Light Water Reactor Plants Research Articles

Development of true stress-true strain relation for various structural materials in light water reactor plants

Development of true stress-true strain relation for various structural materials in light water reactor plants

Study of alkaline carbonate cooling to mitigate Ex-Vessel molten corium accidents

To mitigate adverse effects from molten corium following a reactor pressure vessel failure (RPVF), some new reactor designs employ a core catcher and a sacrificial material (SM), such as ceramic or concrete, to stabilize the molten corium and avoid containment breach. Existing reactors cannot easily be modified to include these SMs but could be modified to allow injectable cooling materials. Current reactor designs are limited to using water to stabilize the corium, but this can create other issues such as reaction of water with the concrete forming hydrogen gas. The novel SM proposed here is a granular carbonate mineral that can be used in existing light water reactor plants. The granular carbonate will decompose when exposed to heat, inducing an endothermic reaction to quickly solidify the corium in place and producing a mineral oxide and carbon dioxide. Corium spreading is a complex process strongly influenced by coupled chemical reactions, including decay heat from the corium, phase change, and reactions between the concrete containment and available water. A recently completed Sandia National Laboratories laboratory directed research and development (LDRD) project focused on two research areas: experiments to demonstrate the feasibility of the novel SM concept, and modeling activities to determine the potential applications of the concept to actual nuclear plants. Small-scale experiments using lead oxide (PbO) as a surrogate for molten corium demonstrate that the reaction of the SM with molten PbO results in a fast solidification of the melt due to the endothermic carbonate decomposition reaction and the formation of open pore structures in the solidified PbO from CO2 released during the decomposition. A simplified carbonate decomposition model was developed to predict thermal decomposition of carbonate mineral in contact with corium. This model was incorporated into MELCOR, a severe accident nuclear reactor code. A full-plant MELCOR simulation suggests that by the introduction of SM to the reactor cavity prior to RPVF ex-vessel accident progression, e.g., core-concrete interaction and core spreading on the containment floor, could be delayed by at least 15 h; this may be enough for additional accident management to be implemented to alleviate the situation.

Finite Element Analysis of Internally Circumferential Surface Cracked Pipe Stainless Steel under Bending and Torsion Loadings

Early Light Water Reactor Plants (LWR plants) in Japan are nearly 30 years old and utilities have been conducting technical evaluations of facility aging. Accurate prediction of fracture behavior for ductile materials is one of the most important issues to guarantee the integrity of the structure such as pipes in plants. If a crack is detected during the in-service inspections, the structural rational maintenance of the cracked pipe has been promoted based on the defect evaluation methods specified ASME Boiler and Pressure Vessel Code and JSME Rules on Fitness-for-Service for Nuclear Power Plants. Recently, experiments and simulations have been performed on pipes subjected to bending and torsion loadings. However, the failure criterion is still under investigation, and it is significant to propose a method to evaluate it appropriately. In this study, a finite element model of a large circumferentially cracked stainless steel pipe was generated and elastic-plastic analysis of stationary crack and crack extension were performed to validity of the results.

Nuclear heat storage and recovery for the APR1400

Detailed here is the proposed design of the tertiary side of a Thermal Energy Storage (TES) System to be interfaced with the steam cycle of a Light Water Reactor (LWR) plant. This study extends the understanding provided by many previous investigations (i.e., beyond thermodynamic considerations) by specifically addressing real world constraints associated with the backfit of such a system to an operating LWR. These constraints relate to the feasibility of design modifications during an extended refueling outage, and to licensing, operational, and maintenance considerations. The Korean designed and built APR1400 plant is selected for integration with the TES tertiary system, resulting in a design which can be readily be adapted to most LWR plant. Heat transfer and transport from and to the nuclear steam cycle is by synthetic oil with high temperature capabilities. Heat storage is in the form of packed beds consisting of crushed rock. While the rates of storage and recovery are constrained by the design of the nuclear reactor and steam plant (i.e., 20% of reactor thermal power during storage, and ~11% during recovery) the total stored energy component of the system can be readily and economically scaled to any desired capacity (at the marginal cost of carbon steel vessels filled with rock). Finally, it is proposed that any nuclear installation with water access could employ bulk thermal energy storage built into an Ultra Large Barge for turnkey delivery with a concomitant reduction in installed cost and risk.

Leak-before-break analyses of PWR and BWR piping concerning size effects

Components of primary circuit piping of light water reactor (LWR) plants fulfil high safety requirements. Beside prevention against any damage, cracking, and leaking, pipes are especially designed to exclude double-ended ruptures, which may cause a high loss of coolant. One option in this strategy, supported by regulations and standards in several countries, is the leak-before-break (LBB) behavior: a crack-like leakage should be detected before it becomes a break.Evidently, many factors influence the LBB behavior of a specific piping section. Material properties, plant operation parameters, assumed accident scenarios and the local geometry determine the result of a LBB analysis as well as the associated safety margins. Therefore, an analysis is traditionally done case by case, and site by site, according to the applicable standard.In this article, typical components of pressure water reactor (PWR) and boiling water reactor (BWR) primary circuits are examined, and their LBB behavior is investigated. The German standard (KTA) and the US approach (U.S.NRC) are compared. The quantitative results are obtained with the GRS codes PROST and WinLeck, which include different methods for fracture assessment and leak rate calculation. A clear trend concerning the pipe size is found, resulting in large safety margins for large-diameter piping.

United States Department of Energy Severe Accident Research Following the Fukushima Daiichi Accidents

The U.S. Department of Energy (DOE) has played a major role in the U.S. response to the events at Fukushima Daiichi. During the first several weeks following the accident, U.S. assistance efforts were guided by results from a significant and diverse set of analyses. In the months that followed, a coordinated analysis activity aimed at gaining a more thorough understanding of the accident sequence was completed using laboratory-developed, system-level best-estimate accident analysis codes, while a parallel analysis was conducted by U.S. industry. A comparison of predictions for Unit 1 from these two studies indicated significant differences between MAAP and MELCOR results for key plant parameters, such as in-core hydrogen production. On that basis, a crosswalk was completed to determine the key modeling variations that led to these differences. In parallel with these activities, it became clear that there was a need to perform a technology gap evaluation on accident-tolerant components and severe accident analysis methodologies with the goal of identifying any data and/or knowledge gaps that may exist given the current state of light water reactor (LWR) severe accident research and augmented by insights from Fukushima. In addition, there is growing international recognition that data from Fukushima could significantly reduce uncertainties related to severe accident progression, particularly for boiling water reactors. On these bases, a group of U.S. experts in LWR safety and plant operations was convened by the DOE Office of Nuclear Energy (DOE-NE) to complete technology gap analysis and Fukushima forensics data needs identification activities. The results from these activities were used as the basis for refining DOE-NE's severe accident research and development (R&D) plan. This paper provides a high-level review of DOE-sponsored R&D efforts in these areas, including planned activities on accident-tolerant components and accident analysis methods.

Environmental fatigue analysis of a U seal in pressure vessel under simulated LWR operation

U seal components are widely used in light water reactor (LWR) plant for the purpose of sealing and connecting vessels and piping. Due to remarkable environmental effects, the environmental fatigue analysis is of crucial important for assessing the life of U seal in LWR applications. The environmental effects are incorporated by multiplying the cumulative fatigue usages (CUF) with an environmental factor (Fen), which is a complex function of many variables such as dissolved oxygen (DO) content, strain rate, temperature. The Fen-factor is found to be greater than 2 at the root of lower U seal branch, which indicates that environmental effects can significantly reduce the fatigue life of the components. Since the fatigue is related and initiated by high local stresses and temperature transient, an elastic–plastic finite element (FE) calculations are carried out. The environmentally assisted fatigue life assessments were performed by implementation of Fen-factor in FE calculation. The results of the case studies suggested that there were a sharp increase in calculated fatigue usage with Fen-factor, and the monitoring data were in consistent with the predicted environmental fatigue usage.

Irradiation effects and micro-structural changes in large grain uranium dioxide fuel investigated by micro-beam X-ray diffraction

Microstructural changes in a set of commercial grade UO2 fuel samples have been investigated using synchrotron based micro-focused X-ray fluorescence (μ-XRF) and X-ray diffraction (μ-XRD) techniques. The results are associated with conventional UO2 materials and relatively larger grain chromia-doped UO2 fuels, irradiated in a commercial light water reactor plant (average burn-up: 40MWdkg−1). The lattice parameters of UO2 in fresh and irradiated specimens have been measured and compared with theoretical predictions. In the pristine state, the doped fuel has a somewhat smaller lattice parameter than the standard UO2 as a result of chromia doping. Increase in micro-strain and lattice parameter in irradiated materials is highlighted. All irradiated samples behave in a similar manner with UO2 lattice expansion occurring upon irradiation, where any Cr induced effect seems insignificant and accumulated lattice defects prevail. Elastic strain energy densities in the irradiated fuels are also evaluated based on the UO2 crystal lattice strain and non-uniform strain. The μ-XRD patterns further allow the evaluation of the crystalline domain size and sub-grain formation at different locations of the irradiated UO2 pellets.

Review of Severe Accident Phenomena in LWR and Related Severe Accident Analysis Codes

Firstly, importance of severe accident provision is highlighted in view of Fukushima Daiichi accident. Then, extensive review of the past researches on severe accident phenomena in LWR is presented within this study. Various complexes, physicochemical and radiological phenomena take place during various stages of the severe accidents of Light Water Reactor (LWR) plants. The review deals with progression of the severe accidents phenomena by dividing into core degradation phenomena in reactor vessel and post core melt phenomena in the containment. The development of various computer codes to analyze these severe accidents phenomena is also summarized in the review. Lastly, the need of international activity is stressed to assemble various severe accidents related knowledge systematically from research organs and compile them on the open knowledge base via the internet to be available worldwide.

TRU burning by dual tier system of LWR-SFR

This paper presents a concept of the dual tier system consisting of the existing light water reactor (LWR) plants and sodium-cooled fast reactor (SFR) for transuranics (TRU) burning for the purpose of downsizing the required SFR. In this system, Pu is combusted by the LWR at first and then the remaining Np, Am, and Pu are destructed by the SFR. The iteration number of Pu combustion by the LWR is chosen to be twice owing to the sodium void reactivity limitation of $6. As a result of combustion calculation, the twice Pu burning of LWR lessens the TRU amount by 27% and changes the composition significantly. Moderator pins of zirconium hydride are deployed to the SFR fuel subassembly so as to enhance TRU burning and reduce the sodium void reactivity. The nuclear calculation found that the core characteristics become similar to the conventional SFR due to the moderator: the sodium void reactivity remains still $4 and the Doppler coefficient becomes −6 × 10 −3 Tdk/dT. This study concludes that this dual tier strategy can downsize the required SFR to approximately 40% of the single tier system of SFR with TRU conversion ratio of 0.6.

Seismic Isolation for Advanced Fast Reactors

We present the results of a survey of the state of seismic isolation technology. The emphasis of the review is placed in the United States. The purpose of this survey was to provide an engineering basis for the use of seismic isolation in the design of nuclear power plants. In particular, the survey is focused on providing a basis for the design of advanced fast reactor (AFR) nuclear power plants. These AFR plants typically have components and piping that are thin walled as opposed to the thick-walled components and piping in light water reactor (LWR) plants. As a result the AFR plants do not have the adequate inherent strength to resist seismic loads that exists in the LWR plants. It is far more desirable, therefore, to reduce the seismic demand on the AFR plants than to require costly measures to strengthen the structures and components. It is believed that the use of seismic isolation is a viable and effective way to provide this reduction in seismic demand. Various types of seismic isolation systems and devices are reviewed along with their strengths and weaknesses. Descriptions of several U.S. seismically isolated buildings are presented. The results of actual performance of seismically isolated buildings are also presented, including representative measurements of accelerations in the structures when subjected to actual seismic events. It is concluded that the seismic isolation technology is well established and that the path forward leading to the use of this technology for AFR nuclear power plants is clear and achievable.

Evaluation Methods for Corrosion Damage of Components in Cooling Systems of Nuclear Power Plants by Coupling Analysis of Corrosion and Flow Dynamics (V) Flow-Accelerated Corrosion under Single- and Two-phase Flow Conditions

The characteristics of flow-accelerated corrosion (FAC) in the wide sense are overviewed. Overlapping effects of flow dynamics and corrosion are important issues to determine the reliability and lifetime of major structures and components of light water reactor plants. FAC in single- and two-phase flows and liquid droplet impingement (LDI) in two-phase flow are typical phenomena due to both interactions. In order to evaluate local wall thinning due to FAC and LDI, a 6-step evaluation procedure for each has been proposed. (1) For FAC wall thinning evaluation, corrosive conditions along the flow path and mass transfer coefficients at the structure surface under single- and two-phase flow conditions were calculated to evaluate high-risk zones for FAC occurrence, and then wall thinning rates were calculated with the coupled model of static electrochemical analysis and dynamic oxide layer growth analysis at the identified high-FAC-risk zones. (2) For LDI local wall thinning evaluation, trajectories of liquid droplets in high-velocity steam, their collision densities and velocities on the pipe inner surface were calculated to evaluate high-risk zones for LDI occurrence, and then local wall thinning rates were calculated at the identified high-LDI-risk zones. For the region with steam velocity higher than 200 m/s, the wall thinning rate was determined mainly by erosion (LDI (erosion)), while for that with velocity lower than 100 m/s, it was determined by corrosion (LDI (corrosion)). (3) The FAC wall thinning model was applied for two-phase flow as well as single-phase flow, while the FAC wall thinning model was applied for LDI (corrosion). The empirical formula was applied for LDI (erosion) after discriminating LDI (corrosion) and LDI (erosion) based on steam velocity and quality.

Evaluation Methods for Corrosion Damage of Components in Cooling Systems of Nuclear Power Plants by Coupling Analysis of Corrosion and Flow Dynamics (V) Flow-Accelerated Corrosion under Single- and Two-phase Flow Conditions

The characteristics of flow-accelerated corrosion (FAC) in the wide sense are overviewed. Overlapping effects of flow dynamics and corrosion are important issues to determine the reliability and lifetime of major structures and components of light water reactor plants. FAC in single- and two-phase flows and liquid droplet impingement (LDI) in two-phase flow are typical phenomena due to both interactions. In order to evaluate local wall thinning due to FAC and LDI, a 6-step evaluation procedure for each has been proposed. (1) For FAC wall thinning evaluation, corrosive conditions along the flow path and mass transfer coefficients at the structure surface under single- and two-phase flow conditions were calculated to evaluate high-risk zones for FAC occurrence, and then wall thinning rates were calculated with the coupled model of static electrochemical analysis and dynamic oxide layer growth analysis at the identified high-FAC-risk zones. (2) For LDI local wall thinning evaluation, trajectories of liquid droplets in high-velocity steam, their collision densities and velocities on the pipe inner surface were calculated to evaluate high-risk zones for LDI occurrence, and then local wall thinning rates were calculated at the identified high-LDI-risk zones. For the region with steam velocity higher than 200 m/s, the wall thinning rate was determined mainly by erosion (LDI (erosion)), while for that with velocity lower than 100 m/s, it was determined by corrosion (LDI (corrosion)). (3) The FAC wall thinning model was applied for two-phase flow as well as single-phase flow, while the FAC wall thinning model was applied for LDI (corrosion). The empirical formula was applied for LDI (erosion) after discriminating LDI (corrosion) and LDI (erosion) based on steam velocity and quality.

1705 発電用設備規格委員会疲労評価タスクの活動状況(OS17-2 低サイクル疲労とその応用-設計疲労線図の検討-)

Fatigue failure is one of the essential phenomena to be considered in the design and integrity assessment of light water reactor plants and the methods for assessing its significance are given in the structural design standards. Decrease in fatigue strength due to environmental effect of high-temperature water ns reported and its introduction into the design standards gains a large importance. This issue and others related to fatigue evaluation for light water reactor plants have been discussed in a task group formed in the committee for power equipments codes organized in JSME. This report summarizes the updated infonnation on the activities in the task group including those on design fatigue curves, environmental effect as well e flaw tolerance approach.

The impact of 238U resonance elastic scattering approximations on thermal reactor Doppler reactivity

The effects of accurate modeling of neutron scattering in 238U resonances are analyzed for typical light water reactor (LWR) and next generation nuclear plant (NGNP) lattices. An exact scattering kernel is formulated and implemented in a newly developed Monte Carlo code, MCSD (Monte Carlo slowing down), which solves a neutron slowing down in an infinite homogeneous medium and is used to generate resonance integral data used in the CASMO-5 lattice physics code. It is shown that the exact scattering kernel increases LWR Doppler coefficients by ∼10% relative to the traditional assumption of asymptotic elastic downscatter for 238U resonances. These resonance modeling improvements are shown to decrease hot full power eigenvalues by ∼200 pcm for LWRs and ∼450 pcm for NGNPs.

J0301-1-4 軸・捩り複合荷重下の疲労寿命評価(規格における疲労関連課題と方向性(1))

Fatigue failure is one of the essential failure modes to be considered in the design of light water reactor plants. Treatment of stress multiaxiality in the fatigue assessment is of practical importance for many components subjected to multiaxial cyclic loading during operation. Based on the results of multiaxial fatigue tests on hollow and solid bar specimens of SUS316NG and SGV410 under combined axial and torsional loading, accuracy of fatigue life prediction based on the conventional parameters as well as a newly proposed two parameter approach proposed here was studied. The conventional parameters tended to provide conservative life prediction for the case of shear-dominated cases especially for 316NG while the new approach was able to give more realistic prediction.

Challenges in coupled thermal–hydraulics and neutronics simulations for LWR safety analysis

The simulation of nuclear power plant accident conditions requires three-dimensional (3D) modeling of the reactor core to ensure a realistic description of physical phenomena. The operational flexibility of Light Water Reactor (LWR) plants can be improved by utilizing accurate 3D coupled neutronics/thermal–hydraulics calculations for safety margins evaluations. There are certain requirements to the coupling of thermal–hydraulic system codes and neutron-kinetics codes that ought to be considered. The objective of these requirements is to provide accurate solutions in a reasonable amount of CPU time in coupled simulations of detailed operational transient and accident scenarios. These requirements are met by the development and implementation of six basic components of the coupling methodologies: ways of coupling (internal or external coupling); coupling approach (integration algorithm or parallel processing); spatial mesh overlays; coupled time-step algorithms; coupling numerics (explicit, semi-implicit and implicit schemes); and coupled convergence schemes. These principles of the coupled simulations are discussed in details along with the scientific issues associated with the development of appropriate neutron cross-section libraries for coupled code transient modeling. The current trends in LWR nuclear power generation and regulation as well as the design of next generation LWR reactor concepts along with the continuing computer technology progress stimulate further development of these coupled code systems. These efforts have been focused towards extending the analysis capabilities as well as refining the scale and level of detail of the coupling. This article analyses the coupled phenomena and modeling challenges on both global (assembly-wise) and local (pin-wise) levels. The issues related to the consistent qualification of coupled code systems as well as their application to different types of LWR transients are presented. Finally, the advances in numerical and computation techniques for coupled code simulations are summarized with outlining remaining challenges.

High-cycle fatigue behavior of type 316 stainless steel at 288 °C including mean stress effect

Establishment of a life evaluation method for structural components is essential for the rational operation and maintenance of light water reactor plants. High-cycle fatigue induced by mechanical vibration or temperature fluctuation is still a major cause of troubles of plants. In this study, high-cycle fatigue tests for austenitic stainless steel SUS316NG were conducted at 288 °C with/without mean stress. Fatigue tests with constant stress amplitude less than a certain level ran out over 10 8 cycles, which suggested SUS316NG has a fatigue limit at 288 °C. The effect of mean stress on fatigue life was investigated through the correlation with the variation of mean strain and strain amplitude. Methods to predict fatigue life under mean stress condition was newly developed based on the concept of ductility exhaustion. The stress-based approach such as the modified Goodman constituting the basis of the current design rule predicted excessively conservative fatigue life, while the developed methods gave more accurate prediction.

Methods for assessing NPP containment pressure boundary integrity

Research is being conducted to address aging of the containment pressure boundary in light-water reactor plants. Objectives of this research are to (1) understand the significant factors relating to corrosion occurrence, efficacy of inspection, and structural capacity reduction of steel containments and of liners of concrete containments; (2) provide the U.S. Nuclear Regulatory Commission (USNRC) reviewers a means of establishing current structural capacity margins or estimating future residual structural capacity margins for steel containments and concrete containments as limited by liner integrity; and (3) provide recommendations, as appropriate, on information to be requested of licensees for guidance that could be utilized by USNRC reviewers in assessing the seriousness of reported incidences of containment degradation. Activities include development of a degradation assessment methodology; reviews of techniques and methods for inspection and repair of containment metallic pressure boundaries; evaluation of candidate techniques for inspection of inaccessible regions of containment metallic pressure boundaries; establishment of a methodology for reliability-based condition assessments of steel containments and liners; and fragility assessments of steel containments with localized corrosion.

Maintenance Cycle Extension in the IRIS Advanced Light Water Reactor Plant Design

New nuclear power generation in the United States will be realized only if the economic performance can be made competitive with other methods of electrical power generation. The economic performance of a nuclear power plant can be significantly improved by increasing the time spent on-line generating electricity relative to the time spent off-line conducting maintenance and refueling. Maintenance includes planned actions (surveillances) and unplanned actions (corrective maintenance) to respond to component degradation or failure. A methodology is described that can be used to resolve, in the design phase, maintenance-related operating cycle length barriers. A primary goal was to demonstrate the applicability and utility of the methodology in the context of the International Reactor, Innovative and Secure (IRIS) design. IRIS is an advanced light water nuclear power plant that is being designed to maximize this on-line generating time by increasing the operating cycle length. This is consequently a maintenance strategy paper using the IRIS plant as the example.Potential IRIS operating cycle length maintenance-related barriers, determined by modification of an earlier operating pressurized water reactor (PWR) plant cycle length analysis to account for differences between the design of IRIS and this operating PWR, are presented. The proposed methodology to resolve these maintenance-related barriers by the design process is described. The results of applying the methodology to two potential IRIS cycle length barriers, relief valve testing and emergency heat removal system testing, are presented.