Reactor Head Research Articles

Improving the quality of performing in-service inspections using manipulators

Many in-service inspections, mainly carried out by the ultrasonic method, were until recently carried out manually or semi-automated with the use of encoders. These types of inspections usually lack data recording and are dependent on the experience of the operator performing these inspections. For these reasons, in recent years one of the important activities carried out within the Research Centre Rez has been the development and production of manipulators for non-destructive testing. This article will describe in more detail the manipulators used for ultrasonic testing of the reactor head nozzles, nozzles of the emergency power supply system of steam generators treated with weld overlay, inspection of low-pressure blade dovetails and others. The developed manipulators will make it possible to refine and improve the repeatability of inspections that were previously carried out manually (more precise guidance of probes, etc.) or were not carried out at all due to inaccessibility for operators. Among other examples of the use of manipulators is in conditions where manual testing is not possible. One of the applications can be measurement using the phased array technique on pipelines during service, when the temperature on the surface of the pipeline can be up to 320°C (the maximum temperature of the primary pipeline of a nuclear power plant).

Earthquake response of head-mounted equipment in advanced nuclear reactors

The seismic response of safety-related equipment mounted on the head of an advanced reactor, including pumps, control rod drive mechanisms, and reactor monitoring devices, will affect the design and layout of many advanced reactors. High earthquake-induced accelerations in such equipment may challenge their seismic qualification and trigger the need for additional support framing on the reactor head. Base isolation is a design solution that can drastically reduce seismic demands on equipment. This article describes a set of earthquake-simulator experiments conducted on a scale-model of a base-isolated reactor vessel including four representations of head-mounted equipment, with frequencies spanning from 4.5 to 27 Hz. Dynamic responses of the head-mounted equipment, including displacements, accelerations, and strains, were measured in the experiments for three support conditions: conventional, and seismically isolated using single concave Friction Pendulum (SFP) bearings and triple Friction Pendulum (TFP) bearings. Seismic isolation was effective at reducing equipment responses (accelerations, displacements, and strains) with respect to those in the conventionally supported vessel across a range of seismic inputs. Companion numerical studies highlight the accuracy to be expected in the calculation of different response quantities for lightly damped equipment. The importance of characterizing damping in head-mounted, safety-related equipment through physical experiments to support design and risk assessment is made clear through the numerical simulations.

Synthesis and Characterization of the Orthorhombic Sn3O4 Polymorph

AbstractAn unexplored tin oxide crystal phase (Sn3O4) was experimentally synthesized via a facile hydrothermal method. After tuning the often‐neglected parameters for the hydrothermal synthesis, namely the degree of filling of the precursor solution and the gas composition in the reactor head space, an unreported X‐ray diffraction pattern was discovered. Through various characterization studies, such as Rietveld analysis, energy dispersive X‐ray spectroscopy, and first‐principles calculations, this novel material was characterized as orthorhombic mixed‐valence tin oxide with the composition SnII2SnIVO4. This orthorhombic tin oxide is a new polymorph of Sn3O4, which differs from the reported conventional monoclinic structure. Computational and experimental analyses showed that orthorhombic Sn3O4 has a smaller band gap (2.0 eV), enabling greater absorption of visible light. This study is expected to improve the accuracy of hydrothermal synthesis and aid the discovery of new oxide materials.

Synthesis and Characterization of the Orthorhombic Sn3O4 Polymorph.

An unexplored tin oxide crystal phase (Sn3O4) was experimentally synthesized via a facile hydrothermal method. After tuning the often-neglected parameters for the hydrothermal synthesis, namely the degree of filling of the precursor solution and the gas composition in the reactor head space, an unreported X-ray diffraction pattern was discovered. Through various characterization studies, such as Rietveld analysis, energy dispersive X-ray spectroscopy, and first-principles calculations, this novel material was characterized as orthorhombic mixed-valence tin oxide with the composition Sn(II)2Sn(IV)O4. This orthorhombic tin oxide is a new polymorph of Sn3O4, which differs from the reported conventional monoclinic structure. Computational and experimental analyses showed that orthorhombic Sn3O4 has a smaller bandgap (2.0 eV), enabling greater absorption of visible light. This study is expected to improve the accuracy of hydrothermal synthesis and aid the discovery of new oxide materials.

Pressure loads generated by gas phase explosions in vessel-pipe geometry due to precompression and subsequent detonation

Reactors used for chemical production plants have a couple of pipes starting or ending in the upper reactor head. For most of the reaction time these pipes are blocked by a shut-off valve which is located at a certain distance from the point where the pipe is flanged to the reactor head. In many reactors the head space is filled with a gaseous mixture which can be explosive, either under upset operating conditions or sometimes even permanently during normal operation. If an accidental ignition occurs inside the head space of the reactor, unreacted mixture can be pushed (“precompressed”) into the attached pipes. When the flame propagating inside the head space of the reactor finally arrives at the point where a pipe is flanged to the reactor head, the pressure of the yet unreacted mixture inside the pipe can have risen up to typically 10 times the initial pressure. In the course of further propagation inside the pipe the flame can undergo the transition from deflagration to detonation (DDT) and the resulting pressures inside the pipe can become extremely large, in particular in those cases when the DDT occurs at the blinded pipe end such that DDT and reflection coincide. In this paper the entire phenomenon will be comprehensively explained and formulae will be derived that allow quantifying the static equivalent pressures to be expected in the attached pipes in dependence on the geometry of the vessel-pipe system, the process parameters (temperature and pressure) and the explosion characteristics of the involved gas mixture. In this way, a detonation pressure resistant design under precompression conditions can be realized. Finally, examples from real life involving ruptured pipes due to precompression with subsequent detonation are presented.

Batch-to-Continuous transition in the specialty chemicals Industry: Impact of operational differences on the production of dispersants

Specialty chemicals are typically produced in large-volume batch reactors characterized by straightforward operation but limited efficiency. Process intensification via transition from batch to continuous operation offers the prospect of strong improvements over current state-of-the-art in energy intensity, process safety, capital cost, and physical footprint. The present study investigated the production of two dispersants to demonstrate the viability of batch-to-continuous transition for the specialty chemicals industry: Succinate ester dispersants, formed in a one-step addition reaction, and succinimide dispersants, formed via a two-step reaction with water as by-product. While the ester process showed identical dispersant yields in batch and continuous operation under identical conditions, the succinimide process produced significantly higher yields in the batch compared to the continuous reactor. This could be traced back to an operational difference between these processes: In the batch process, the by-product water is continuously removed from the reactive mixture while purging the reactor head space, rendering the dehydration step effectively irreversible. In contrast, to prevent steam-induced foaming, the continuous reactor is operated at high-enough pressure to maintain water in the liquid phase. This renders the dehydration step incomplete due to thermodynamic equilibrium limitations. However, the full succinimide product yield could be recovered via subsequent continuous drying using a thin film evaporator. This product was indistinguishable from the batch product and fulfilled commercial specifications. Finally, full reversible reaction kinetics were derived based on the continuous reactor operation. Comparison between batch operation results and reactor modeling predictions indicated that the slow apparent kinetics of the industrial batch process is the result of mass transfer limitations due to slow water evaporation, and not caused by limitations in the intrinsic reaction kinetics.

How Risk-Informed Decisions Can Fail: Lessons from a Nuclear Power Plant Incident

In 2002, a large hole was discovered in the reactor vessel head of the Davis-Besse plant located in Ohio, US. Had the reactor pressure vessel ruptured, a major public safety event would have occurred. Further investigations revealed that, by the end of 2001, a shutdown order was drafted by the Regulatory Agency. Nonetheless, through a risk-informed decision-making approach, the Agency changed the decision related to the issuance of the order. Operations were then allowed to continue for six more weeks, despite the high likelihood that the reactor was operating with leaking cracks in the nozzles located at the reactor head. The reactor vessel head wastage, which had evolved unnoticed for years, was primarily due to borated water that leaked through a nozzle crack onto the carbon steel portion of the reactor head, resulting in boric acid corrosion. This paper aims to explore the Agency’s decision-making to identify factors that might have contributed to the flawed decision. A better comprehension on the organizational factors and social processes that can adversely affect risk-informed decisions is crucial to enhance operational safety in high hazard industries. The paper concludes by suggesting possible ways to guard against the identified factors.

Seismic isolation: A pathway to standardized advanced nuclear reactors

Standardizing advanced nuclear reactors is a pathway to substantially reducing their overnight capital cost and achieving parity with other power sources, including renewables and fossil fuels. The seismic load case has thwarted standardization of nuclear power plants because site-specific seismic hazard and local near-surface geology has triggered soil-structure-interaction analysis, design, equipment qualification, regulatory review, and licensing, ensuring that each build is different. To achieve standardized or site-independent certified advanced reactor designs, the impact of the seismic load case on the engineering and construction cost and time must be substantially mitigated. Seismic isolation is a mature technology that has been used for more than 30 years in non-nuclear sectors to substantially reduce earthquake demands in buildings and other infrastructure. In this paper, seismic isolation is used to enable standardization of advanced reactor designs, aimed at the complete re-use of a site-independent, certified design and repeated procurement of safety-class equipment. A pathway to standardized designs using seismic isolation is demonstrated for two fundamentally different advanced reactors: a molten salt reactor and a high temperature gas reactor. Each reactor building is equipped with three specialized pieces of safety-class equipment, namely, a reactor vessel, a steam generator, and a control rod drive mechanism housing that is attached to the reactor head. Analysis is performed per ASCE and ASME standards to design the buildings and the equipment for two base conditions: conventional (fixed base) and base isolated. The impact of the seismic load case is characterized for the reinforced concrete walls in the buildings and for the equipment, measured using vessel wall thickness and horizontal accelerations. The analysis results show that the fixed-base buildings, designed for a site of low seismic hazard (peak ground acceleration, PGA = 0.15 g) could be constructed at a site of much greater seismic hazard (PGA = 0.7 g) if seismic base isolation is employed. Importantly, the scope of the site-specific analysis, design, and qualification would be limited to the seismic isolators and the isolated substructure, drastically reducing plant-specific engineering, review, and licensing, and time to construction start. Regulatory challenges and opportunities with standardized reactor designs are identified.

Safety Evaluation of Clearance of Radioactive Metal Waste After Decommissioning of NPP

The Kori-Unit 1 nuclear power plant, which is scheduled to be decommissioned after permanent shutdown, is expected to generate large amounts of various types of radioactive waste during the decommissioning process. Among these, nuclear reactors and internal structures have high levels of radioactivity and the dismantled structure must have the proper size and weight on the primary side. During decommissioning, it is important to prepare an appropriate and efficient disposal method through analysis of the disposal status and the legal restrictions on wastes generated from the reactors and internal structures. Nuclear reactors and internal structures generate radioactive wastes of various levels, such as medium, very low, and clearance. A radiation evaluation indicates that wastes in the clearance level are generated in the reactor head and upper head insulation. In this study, a clearance waste safety evaluation was conducted using the RESRAD-RECYCLE code, which is a safety evaluation code, based on the activation evaluation results for the clearance level wastes. The clearance scenario for the target radioactive waste was selected and the maximum individual and collective exposure doses at the time of clearance were calculated to determine whether the clearance criteria limit prescribed by the Nuclear Safety Act was satisfied. The evaluation results indicated that the doses were significantly low, and the clearance criteria were satisfied. Based on the safety assessment results, an appropriate metal recycle and disposal method were suggested for clearance, which are the subject of the deregulation of internal structures of nuclear power plant.

Ultrasonic ranging technique for obstacle monitoring above reactor core in prototype generation IV sodium-cooled fast reactor

Abstract As the refueling of a sodium-cooled fast reactor is conducted by rotating part of the reactor head without opening it, the monitoring of existing obstacles that can disturb the rotation of the reactor head is one of the most important issues. This paper deals with the ultrasonic ranging technique that directly monitors the existence of possible obstacles located in a lateral gap between the upper internal structure and the reactor core in a prototype generation IV sodium-cooled fast reactor (PGSFR). A 10 m long plate-type ultrasonic waveguide sensor, whose feasibility has been successfully demonstrated through preliminary tests, was employed for the ultrasonic ranging technique. The design of the sensor's wave radiating section was modified to improve the radiation performance, and the radiated field was investigated through beam profile measurements. A test facility simulating the lower part of the upper internal structure and the upper part of the reactor core with the same shapes and sizes as those in the PGSFR was newly constructed. Several under-water performance tests were then carried out at room temperature to investigate the applicability of the developed ranging technique using the plate-type ultrasonic waveguide sensor with the actual geometry of the PGSFR's internal structures.

Leningrad NPP-2 start-up loss of power test and its simulation with use of KORSAR/GP code

Abstract The paper deals with the results of the Leningrad NPP-2 start-up loss of power test and its simulation with use of KORSAR/GP code. The test was held at the 1st unit with VVER 1200 reactor on the 29th of March. During the test the steam generators were refilled with EFW system at first. This system is connected to the SG steam space. Therefore, injection of water leads to the secondary pressure decrease. This phenomenon is difficult for simulation due to complicated structure of the SG internals. Also, the test provides unique data for the reactor facility cooldown at natural circulation in the primary circuit. The important data for reactor head and pressurizer cooldown were obtained. Thus, the test data provides the validation of several processes such as: emergency feedwater injection into the steam generator; steam generator filling with “cold” water from AFWS; pressure regulation at filled pressurizer; cooldown of the reactor head and the coolant circulation beneath the head. The results of the test were used for the validation of KORSAR/GP code, unit nodalization and equipment models.

Development of an In-Vessel CEDM for Small Modular Reactors

An ex-vessel–type control element drive mechanism (CEDM) is installed on the nozzles of a reactor vessel closure head. However, there has been a demand for locating CEDMs inside the reactor vessel because doing so provides many benefits such as eliminating rod ejection accidents, minimizing pressure boundary penetration, and simplifying the reactor head area structure. Accordingly, an in-vessel CEDM has been developed to operate inside the nuclear reactor. Developing an in-vessel CEDM is very challenging because of the harsh environment of high temperature, high pressure, high radiation, and submerged condition. During intense research work for several years, KEPCO E&C designed an in-vessel CEDM and studied its feasibility by electromagnetic analysis and thermal analysis. Then, a prototype of the in-vessel CEDM was manufactured to prove its performance and operability at normal reactor operating conditions. This paper introduces key design work and validation tests of the in-vessel CEDM including an in-vessel position indicator.

Crack growth rate evaluation of alloys 690/152 by numerical simulation of extracted CT specimens

Abstract While nickel-based alloys have been widely used for power plants due to corrosion resistance and good mechanical properties, during the last couple of decades, failures of nuclear components increased gradually. One of main degradation mechanisms was primary water stress corrosion cracking at dissimilar metal welds of piping and reactor head penetrations. In this context, precise estimation of welding effects became an important issue for ensuring reliability of them. The present study deals with a series of finite element analyses and crack growth rate evaluation of Alloys 690/152. Firstly, variation of residual stresses and equivalent plastic strains was simulated taking into account welding of a cylindrical block. Subsequently, extraction and pre-cracking of compact tension (CT) specimens were considered from different locations of the block. Finally, crack growth curves of the alloys and heat affected zone were developed based on analyses results combined with experimental data in references. Characteristics of crack growth behaviors were also discussed in relation to mechanical and fracture parameters.

The Design Analysis of Resistance to Brittle Crushing of WWER-1000 Reactor Head

The basic equipment of reactor plant WWER-1000 is operated under the severe conditions. The reactor head is no exception because this construction is subjected to severe dynamic and temperature influence which causes a degradation of its material properties during the operation. This construction is made of rugged material which is capable of sustaining all impacts occurring at the planned operation lifetime, but some head sections are more subjected to a brittle crushing.

Design methodology for insulating and cooling of a small modular reactor head by high temperature structural-thermal-fluid analysis

ABSTRACTThe design of a sodium-cooled fast reactor (SFR) head can be complicated due to its shape and functions. The head is a component placed in the pressure boundary to shield nuclear radioactive radiation. At the same time, it needs to seal the reactor vessel, support penetrating components, and minimize heat losses. This paper presents a new insulating and cooling design concept of a small SFR head. For a new design, this study shows a comprehensive design approach considering fluid-thermal-structural computations. The interactive design approach refers to dependent simulation steps of three-dimensional (3D) thermal-structural, one-dimensional (1D) heat-transfer, and 3D computational fluid dynamics (CFD) analysis. This multi-domain approach was applied to the head of the large sodium integral effect test facility called sodium test loop for safety simulation and assessment (STELLA-2). And the STELLA-2 head design was proposed as a thick plate with a sandwich type of insulation, cooling the perimeter annulus of the round head-top surface. For the structural design, the ASME design code was utilized, and the head temperature of 346 °C was calculated as its initial design temperature target. In an axial heat-transfer mode from the in-vessel to the head, a 1D finite element model gave 57 and 75 mm insulation thicknesses with a thermal conductivity of 0.07 W/m/K. The cooling effectiveness of the proposed head design was shown through a commercial CFD package.

Evaluation of Interface Defects in Inaccessible Reactor Shrink Fit Nozzle Welds Using Ultrasonic Waves

This study proposes an effective method to inspect inaccessible nuclear power reactor head nozzles using interface waves that propagate along the shrink fit boundary of a reactor head. The reactor head is relatively thick, which makes it difficult to inspect from the outside by conventional ultrasonic testing. However, interface waves can propagate a long distance from a fixed transducer position. The inside of the nuclear reactor has limited access due to the high radiation, so the transducers are located outside the nuclear reactor head, and interface waves propagate into the nuclear reactor to detect defects. A numerical simulation and experiments were carried out to validate the method. Various defect cases that simulate field failures are also presented, and the proposed technique shows satisfactory defect classification.

Integral seismic analysis of reactor head package and control rod drive mechanism

Integral seismic analysis of reactor head package and control rod drive mechanism

Dynamic Modelling and Simulation of Pilot Scale Anaerobic Digestion Plant Treating Source Separated Food Waste and Effect of Recycling Sludge

This study investigates the effect of recycling sludge and stability of pilot scale anaerobic digestion plant treating organic fraction of source separated food waste. Pilot plant comprises of pre-treatment, anaerobic digestion and post treatment. Anaerobic digestion is carried out under mesophilic conditions. At flow rate of 0.221 m3/d, slurry of food waste is introduced semi-continuously every day within 2 hr time span. Anaerobic digestion model No.1 (ADM1) was applied for modelling and simulation of continuous stirred tank anaerobic reactor including recycling and was implemented in AQUASIM 2.1f. Input food waste was characterized and parameters were determined as dictated by ADM1. Process parameters were obtained from pilot plant and kinetic parameters are standard parameters given in ADM1. Input Organic Loading Rate(OLR) to reactor is 4.81 kgCOD/m3.d and Hydraulic Retention Time (HRT) was 20 days. Pilot scale AD plant was simulated for 40 days. It was found that biogas production rate and gas composition vary according the intermittent feeding pattern and reactor head space contains average compositions of CH4 and CO2 of 56% and 30% (v/v) respectively. Average biogas production rate under this condition was 16.4 m3/d. Recycling of sludge in terms of biomass was implemented in model as an advective link. When model was simulated at 40 day solid retention time (SRT), average biogas production rate increased by 31%. By varying OLRs to reactor, stability was investigated. When input OLR was doubled in terms of hydraulic load, anaerobic reactor became unstable producing H2 (38% v/v) and CO2 (40%v/v) in reactor head space. Under this condition, bulk liquid phase pH was 4.78. Instability is further confirmed by accumulation of volatile fatty acids and inhibition of strict methanogens. ADM1 can be applied to model and simulate pilot scale anaerobic digestion plant and to screen different options before scaling up into large scale plants.

Residual stress assessment of nickel-based alloy 690 welding parts

Failures of welding parts in nuclear power plants have increased significantly. One of main degradation mechanisms was initiation and growth of primary water stress corrosion cracking accelerated by residual stress at dissimilar metal welds of piping, reactor head penetrations and nozzles connecting major components. Therefore, estimation of the welding residual stress became an important issue for ensuring reliability of them. The present study deals with verification of welding simulation techniques through sensitivity analyses of bottom-mounted instrumentation nozzle mock-up and control rod drive mechanism (CRDM) nozzle mock-up, examination of residual stress effects for a welded cylindrical block and application of the residual stress evaluation method to real penetration nozzles of CRDM made of nickel-based alloy 690.

판형 웨이브가이드 초음파 센서를 이용한 소듐냉각고속로 원격주사 검사기법 개발

In a sodium-cooled fast reactor, which is a Generation-IV reactor, refueling is conducted by rotating, but not opening, the reactor head to prevent a reaction between the sodium, water and air. Therefore, an inspection technique that checks for the presence of any obstacles between the reactor core and the upper internal structure, which could disturb the rotation of the reactor head, is essential prior to the refueling of a sodium-cooled fast reactor. To this end, an ultrasound-based inspection technique should be employed because the opacity of the sodium prevents conventional optical inspection techniques from being applied to the monitoring of obstacles. In this study, a ranging inspection technique using a plate-type ultrasonic waveguide sensor was developed to monitor the presence of any obstacles between the reactor core and the upper internal structure in the opaque sodium. Because the waveguide sensor installs an ultrasonic transducer in a relatively cold region and transmits the ultrasonic waves into the hot radioactive liquid sodium through a long waveguide, it offers better reliability and is less susceptible to thermal or radiation damage. A 10 m horizontal beam waveguide sensor capable of radiating an ultrasonic wave horizontally was developed, and beam profile measurements and basic experiments were carried out to investigate the characteristics of the developed sensor. The beam width and propagation distance of the ultrasonic wave radiated from the sensor were assessed based on the experimental results. Finally, a feasibility test using cylindrical targets (corresponding to the shape of possible obstacles) was also conducted to evaluate the applicability of the developed ranging inspection technique to actual applications.