Plant Life Extension Research Articles

Residual Creep Life Assessment of High-Temperature Components in Power Industry.

A large percentage of power, petroleum, and chemical plants over the world were in operation for a long duration with the corresponding critical components being used beyond the design life of 30 to 40 years. It is generally more cost-effective to refurbish or modernize the degraded equipment or components, rather than to construct a new plant. Therefore, a reliable plant life extension assessment that can evaluate the critical components is needed. The key element in plant life extension is the residual life assessment technology. However, at present, there is still no general consensus among the industry players on the approach to adopt when performing residual life assessment for such a critical damage mechanism as creep. In this article, a three-level residual life assessment methodology is proposed as a general approach for high-temperature components prone to creep. A detailed validation of the selected guidelines and calculation models is also described. Eventually, an application of the three-level methodology to a real industrial case study is presented.

Methodology for lifetime characteristics assessment of mechanical equipment in nuclear power plants. Ageing management of NPP mechanical equipment

In the 20s of the twenty-first century many nuclear power plants around the world have reached the final years of their design lifetime. The technical aspects of Plant Life Extension (PLEX) activities involve tests, in addition to the routinely performed ones, of the materials and structures of reactor units, as well as the carrying out operability analyses. In relation to this, the absence of regulatory and methodological basis becomes apparent, in Bulgaria as well. In recent years, new approaches have been adopted for evaluation of lifetime characteristics of components. The algorithm implemented is based on the active mechanism of degradation of the mechanical properties of materials. This paper discusses PLEX stages and corresponding activities. A methodology has been developed to assess the lifetime characteristics of Nuclear Power Plant (NPP) components such as to suit the stages of plant life extension and long-term operation. The methodology consists of: 1) Equipment classification; 2) Identification of the major degradation mechanisms of the mechanical properties of materials; 3) Determining the degradation effects of the mechanical properties of materials; 4) Identification of methods to control degradation effects (ageing); 5) Development of methodologies for evaluation of the ageing effects.

Investigation of nuclear concrete

Recent times, life extension of nuclear power plant (NPP) plays one of the crucial roles in energetics and especially in nuclear industry. Key members of NPP's long term operation are reactor pressure vessel (RPV) and its cavity (concrete shell surrounding nuclear reactor). There is big interest in prediction of the state of biological shielding concrete around RPV. Many countries all over the world solve more or less the same problem: how to predict its state? Recent article provides with an opportunity of biological shielding concrete non-destructive testing (NDT). Two NDT methods will be performed as a possible solution of the problem. Due to high radiation, mechanical manipulators have been applied in order to attach ultrasonic or acoustic-emission transducers onto concrete surface or steel liner surface. The results will show 1) the possibility of the measurement near nuclear reactor in its outage, 2) the testing of RPV cavity from Greifswald's NPP NORD 3) and also the measurement of concrete samples irradiated by neutron and gamma radiation in light water reactor LVR-15. All presented material leads to the same goal: prolong the durability of old nuclear power plants and/or access their reliability and/or guarantee their safety.

Cable Aging Management Program Implementation in Krško NPP-NEK

As a requirement for plant life extension for more than 40 years some additional Aging Management Programs (AMP) for passive equipment have to be implemented. The article presents overview of Cable Aging Management Program (CAMP) implementation. Program defines basic rules and initial activities for identification of adverse operation environment parameters that could lead to accelerated aging of specific materials. Samples of cables are selected based on nuclear safety and electrical equipment criticality for inspection and testing, to check functionality and prevent unexpected failure during normal operation. Acceptance criteria for environment parameters and diagnostic testing have been set. Initial visual inspection of cables condition in adverse environment and testing of sampled cables and environment are giving results for in time preventive measures. First cable aging management program is implemented since 2010 and its experience could be accommodated to other companies with cables recognized as key components.

FORMULA FOR THE DURABILITY OF OIL AND GAS FIELDS

The article is concerned with the issues of the status and results of long-term development of oil and gas fields, with consideration and justification of the factors that influence the extension of their further cost-effective development. To evaluate the potential factors of “plant life extension” of hydrocarbon reservoirs, the Munaily field in above-salt deposits of the southeastern part of the PreCaspian Depression has been designated as a pilot option. Pursuant thereto, the sequence of the major stages in the study and development, as well as the specific features of this field are provided. The forecast made by the authors for the longevity of field production is based on the factor of natural replenishment of reserves, which can be achieved through the establishment of certain restrictive measures and observance of the optimum operation mode appropriate for the conditions of each particular field. This refers to the setting of optimal oil withdrawals, the average daily flow rate of the well, etc. Determination of this pattern and its use in practical terms has favorable historical prerequisites, which is confirmed by the example of numerous objects under development in Russia and Kazakhstan. When substantiating this forecast, the main features of the geological structure of the Munaily field have also been taken into account, among which considerable importance in evaluating the mode of occurrence and productivity of above-salt deposits in the Pre-Caspian Depression is given to the presence and impact of underlying complexes of sediments. This concerns the salt-bearing section of the Kungurian age, the subsalt Paleozoic strata, and the influence of fault tectonics, which contributes to the development in the field section of certain oil and gas feed channels due to disturbances of various order, zones of decompaction, and fracture porosity of the rock formations. Following the results of the analysis, the necessity of compliance with the rehabilitation cycle of 10-15 years has been substantiated, which allows restoring the internal indicators of the energy state and the volume of hydrocarbons in line with the existing model of the reservoir (trap) of the field to the level at which the continuation of further cost-effective development is possible.

Arnstein’s Ladder of Citizen Participation in the Life Extension and Decommissioning Process of the Kori-1 Nuclear Power Plant

문재인 정부에서 추진되었던 월성 1호기의 폐쇄는 정치적 쟁점으로 부각되며, 사회적 혼란을 불러일으키고 말았다. 심지어는 대통령 선거를 앞둔 시점에서까지 대선 후보들이 탈원전 정책의 철회를 공약으로 내세우면서, 월성 원전 관련 갈등은 차기 정부에서도 해소되지 않을 것으로 예상될 정도이다. 반면에 박근혜 정부에서 진행되었던 고리 1호기의 폐로 과정은 적어도 국가적인 차원에서는 그다지 심각한 논란 없이 원만하게 결정이 내려질 수 있었다. 이에 본 논문은 고리 1호기의 수명 연장 및 폐로 관련 결정 과정을 “시민참여”라는 관점에서 분석하고 있다. 구체적으로는 이러한 연구 목적을 달성하기 위해 아른슈타인의 참여 사다리 모형이 분석틀로 활용되었다. 연구 결과 “수명 연장” 과정에서는 조작 및 처방이라는 ‘참여 결핍’ 수준의 시민참여만이 가능했던 것으로 확인되었다. 반면에 “폐로 결정” 과정에서는 후쿠시마 사고, 정전 사태, 부품 성적서 조작 등의 복합적인 요인들로 인해 고지･협의･회유 단계의 ‘명목 참여’까지 달성되었던 것으로 밝혀졌다. 다만 중앙정부가 결정권을 보유하고 있는 원자력 정책의 특성상 ‘시민 권력’ 수준의 시민참여는 이뤄지지 않았었다. 결론적으로 본 논문에서는 현재 시점에서 잠재적 갈등 요인으로 남아있는 고리 2호기의 수명 연장과 관련해서 시민참여를 강화해야 한다는 단기 및 장기의 정책적 함의가 제시될 수 있었다.

Thermal hydraulic and mechanical behaviour of VVER-1000 reactor lower plenum in the late phase of severe accident

One of the most critical issues for determining the life extension of a nuclear power plant is the structural integrity analysis of a reactor pressure vessel (RPV). This work, examines the specifics of thermal-mechanical analysis, for the lower plenum of the reactor pressure vessel of VVER-1000, during the late in-vessel phase of a severe accident. Thermal-hydraulics and structural models of ANSYS FLUENT17.2 are prepared to obtain both the radial, axial temperatures and the stress field across the vessel wall thickness. Boundary conditions obtained from the ASTEC code are applied as input to the ANSYS FLUENT simulation for corium behavior in lower plenum: 2D meshing of vessel lower plenum, two corium layers (oxide layer and metal layer).The variation of stress, strain and damage with time at the critical axial layer are investigated. The analysis identifies the worst crack orientation and location. The results show without external flooding of the vessel, a pure thermal failure (melt-through) is highly probable due to the high thermal loads imposed by the decay power. Applied external flooding of the vessel would help the stability of the vessel and the retention of the molten corium inside the reactor pressure vessel and can prevent the vessel failure.

Flaw Evaluation Procedure for Cast Austenitic Stainless-Steel Materials Using a Newly Developed Statistical Thermal Aging Model

Abstract Thermal embrittlement of some cast austenitic stainless steels (CASS) occurs at reactor operating temperatures leading to very low fracture toughness. Because of their low aged toughness with high variability, flaw evaluations of CASS material need to be established with an understanding of the materials aged condition, especially since most U.S. pressurized water reactor (PWR) nuclear plants have been given plant life extensions for 60-year operation. A flaw evaluation procedure for CASS materials is presented here using a new statistical model developed to predict the toughness of fully aged CASS using the materials' chemical compositions. In this procedure, the dimensionless-plastic-zone-parameter (DPZP) analysis is used to determine when limit-load is applicable and also approximate the elastic-plastic correction factor (Z-factor) to predict the failure stress for CASS pipe/fittings with a circumferential surface crack. The procedure was validated against several CF8m pipe test results which include various pipe diameters, crack sizes, ferrite contents, failure modes. The as-developed flaw evaluation procedure was also used to determine the Z-factors for four different pipe diameters for a database of 274 pipe/elbows in U.S. PWR plants –solving 1096 sample problems to understand what range of Z-factors in U.S. PWR plants (for CF8m CASS materials). Finally, the applicability of the CF8m-based statistical model for use with CF3 and CF8 CASS materials was also verified with available test results. The outcome of this work has been implemented in an ASME boiler and pressure vessel (BPV) code case and has been approved in 2020 as Code Case N-906.

Plant Life Extension programs - useful resources for preservation and amplification of Nuclear Power Plants efficiency

"Plant Life Management (PLIM) is now usually taken to mean those many activities which are to do with the normal operational management of a nuclear power plant to maintain its 'good condition' and to enable operators or Owner(s) to meet the plant's intended amortisation period, operational life or design operational life. That is, it is to do with existing planned Plant Operational or Service Life assurance and not necessarily to do ith meeting an additional period or extended operation time in excess of that intended at the design or evaluation stage, which is usually known as Plant Life Extension (PLEX). The paper presents considerations regarding the ageing management of the NPP’s key components, Plant Life Management (PLIM), refurbishment as activity strictly necessary for Plant Life Extension, and the Romanian refurbishment program as part of PLIM from Cernavoda CANDU NPP. "

Role of Cavity Formation on Long-Term Stress Corrosion Cracking Initiation: A Review

Plant life extension from the initial licensed life to beyond 60 years is now being discussed for light water cooled nuclear powered reactors (LWRs). Reliable prediction for material degradation is extremely important to keep the reliability of LWRs during such long-term operation. One of the specific perspectives for this prediction is to take into account the changes in material properties during long-term operation, such as cavity formation, even at the LWR operating temperature. The mechanism of cavity formation and the associated phenomena are closely intertwined with interdisciplinary technological and scientific knowledge. Therefore, historical key knowledge from both phenomenological and fundamental research studies related to cavity formation was first reviewed to understand the overall picture. Subsequently, current research results related to long-term stress corrosion cracking initiation in the LWR environment were summarized to explain what is known, what is still unknown, and what are the critical remaining subjects.

Role and importance of NDE in nuclear power plant life extension

Nuclear power plants are ageing but there is a need for their further use. Ageing may lead to the gradual degradation of the safety margins of structures and components. Service life extension of nuclear power plants became a world tendency, the key condition of which is the assurance of components’ structural integrity until the end of the extended lifetime. Non-destructive evaluation is a solid pillar of structural integrity assessment, and in the case of nuclear power plants, the general objectives (social and commercial) of non-destructive evaluation are highly relevant.The paper will show examples on the substantial role and importance of non-destructive evaluation in ensuring nuclear component’s long-term integrity. The examples are selected from the technical justification of service life extension of Paks Nuclear Power Plant, Hungary. They will cover ultrasonic and eddy current examinations of critical areas and sub-components of the reactor pressure vessel, the reactor internals, the steam generator and some pipeline systems. The results contributed to completion of key time-limited ageing analyses which were absolutely necessary for getting the operating license beyond the design life.

Nordic research and development cooperation to strengthen nuclear reactor safety after the Fukushima accident

Abstract NKS (Nordic Nuclear Safety Research) is a forum for Nordic cooperation and competence sharing within nuclear safety and emergency preparedness, serving as an umbrella for Nordic initiatives and interests. An overview of the funded activities within the nuclear safety program (NKS-R) during the last few years is presented in this article. The presented research covers the areas of thermal hydraulics, severe accidents, risk analysis and probabilistic methods, organisational issues and safety culture, decommissioning and plant life management and extension. Activities are focused towards practical and directly applicable scientific results and competence building. NKS-R funds research activities with particular relevance for the development of Nordic reactor safety research, and promotes participation of young scientists in the activities. The activities involve experimental and analytical studies, workshops, case studies and seminars, and bring together technical research organisations, universities, authorities, industries and consultant companies from the Nordic countries. As exemplified in this report, The NKS-R program provides support with high additionality particularly in smaller R&D projects and pilot projects, and offers the opportunity for network building and valuable collaborations between experts at the leading Nordic research facilities within reactor safety.

The Climate and Economic Rationale for Investment in Life Extension of Spanish Nuclear Plants

Spain's seven operating nuclear plants currently provide more than 20% of its electricity. Each of these began operation in the 1980s and is approaching the end of its 40-year design life. Extending their lives will require additional investments. Should Spain make the investment and extend their lives, or should they be retired at the end of their design life? We show that investing in nuclear plant life extensions is the least- cost alternative for further reducing GHG emissions. We also show that in assessing the cost of renewable alternatives it is critical to take into account the time profile of the available renewable resource. Solar PV and especially wind capacity were expanded significantly since 2000, and significantly greater penetration, especially of solar PV, is promised out to 2030 in order to reduce GHG emissions still further. We show that at these expanded penetration levels, curtailment becomes a significant determinant of system cost. This significantly improves the relative value of nuclear life extensions as a contributor to reducing GHG emissions.

Radiation-induced bcc-fcc phase transformation in a Fe[sbnd]3%Ni alloy

The issue of neutron irradiation embrittlement of Reactor Pressure Vessel steels must be considered for Nuclear Power Plant life extension. This phenomenon partly arises from the existing interactions between dislocations and nanometric clusters composed of Cu, P, Si, Mn and Ni. The latter alloying element, playing a key role in the evolution of solute enriched clusters under irradiation, is the focus of this publication. To assess the effect of Ni on microstructure evolution under irradiation, particle accelerator based experiments were conducted. An under-saturated Fe3at.%Ni alloy was irradiated with self-ions, at 673 K, up to ∼1.2 dpa. Then, the microstructural damage was characterized, at the atomic scale, by conventional Transmission Electron Microscopy, Scanning Transmission Electron Microscopy coupled to Energy Dispersive X-ray Spectroscopy and Electron Energy Loss Spectroscopy, while chemical features were investigated by Atom Probe Tomography. Informations obtained by combining these coupled techniques provide evidence for the formation of a FCC phase, containing 25 at.%Ni, which can be either the disordered γ phase or the ordered L12 type Fe3Ni phase. The metastable or stable state of this FCC phase is discussed in the light of what is known from the literature. It is the first time that this BCC-FCC phase transformation is observed in an under-saturated α-FeNi alloy and this likely occurred via a Radiation Induced Precipitation (RIP) mechanism. Ni atom segregation is observed on cavities, dislocation lines and dislocation loops. The latter constitute nuclei for precipitates, leading to the formation of an additional segregation site for Ni: the precipitate FCC - matrix BCC nearly coherent interface. Similar mechanisms are argued to be operating also in high Ni RPV steels under neutron irradiation.

Determining the electrical and thermal resistivities of radiolytically-oxidised nuclear graphite by small sample characterisation

Property measurement of graphite extracted from operational nuclear reactors is vital for forward prediction of the graphite's properties and integrity as part of continued operation and plant life extension assessments. Limited core monitoring material is available, so robust and efficient measurement techniques are required for samples with constrained dimensions. The validation of a technique for measuring electrical resistivity in small graphite samples (largest dimension < 20 mm) is described with an empirical validation of the geometry correction factors. The electrical and thermal characterisation of 175 radiolytically-oxidised Gilsocarbon-grade graphite samples extracted from four advanced gas-cooled reactors (AGRs) is presented, including a subset that were further oxidised in an accelerated-irradiation test programme. The selected material sampled radiolytic-oxidation induced weight losses of up to 59% and doses of up to 193 × 1020 n.cm−2 equivalent DIDO nickel dose, or ca. 28 displacements per atom. For all the measured oxidation states, a linear correlation between thermal diffusivity and density is demonstrated, with power law functions found to describe the trends in thermal and electrical resistivity with density. The effect of the oxidation on the phenomenologically-different heat and electrical transport mechanisms is considered and electrical resistivity, ρE, is shown to be an excellent proxy for thermal resistivity, ρT, with a calibration function ρE ∝ ρT0.66. For low weight loss samples, including virgin Gilsocarbon, a linear approximation is also described.

Ageing management in NPP. Effectiveness of the methods for control, examination and monitoring in relation to mechanisms of degradation of mechanical properties

This paper describes ageing management activities for Long Term Operation (LTO) of components of Kozloduy NPP EAD. The stages of LTO Project are: Stage 1 - implementing a comprehensive assessment of the actual condition of the equipment; Stage 2 - providing of a Complex Analysis; Stage 3 – licensing of Plant Life Extension (PLEX) for long-term operation (LTO). The main activities are assessment of actual condition of the equipment and building; review of methods for control (non-destructive), examination and monitoring; assessment of effectiveness of methods; necessity of additional control/ examination/ service. The paper describes some mechanisms of degradation of mechanical properties, methods for control and criterions for their effectiveness.

Analysis of Power Uprate in Nuclear Power Plants

The three main strategies used by the utilities in order to obtain more benefits from the nuclear installations focused on electricity production and in order to satisfy their demand without carbon dioxide emissions are: (1) Power uprate, (2) plant life management and (3) plant life extension. Alone in the United States, the Nuclear Regulatory Commission has approved more than 140 Nuclear Power Plant (NPP) power uprates since 1977, which represent the equivalent of five new Nuclear Power Plants (NPPs). While in the rest of the world the NPP uprates represent the equivalent of two new NPP. This study is a compilation of the power uprate experience in the whole world and a discussion about important problems detected due to this process in Light Water Reactors (LWR), specifically for Boiling Water Reactors (BWR). The power uprate involves the reanalysis of many topics in order to assure that the safety operational margins are kept. The new radiological consequences, structural integrity of the systems, vibrations, core heat balance and thermal-hydraulics behavior under transient and accident conditions, are some of the aspects to be considered during power uprate processes. As special case the Laguna Verde Power Nuclear Power Plant (LVNPP) Extended Power Uprate (EPU) is presented, whose experience indicated that the steam dryer loads are crucial for uprates in BWR´s.

Analysis for the Effect of Spatial Discretization Method on AP1000 Reactor Pressure Vessel Fluence Calculation

Maintaining the structural integrity of the reactor pressure vessel (RPV) is a critical concern related to the safe operation of nuclear power plants. To estimate the structural integrity over the designed lifetime and to support analyses for a potential plant life extension, an accurate calculation of the fast neutron fluence (E&gt;1.0 MeV orE&gt;0.1 MeV) at the RPV is significant. The discrete ordinates method is one of the main methods to solve such problems. During the calculation process, many factors will affect the results. In this paper, the deviations introduced by different differencing schemes and mesh sizes on the AP1000 RPV fast neutron fluence have been studied, which are based on new discrete ordinates code ARES. The analysis shows that the differencing scheme (diamond difference with or without linear zero fix-up, theta weighted, directional theta weighted, and exponential directional weighted) introduces a deviation within 4%. The coarse mesh (4 × 4 cm meshes inXYplane) leads to approximately 23.7% calculation deviation compared to those of refined mesh (1 × 1 cm meshes inXYplane). Comprehensive study on the deviation introduced by differencing scheme and mesh size has great significance for reasoned evaluation of RPV fast neutron fluence calculation results.

Pipe/duct system design for tornado missile impact loads

For nuclear power plant life extension projects, it may be convenient and in some instances necessary to locate safety-related steel ducts and pipes outside of the main structures, exposing them to extreme environmental loads such as tornado missile impact. Examples of this application include emergency firewater lines and Control Room vent ducts. A typical exposed commodity run could be comprised of a rectangular or circular cross-section with horizontal and vertical segments supported at variable spans off of roof and wall panels, respectively. Efficient and economical design of such a tornado-impacted duct or pipe system, consisting of the commodity and its supports, must exploit all of the system's capability to absorb the impact energy by deforming plastically to the fullest extent allowable. Energy can be absorbed locally in the vicinity of impact on the commodity, globally through rotation at flexural plastic hinges, and through yielding of the supports. In this paper a simplified NDOF lumped parameter nonlinear analysis methodology is presented and applied to the coupled commodity/support system subjected to tornado impulse loading. The analysis methodology is confirmed using a detailed ANSYS nonlinear finite element model. Optimization of the initial trial design is achieved by progressively decreasing the support resistances, while monitoring the response ductilities throughout the system. Evaluation methodologies are provided for the four types of plastic deformation responses which occur in the system: local response in the immediate vicinity of impact, flexural and membrane response of the sidewall out to one or two times the commodity depth beyond the point of impact, global response of the commodity as a beam spanning between supports, and the shear and flexural response of support. The inelastic responses are evaluated against AISC N690 acceptance criteria (ANSI, 2006), supplemented as appropriate by triaxiality considerations for inelastic strain in the pipe material.

Non-Environmentally Assisted Cracking Corrosion Concerns Affecting Life Extension of Light Water Reactors

Corrosion is a major service life-limiting mechanism for both pressurized water reactors (PWR) and boiling water reactors (BWR). While most of the corrosion research emphasis in the nuclear corrosion community has been focused on environmentally assisted cracking (EAC) of austenitic stainless steels and nickel-based alloys and weld metals, in particular stress corrosion cracking (SCC), there are other corrosion phenomena that seriously affect plant life extension that cannot be ignored. This paper presents three other important corrosion areas, i.e., general corrosion of the light water reactor (LWR) containments, flow-accelerated corrosion of carbon steel piping systems, and the corrosion of buried piping.