alkali cement, use of special aggregates and fibers, use of corrosion inhibitors, use of cathodic protection, etc. The review concludes with an overview of present methods and possible recommendations used to enhance the quality of concrete towards preventing concrete degradation and increasing the lifetime of NPPs.

Application of Distributed Optical Fiber Sensing Technology to the Detection and Monitoring of Internal Swelling Pathologies in Massive Concrete Blocks.

This paper presents an experimental application of Distributed Optical Fiber Sensors (DOFS) for the Structural Health Monitoring (SHM) of concrete structures affected by internal swelling pathologies. In the framework of a large research project aiming to assess the possible extension of the operating lifetime of nuclear power plants from 40 to 60 years, massive blocks were cast from reactive concrete mixtures intended to develop delayed ettringite formation and alkali–silica reaction. These blocks were subjected to specific ageing conditions to initiate and accelerate the concrete pathologies. Some of the blocks were instrumented with DOFS bonded to the surface and embedded in the concrete. Using an interrogator device based on Rayleigh backscattering and a suitable procedure to eliminate temperature effects, distributed strain measurements were then performed at different time intervals. The first results of this ongoing study made it possible to demonstrate the feasibility and effectiveness of this sensing technology for detecting and monitoring expansion induced by swelling pathologies in representative-scale concrete structures.

Multiphase Eulerian-Eulerian CFD supporting the nuclear safety demonstration

Nuclear safety demonstration is a key issue where thermal hydraulics plays a central role, whether at the design stage or for nuclear power plants lifetime extension. Combined with huge efforts dedicated to simulation software developments during the past decades, this context naturally leads to a progressive introduction of CFD (Computational Fluid Dynamics) tools to support the nuclear safety demonstration. However, this progressively increasing use of CFD also implies strong requirements related to nuclear safety issues. A declination of this methodology will show what is the current place of multiphase CFD in nuclear safety, and what are the necessary prerequisites to use it in this context. Including the development and validation of dedicated models, allowing for example, all-flow regimes transitions. An illustration of application on one of the most challenging case, involving all these flow phenomena with heat and mass transfer will be described to show where we are in terms of limit of usage for the safety demonstration purpose.

Environmental Fatigue Analysis of nuclear components within the framework of INCEFA-SCALE project

INCENFA-SCALE is a five-year project funded by the EC Horizon2020 programme, successor of the INCEFA-PLUS project. INCEFA-SCALE started in October 2020. The objective of this project is to improve the capacity to predict the lifetime of Nuclear Power Plant (NPP) components subjected to environmental assisted fatigue. The project starts off by analysing the existing data and then provides new environmentally assisted fatigue data which allow the laboratory test outcomes to be applied to components with real geometries and loads. So far, the data mining of different finished projects (INCEFA-PLUS, USNRC, EPRI, MHI and AdFaM) has been carried out, and test conditions for filling the knowledge gaps have been established. Moreover, the test matrix for 2022 has been defined. In this first phase, tests are focused on producing reference data, analysing complex waveforms (variable amplitude) and the effect of the surface finish. The next testing phases will focus on particular conditions: multi-axial tests, notches, stress/strain gradient effect and size effect. Furthermore, the microstructural analysis of common materials and a guideline for fatigue striation measurement on the fracture surface have been developed. This article provides an update on the project status and the advances made in data analysis, mechanical understanding and testing conditions.

The Carbon Value of Nuclear Power Plant Lifetime Extensions in the United States

Nuclear power is currently the single largest carbon-free source of electricity in the United States. The climate mitigation cost savings of the existing U.S. nuclear fleet is denominated in hundreds of billions of dollars [net present value (NPV)] based on an integrated assessment modeling of the U.S. energy system within a globally consistent framework. Lifetime extensions of the existing nuclear fleet from 40 years to 60 and 100 years resulted in $330 billion to $500 billion (all figures are in U.S. dollars) (NPV) of mitigation cost savings for the United States under a deep decarbonization scenario consistent with limiting global temperature change to 2°C. The addition of new nuclear deployments in the United States increased the total U.S. mitigation cost savings of the 2°C climate goal by up to $750 billion (NPV). Immediate actions are required in the United States and globally to achieve net-zero carbon emissions by mid-century, and once achieving net-zero emissions, they must remain at net-zero indefinitely. Lifetime extensions of the existing nuclear fleet, in the United States and globally, support urgent near-term emissions reduction goals. Additionally, the longevity of nuclear power technologies reduces the need for new capacity additions of all carbon-free electricity sources and supports long-term actions necessary to maintain net-zero emissions.

Uncertainty assessment for the displacement damage of a pressurized water reactor vessel

The neutron irradiation-induced mechanical property change is a major challenge for nuclear materials. For a Pressurized Water Reactor (PWR), neutron irradiation embrittlement of the reactor pressure vessel is one of the most important issues that determine the whole nuclear power plant lifetime. The present work focuses thus on the uncertainty assessment for the neutron irradiation displacement damage of a PWR vessel. The uncertainties of neutron flux, nuclear data, and atomic displacement models are propagated to Displacement per Atom (DPA) rates with a special emphasis on the importance of correlation matrices. In addition to the uncertainty propagated from the threshold displacement energy, the total uncertainty of damage energy rate is about 10% (at 1 σ ), whereas the uncertainty is largely underestimated if the correlations of damage cross section and neutron flux are not considered. Apart from the complete uncertainty propagation shown in this paper, combining the damage energy rate uncertainty and the displacement threshold energy uncertainty is recommended for estimating the calculated DPA rate uncertainty. Besides the uncertainty, the biases induced by the model defects are preliminarily estimated.

Study of a Gamma Radiation Impact on Concrete Properties Under Severe Accident Conditions

Abstract Concrete is an important structural material used in nuclear power plant (NPP) design. Due to relatively high amount of hydrogen as well as the presence of heavier elements, it also acts as a biological shielding. One of the important tasks for prolongation of operational life time is the determination of concrete components' condition after long-term irradiation. The paper aims to present the current activities in the CV Řež institute (Research Centre Řež—CVR) regarding the investigation of ionizing radiation effects on concrete properties. In its first part, the paper deals with experimental identification of the character of mixed neutron and gamma spectra in the concrete part of the VVER-1000 Mock-Up. Using the knowledge, the radiation field character can be scaled up to the commercial power plants with VVER-1000 light water reactor. It also provides justification for usage of the 60Co source for performed irradiation experiments with concrete. The second part of the article describes the experimental studies of the properties of gamma-irradiated concrete samples by strong 60Co source. This irradiation experiment can be understood as the first step in characterizing concrete degradation as gamma flux in biological shielding is significantly higher than that of neutron flux. In order to better understand the concrete properties and the behavior under irradiation, nondestructive as well as destructive testing methods were applied. We found that after 48 days of irradiation by the 60Co source the sample obtained dose from gamma corresponding to approximately 1% of the total during the NPP lifetime operation. Concrete microstructure degraded and the modulus of elasticity slightly decreased within 5%. Conversely, destruction tests prove significant flexural strength decrease by 27% in case of normal test and by 63% at the loss of coolant accident (LOCA) test.

Microstructure, mechanical properties and IASCC susceptibility of stainless steel baffle bolts after 30 years of operation in a PWR

Neutron irradiation in the core of a nuclear power plant can change the material properties of stainless steel significantly. In this paper cold worked AISI 316 baffle bolts retrieved from Doel 1 PWR were investigated with the aim to assess the structural degradation due to neutron irradiation by analyzing microstructure, mechanical behavior and stress corrosion susceptibility. The baffle bolts had received neutron doses ranging from 6 to 59 dpa covering a major part of a nuclear power plant's lifetime. First, microstructural investigations were carried out to reveal the defect and dislocation density and the amount of void swelling. Then mechanical properties were determined using tensile tests and fracture toughness tests. The irradiation assisted stress corrosion cracking (IASCC) susceptibility was investigated by slow strain rate testing and crack growth rate testing. Results were compared with literature values and found to be in good agreement. Main conclusions were that void swelling was small even for the highest dpa values and that the fracture toughness values saturated above 15 dpa. The material was susceptible to IASCC as shown by SSRT and crack growth rate measurements exhibiting intergranular fracture morphology.

Factors limiting lifetime of nuclear power plants with pressurized-water reactors

This paper builds on papers NED 9185 and NED 9377. Most of the referred supporting information comes from author‘s doctoral thesis (Křivánek, 2018) Factors Limiting Life Time of Nuclear Power Plants with Pressurized-water Reactors published in 2018. The paper provides an overview of operation plans in countries operating PWR reactors, discusses the most important factors limiting lifetime of NPPs leading to permanent shut-down of units and discusses in detail technical factors which can potentially limit lifetime of PWR reactors in future. Conclusions of this article show increasing importance of ageing management and technological obsolescence management which will play a key role for operation of PWR reactors for 60 years but mainly for operation beyond 60 years.

Evaluation of neutron radiation damage in zircaloy fuel clad of nuclear power plants: a study based on PKA and dpa calculations

During the lifetime of nuclear power plants, irradiation and mechanical interactions can cause permanent damage to their structural materials. In nuclear reactors, one of the most important of these materials is fuel clad that, in addition to its main role in transferring heat from fuel to coolant, restrains most of the radioactive fission products within its volume. In this article, neutron-induced radiation damage based on Primary Knock-on Atom (PKA) and displacement per atom (dpa) calculations are evaluated on fuel clad in Hot Fuel Assembly (HFA) in a WWER-1000 reactor core as a case study. New couplings of existing nuclear and radiation damage codes and methods are developed for dpa calculations and results are compared together and against experimental results. In the first step, the full reactor core is simulated with MCNPX code (v2.7) to determine the HFA and its relevant neutron-energy spectrum. SPECTER, SPECTRA-PKA and PTRAC card of MCNPX are then employed to evaluate the PKA spectrum under neutron bombardment. Finally, dpa calculated with the Binary Collision Approximation approach is achieved through the use of SRIM-2013 code. A MATLAB interchange program is developed to prepare and transfer the data between the mentioned codes and SRIM. Comprehensive comparisons are performed between results in terms of dpa obtained by (i) NRT calculation (NRT-dpa) (SPECTER and SPECTRA-PKA results), (ii) arc-dpa calculation (MCNPX+arc-dpa), (iii) BCA calculation (PTRAC+SRIM) and (iv) coupled calculation (SPECTRA-PKA+SRIM) (all generations of knock-on atoms and clusters are tracked in the last two methods). These comparisons reveal the portions of damage in Zircaloy due to PKAs, secondary knock-on atoms and cascade atoms, the effect of position-depended PKA and the fraction of displaced atoms that come back to their lattice positions as a result of thermal effects and annealing. Effects of elemental PKA are obtained by means of SPECTRA-PKA code and a visualization map of radiation damage in the clad is presented as an outcome of the SRIM code simulation. Results confirm that value of dpa calculated by MCNPX+arc-dpa is closer to experimental ones as the recombination effect in radiation damage is considered in this method.

Enhanced prioritisation of prospective scenarios for power generation in Spain: How and which one?

Many countries are implementing energy policies that focus on the transformation of national power generation mixes into highly efficient and sustainable systems. In the case of the Spanish electricity sector, regulations applicable to existing coal thermal plants and the proximity to the end of the lifetime of nuclear power plants arise as critical concerns to be addressed. In this respect, a sensible energy strategy is required in order to fill the future gap of coal and nuclear power plants. Within this context, this work applies a novel methodological framework for the prioritisation of 15 energy scenarios formulated for power generation in Spain. In addition to a business-as-usual scenario, alternative ones are built based on potential restrictions on fossil-based technologies and lifetime extensions of coal and nuclear power plants. The novel multi-criteria decision analysis approach for enhanced, sustainability-oriented energy planning is based on the combined use of energy systems modelling, life cycle assessment, and data envelopment analysis. Besides conventional techno-economic results, the evolution of life-cycle indicators endogenously integrated into the energy system model of Spain is provided. The subsequent use of a dynamic data envelopment analysis model leads to identify and rank the most suitable energy scenarios according to sustainability criteria.

Geometrical study of ventilation system openings of pump room in nuclear power plant

Fire safety is important throughout the lifetime of nuclear power plant (NPP), from design to construction throughout plant operation. The integrity of the shutdown cables located in the pump room in NPPs is very important in the fire protection system. Therefore, the present numerically study investigates the effect of outlet vent opening geometry and the location of the ventilation system specifications on the heat transfer in pump room during a fire. Three-dimensional ANSYS- FLUENT Computational Fluid Dynamics with Realizable k–ε turbulence model is employed in the present work. The results found that the geometry (aspect ratio) of the outlet opening vent has major effects on the shutdown cable protection from fire according to their locations. Furthermore, as the aspect ratio for outlet vent decreases, then the suction of hot gasses from the upper half of the room is improved. Consequently, the temperature of the heat source and target decreases. When the two inlet and outlet vents are located at a same lower level from the ground the inlet velocity should be 1.25 m/s, while, when the two vents are located at two different level the inlet velocity should be 0.75 m/s in order to protect the shutdown cable insulation.

METAL WASTE MANAGEMENT AND RECYCLING METHODS IN THE NUCLEAR POWER PLANT DECOMMISSIONING AND DISMANTLING PROCESS

Decommissioning and dismantlement of nuclear power plants (NPP) present the final phase of a nuclear power plant lifetime and are an essential part of an efective nuclear safety culture. While some countries are implementing nuclear exit strategies and others are building new nuclear power plants or prolonging their lifetimes, the decommissioning of facilities is imminent. This paper gives a short overview on decommissioning and dismantling strategies and techniques as an introduction to the applicability of radioactive metal waste recycling and reuse. General estimations suggest that the majority of the waste mass resulting from decommissioning and dismantling is radiologically unrestricted material i.e. exempt waste. The remaining one-third is the radioactive material with different levels of radioactivity. To present radioactive metal scrap recycling in practice, this paper addresses estimations on metal waste amounts resulting from decommissioning, the use of melting as a possible recycling method (with regard to secondary waste generated) and aspects of environmental and health impacts. Additionally, the best practice example of the decommissioning process in a pressurized water reactor (PWR) with NPP technology is given to provide context for the possible future decommissioning of NPP Krsko. After consideration of the available data on decommissioning waste management, the conclusion is that while the decision on decommissioning strategies and technologies is not straightforward but unique to each country, current experiences give important context to the complexity and potential of decommissioning and provide a basis for future actions.

Effects of 3.5 MeV Proton Irradiation on the Surface of Iron

Radiation damage and its effects on Reactor Pressure Vessel (RPV) steel, an Iron (Fe) based alloy, have been important issues to assess the lifetime of nuclear power plants since nuclear reactor operation changed from an annual cycle to a high burn-up and extended cycle mode in the late 1990s. In this study, using an MC-50 accelerator, 99.99 % iron was irradiated with 3.5 MeV protons with doses in a range from 1x1010 to 1x1012 protons/cm2. First, the proton ion range was estimated at 45 μm using SRIM simulation; Optical Microscope (OM) observation results were in good agreement with the simulation. X-Ray Diffraction (XRD) analysis showed that the diffraction peak on the (110) plane, which is the most compact plane of the BCC structure, was intensified, probably due to the stress concentration stemming from the implanted proton atoms on the plane. Moreover, SEM results revealed that sputtered and recoiled iron atoms from the surface were deposited on the surface and that cracks developed and propagated at the sputtered spot.

Systems Thinking Safety Analysis: Nuclear Security Assessment of Physical Protection System in Nuclear Power Plants

The dynamical assessment has been performed in the aspect of the nuclear power plants (NPPs) security. The physical protection system (PPS) is constructed by the cyber security evaluation tool (CSET) for the nuclear security assessment. The systems thinking algorithm is used for the quantifications by the Vensim software package. There is a period of 60 years which is the life time of NPPs' operation. The maximum possibility happens as 3.59 in the 30th year. The minimum value is done as 1.26 in the 55th year. The difference is about 2.85 times. The results of the case with time delay have shown that the maximum possibility of terror or sabotage incident happens as 447.42 in the 58th year and the minimum value happens as 89.77 in the 51st year. The difference is about 4.98 times. Hence, if the sabotage happens, the worst case is that the intruder can attack the target of the nuclear material in about one and a half hours. The general NPPs are modeled in the study and controlled by the systematic procedures.

The Power Distribution and Neutron Fluence Measurements and Calculations in the VVER-1000 Mock-Up on the LR-0 Research Reactor

Abstract The power density distribution in a reactor has significant influence on core structures and pressure vessel mechanical resistance, as well as on the physical characteristics of nuclear fuel. This quantity also has an effect on the leakage neutron and photon field. This issue has become of increasing importance, as it touches on actual questions of the VVER nuclear power plant life time extension. This paper shows the comparison of calculated and experimentally determined pin by pin power distributions. The calculation has been performed with deterministic and Monte Carlo approaches. This quantity is accompanied by the neutron and photon flux density calculation and measurements at different points of the light water zero-power (LR-0) research reactor mock-up core, reactor built-in component (core barrel), and reactor pressure vessel and model. The effect of the different data libraries used for calculation is discussed.

Uncertainty and sensitivity analyses for age-dependent unavailability model integrating test and maintenance

Abstract The interest in operational lifetime extension of the existing nuclear power plants is growing. Consequently, plants life management programs, considering safety components ageing, are being developed and employed. Ageing represents a gradual degradation of the physical properties and functional performance of different components consequently implying their reduced availability. Analyses, which are being made in the direction of nuclear power plants lifetime extension are based upon components ageing management programs. On the other side, the large uncertainties of the ageing parameters as well as the uncertainties associated with most of the reliability data collections are widely acknowledged. This paper addresses the uncertainty and sensitivity analyses conducted utilizing a previously developed age-dependent unavailability model, integrating effects of test and maintenance activities, for a selected stand-by safety system in a nuclear power plant. The most important problem is the lack of data concerning the effects of ageing as well as the relatively high uncertainty associated to these data, which would correspond to more detailed modelling of ageing. A standard Monte Carlo simulation was coded for the purpose of this paper and utilized in the process of assessment of the component ageing parameters uncertainty propagation on system level. The obtained results from the uncertainty analysis indicate the extent to which the uncertainty of the selected component ageing data set influences the performed unavailability calculations on system level, as well as they present sensitivity insights on the equipment. Sensitivity analyses were additionally conducted. The obtained results indicate sensitivity insights associated to the coded Monte Carlo simulation itself as well as component ageing effects sensitivity judgements related to the selected system unavailability calculation.

Plant life management and modernisation: Research challenges in the EU

The European network of excellence NULIFE (nuclear plant life prediction) has been launched with a clear focus on integrating safety-oriented research on materials, structures and systems and exploiting the results of this integration through the production of harmonised lifetime assessment methods. NULIFE will help provide a better common understanding of the factors affecting the lifetime of nuclear power plants which, together with associated management methods, will help facilitate safe and economic long-term operation of existing nuclear power plants. In addition, NULIFE will help in the development of design criteria for future generations of nuclear power plant. NULIFE was kicked-off in October 2006 and will work over a 5-year period to create a single organization structure, capable of providing harmonised research and development (R&D) at European level to the nuclear power industry and the related safety authorities. Led by VTT (Technical Research Centre of Finland), the project has a total budget in excess of 8 million euros, with over 40 partners drawn from leading research institutions, technical support organizations, electric power utilities and manufacturers throughout Europe. NULIFE also involves many industrial organizations and, in addition to their R&D contributions, these take part in a dedicated End User Group. Over the last 15 years the European Commission has sponsored a significant number of R&D projects under the Euratom Framework Programme and its Joint Research Centre has developed co-operative European Networks for mutual benefits on specific topics related to plant life management. However, their overall impact has been reduced due to fragmentation. These networks are considered forerunners to NULIFE. The importance of the long-term operation of the plants has been recognized at European level, in the strategic research agenda of SNETP (Sustainable Nuclear Energy Technology Platform). In NULIFE, the joint EU-wide coordinated research strategy for plant life integrity management and long-term operation has been defined. Mapping exercise of expertises performed under NULIFE confirmed that NULIFE R&D resources are versatile and high quality. In addition to the wide range of technical expertise available, these are widely spread at geographical and organizational level. Four expert groups, with identified members and links to national programmes, have now produced state-of-the-art type reports related to their expertises. Stress corrosion cracking and thermal fatigue pilot projects have finished concluding reports. Several project proposals have been introduced and optimised for new NULIFE pilot projects or other R&D projects. Based on NULIFE business plan, the discussion of long-term business plan, operational model and statute of the future NULIFE institute has been started. NULIFE maintains the sustainability of nuclear power by focusing on the continued, 60+ years of safe operation of nuclear power plants.

Aging and Life Management System of Reactor Pressure Vessel

Reactor pressure vessel (RPV), the only key component that can not be replaced in nuclear power plants (NPPs), is the main barrier against the radioactive leakage. The lifetime of NPPs is dependent heavily on the life of RPV, and thus, the aging and life research on a RPV is a key factor in determining the life extension of NPPs. The purpose of this paper is to introduce an aging and life management system for an operating RPV which can be used as a reference of the lifetime extension. In order to realize the objective, an aging and life management system was developed. It is an comprehensive knowledge management system that integrates decentralized information and serves as a valuable data center. Based on the storage and management of RPV state information and operation data, this system provides real-time monitoring of important operating parameters, evaluation of irradiation embrittlement, and RPV aging assessment. Therefore, it is anticipated that the developed system can be used as an efficient tool for aging and life estimation of RPV.