Reactor Building Research Articles

Influence of Vertical Equivalent Damping Ratio on Seismic Isolation Effectiveness of Nuclear Reactor Building

This paper aimed at evaluating the influence of different vertical equivalent damping ratios of a 3-dimensional combined isolation bearing (3D-CIB) as regards seismic response and isolation effectiveness. A comparative study of the seismic response in terms of acceleration floor response spectra (FRS), peak acceleration, displacement response of the nuclear reactor building, and dynamic response of the 3D-CIB was carried out. The results showed that: (1) the horizontal FRS is slightly influenced by the vertical equivalent damping ratio of 3D-CIB, whereas the increase of the vertical equivalent damping ratio has a significant effect on reducing the vertical FRS; (2) the peak vertical acceleration increased with the decrease in the vertical equivalent damping ratios of 3D-CIB and the difference of peak accelerations calculated by the damping ratio of 20 and 25% is within 10%; (3) the increase of the vertical equivalent damping ratio is capable of reducing the horizontal displacement and the rocking effect of the superstructure, and effectively controlling the vertical displacement amplitude; and (4) the vertical equivalent damping ratio of 3D-CIB has a slight effect on its axial force. Consequently, it is demonstrated that the increase of the vertical equivalent damping ratio is advantageous for isolation effectiveness. From the view of displacement control, it is suggested that the 3D-CIB with the vertical an equivalent damping ratio of 15~20% is appropriate and acceptable.

Isolation performance of a small modular reactor using 1D periodic foundation

Photonic crystals or man-made periodic materials own band gap characteristics, which can block the elastic wave propagate through the structure and can be used to mitigate the seismic waves. Firstly, the frequency band gap of a one-dimensional (1D) layered periodic foundation composed of reinforced concrete (RC) and rubber is derived by using the Floquet-Bloch theory. Secondly, a scaled small modular reactor (SMR) building was fabricated and carried out to evaluate the attenuation domain characteristics and isolation effectiveness of one-dimensional periodic foundation (1D PF) through shaking table tests. In addition, the finite element models (FEM) of SMR with 1D PF and reinforced concrete foundation (RCF) were established to verify the efficiency and the isolation mechanism of the 1D PF. Finally, the numerical and experimental results show that the 1D PF can effectively mitigate the seismic response and does not increase the relative displacement with a little overturning moment when the predominant frequencies of the seismic waves fall into the frequency band gap of the 1D PF.

3D FEM Soil-Structure Interaction Analysis for Kashiwazaki–Kariwa Nuclear Power Plant Considering Soil Separation and Sliding

In the 2007 Niigataken Chuetsu-oki earthquake, soil settlement, considered to be a result of the relative displacement between the soil and the structure, occurred alongside buildings at Kashiwazaki–Kariwa nuclear power plant. The objective of this study is to evaluate accurately the influence of the nonlinear behavior of the soil-structure interface on the response of the reactor building using a three-dimensional finite element model. To achieve this, we modeled the separation and sliding from sidewalls during a severe earthquake using joint elements, and evaluated the effect on the horizontal response of the structure. Through the soil-structure interaction analyses based on the recorded seismic data on the foundation for unit 7 reactor building using the three-dimensional finite element model, it was confirmed that the simulated horizontal responses of the building corresponded to the recorded seismic data. In addition, the comparison of the cases with and without separation and sliding revealed that the separation and sliding of soil from sidewalls reduces the embedment effects of the structure and that the reduction causes an occurrence of the foundation uplift. Moreover, in this study, it was clarified that the increase of the foundation uplift affected both the soil pressure characteristics beneath the foundation and along the sidewalls, and the maximum acceleration response of structures. In this study, the separation and sliding of soil from sidewalls had only a limited influence on the structural responses; however, under severe seismic excitation, it is expected that the embedment effects will be significantly reduced due to the separation and sliding of soil. Therefore, a precise response evaluation of the building including the nonlinear behavior due to separation and sliding should be considered in the seismic response analyses under severe seismic excitation.

Radon Concentrations in a Nuclear Reactor Center in Brazil.

Radiation workers are normally exposed to doses resulting from their day-to-day activities. Besides that, background radiation, such as radon, can contribute to the exposure rates. The radionuclide 222Rn is a noble gas belonging to the uranium series, and its indoor concentration in the air depends on the exhalation from surrounding soil and the exhalation from building materials. Radon exhaling from porous building materials containing high uranium concentrations can become a significant exposure factor in areas with limited ventilation. The objective of this study was to evaluate the 222Rn concentrations in the radiochemistry and radiometric laboratories in the nuclear reactor building of the Nuclear Reactor Center (CERPq) located in the Institute of Nuclear and Energy Research (IPEN), São Paulo, Brazil. Measurements were done using a Radon Gas Monitor, model RAD7, equipped with a solid-state alpha detector. A passive method (SSNTD) was also used, consisting of square pieces of C-39 foils (2.5 cm × 2.5 cm) placed within small diffusion chambers. The CR-39 detectors were etched in KOH 30% solution at 80 °C for 5.5 h in a constant-temperature bath. After etching, the detectors were washed, dried, and scanned using a microscope to obtain the track density measurements. The activity concentrations measured with both techniques varied from 52 to 103 Bq m-3 in the studied areas of the CERPq. These values may be compared to the reference level of 100 Bq m-3 established by the World Health Organization to ensure safety environments.

Modeling of the Radiological Behavior During LOCA in Open Pool Type Reactor

This study aims to model the radiological consequences that result from loss of coolant accident in open pool research reactor of 22 MW power. The loss of coolant accident results from rupture in the cooling systems of the reactor and consequently results in decreasing the water level in the reactor pool. Decreasing the water level, that represents the shielding material above the reactor core, results in increasing the dose rate in the top region in the reactor pool as well as the reactor building. LOCA normally results in shutting down the reactor due to the radioprotection system provided in the upper region of the reactor pool. A failure in shutdown systems of the reactor during LOCA is rare but expected and it represents the worst case. So, evaluating LOCA from the radiological point of view must be investigated for the safe shutdown mode as well as the failure of shutdown case. The study also includes occurring LOCA for shutdown reactor taking the decay time into consideration. ORIGEN-S module from scale system was used to estimate the delayed neutrons and gamma rays from the core for the shutdown mode. KENO-VI module from Scale system was used to estimate the prompt neutrons and gamma rays generated from the core for the failure case. Then, MAVRIC/Scale module was used to estimate the dose rate at the top point of the reactor pool for the two cases during LOCA. Studying the dose rate during LOCA aims to determine the water level that verifies the radiological safety limit at the top point of the reactor pool for the two cases. The calculations introduced a relationship between the water level that verifies the permissible radiological limit (WL) and the decay time (DT) in the range between 0 to 30 d.

Safety Investigation of Hazardous Materials Released from the Combined High Temperature Gas Cooled Reactor – Hydrogen Production Plant Using ALOHA Software

Safety Investigation of Hazardous Materials Released from The Combined High Temperature Gas Cooled Reactor (HTGR) – Hydrogen Production Plant Using ALOHA software has been carried out. Currently, most of studies for HTGR-hydrogen plant are focused only on the impact of hydrogen presence to the HTGR plant safety. Therefore, the objective of this study was to investigate the effect of the presence of natural gas and synthetic gas from Steam Methane Reforming hydrogen plant on the combine HTGR-Hydrogen production system using ALOHA software. Three selected hazardous materials: CH 4 , CO and H 2 were analyzed. The selected potential hazards of the hazardous materials after leaking from the pipe were downwind suffocation/toxication, flammable area and blast area from vapor cloud explosion. Two types of parameter, i.e., meteorological dispersion (including wind speed, temperature, humidity, nuclear building air changes for day and night) and source release parameters (including pipeline length, and distance from the reactor building to the hydrogen plant), were selected for this study. The effects of the parameters on the hazard distance were then analyzed. The study shows that hydrogen detector needs to be installed at the plant to ensure safety of field operator. Furthermore, CO adsorber and H 2 recombiner should be installed at the Reactor HVAC system for CO poisoning and H 2 fire protection. Provision of a separation distance of more than 250 meters or construction of a blast barrier between the reactor building and the hydrogen plant is also recommended to protect the reactor from H 2 explosion hazard.

Seismic Performance of Deeply Embedded Conceptual Advanced Reactors: Three-Dimensional Nonlinear Soil-Structure Interaction Analyses

The U.S. Nuclear Regulatory Commission’s functional containment concept provides advanced nuclear power plant designers with more flexibility in terms of the civil/structural design if the appropriate set of barriers for prevention of radioactive material release exist. Some of the conceptual advanced reactor structures, without the traditional pressure boundaries of large containment structures, are proposed to be deeply embedded or buried into soil. This approach is expected to provide (1) lesser seismic demands on the structures and safety-critical structures, (2) eased regulatory efforts and overall design against other external hazards such as aircraft impact, and (3) overall cost savings. One of the important aspects of assessing the technical and economic viability of deeply embedding advanced reactor buildings is to assess the seismic performance with the understanding of effects with material and geometric nonlinearities. This study investigates the seismic response of deeply embedded or buried advanced reactors by conducting three-dimensional nonlinear soil-structure interaction analyses. Although the results indicate that there is a general trend of decreased seismic response with increased embedment depths, the change in the dynamic environment with different embedment depths and the nonlinear environment under high-intensity seismic inputs may result in increased peak response at increased embedment depths.

Delayed gamma radiation simulation in case of loss of water event using Monte Carlo method

Monte Carlo N-Particle (MCNP) transport code accelerated by AutomateD VAriaNce reducTion Generator (ADVANTG) code was used to simulate the transport of delayed gamma particles emitted from irradiated TRIGA fuel. At the first step, this method was verified on a simple experiment where irradiated fuel was packed inside a transport cask. The calculated dose rate field in the vicinity of the transport cask was compared to the measurements. Due to an acceptable agreement, the method was later used to analyse Loss of Water Event (LOWE) for the JSI TRIGA research reactor main and spent fuel pool.Main results of the analysis show that in the case of LOWE for the reactor pool, the reactor building has to be evacuated since dose rates exceed 2 mSv/h. The highest dose rates are expected at the reactor platform. The radiation from the core is backscattered from the ceiling and affects the dose rates inside the whole reactor hall and control room. In the worst case scenario, the operating personnel can start with corrective actions 10 days after the last reactor operation, when dose rates at the reactor platform drop below 20 mSv/h.If LOWE occurs for the spent fuel pool, the reactor building has to be evacuated. However, the corrective actions by the operating personnel can be started immediately after the accident. This is due to the fact, that the fuel can only be transported into the spent fuel pool after a three-month cooling period inside the main reactor pool, resulting in a lower fuel activity.

From Exploration to Re-Enactment: Instructional Uses of a Desktop Virtual Environment for Training Nuclear Plant Field Operators

Despite the growing use of virtual environments for training in complex industrial settings, we have little understanding of how these innovative settings transform training and trainers everyday work. This study investigates the instructional use of an industrial building’s virtual environment by expert trainers during a training session for nuclear power plant field operators. Drawing from the course-of-action theoretical and methodological framework, field notes, continuous video-recording of the training sessions and verbalisations during post-training self-confrontation interviews with the trainers were collected and analysed. The results point out four typical instructional uses of the virtual environment in authentic settings: (a) showing the material elements and spatial layout of certain areas of the reactor building, (b) displaying safe and typical paths through the building, (c) explaining functional aspects to help trainees develop an operating model of the nuclear building, and (d) sharing salient experience through real-life anecdotes. These typical uses and their related learning dimensions are anchored in the re-enactments of expert trainer’s embodied past events. The discussion develops the counterintuitive idea that from an instructional point of view, the intensified immersion of trainers afforded by the virtual environment seems less influential than emerging practice-based learning experiences. We conclude with new possibilities for improving learning through and for work thanks to re-enactment of expert trainers’ past work practices.

Evaluation of the reactor building environmental conditions under a LOCA for the decommissioning Chinshan plant

Both units of the Chinshan plant have been approved for decommissioning. However, fuel assemblies cannot be discharged from the reactor core because the spent fuel pool (SFP) is almost full. Currently, the reactor remains open and connected to the SFP. This configuration would persist for years. Considering this, the nuclear regulatory authority requires the safety-grade instruments in the reactor building to remain operable under accident conditions. However, a loss of coolant accident in this configuration was not analyzed before. In this study, thermal analysis using GOTHIC is performed to evaluate the environmental conditions in the reactor building. The reactor, SFP, primary containment and reactor building are included. The results show that the reactor building temperature remains below 48.9 °C during 6 days following the accident. Thus, recovering any cooling function in 6 days following the accident is suggested. This ensures that the reactor building space is in a mild environment.

Generation of Design Time History Complying With Japanese Seismic Design Standards for Nuclear Power Plants

Seismic designs for Korean nuclear power plants (NPPs) under earthquakes’ design basis are noticed due to the recent earthquake events in Korea and Japan. Japan has developed the technologies and experiences of the NPPs through theoretical research and experimental verification with extensively accumulated measurement data. This paper describes the main features of the design-time history complying with the Japanese seismic design standard. Proper seed motions in the earthquake catalog are used to generate one set of design time histories. A magnitude and epicentral distance specify the amplitude envelope function configuring the shape of the earthquake. Cumulative velocity response spectral values of the design time histories are compared and checked to the target response spectra. Spectral accelerations of the time histories and the multiple-damping target response spectra are also checked to exceed. The generated design time histories are input to the reactor building seismic analyses with fixed-base boundary conditions to calculate the seismic responses. Another set of design time histories is generated to comply with Korean seismic design procedures for NPPs and used for seismic input motions to the same reactor containment building seismic analyses. The responses at the dome apex of the building are compared and analyzed. The generated design time histories will be also applied to subsequent seismic analyses of other Korean standard NPP structures.

Development and analysis of seismic experience database of structures, systems and components in nuclear power plants based on investigation reports and maintenance records

In this paper, for the development of reasonable seismic capacities and the assessment of the actual seismic margin of components and piping, we have developed a database of their seismic experience in Japanese NPPs resulting from actual large earthquakes, and investigated the relationships among the damage state of each component and piping and its location, aseismic class and damaged functions using the database. The seismic experiences of 30 plants affected by six recent large earthquakes were surveyed and investigated, and more than 1000 experience data were extracted. A unified damage level was also defined to standardize the damage level classification of the experience data, and the damage level of experience data of each component and piping was determined in accordance with the condition of the damage in terms of the support, boundary and active functions of the components and piping. The database includes some cases of function deterioration or function loss. However, considering a sufficient number of no damage cases in the database and the total number of components and piping, the proportion of these cases was considered as low. Also, whereas there were a few (damage) cases observed in their reactor buildings, no cases of serious damage case to components with important functions related to nuclear safety were observed.

Integration of LiPb loops for WCLL BB of European DEMO

The eutectic alloy Lithium Lead (LiPb) enriched at 90 % in 6Li is the breeder material for one of the candidate European Breeding Blanket (BB) concepts. Currently under investigation for DEMO reactor, the Water Cooled Lithium-Lead (WCLL), and for the WCLL Test Blanket Module (TBM) that will be qualified in the ITER reactor. The LiPb alloy is used as tritium breeder, neutron multiplier and tritium carrier. The design of the LiPb loops is currently under study and the conceptual design of the main loop components has been completed. For this reason, it becomes mandatory to proceed with the integration of the LiPb loops in the EU DEMO Tokamak building, checking the consistency of the different systems design to be integrated in DEMO reactor building. CAD design and integration of the entire LiPb loops are shown taking into account the building areas assigned, the interfaces with the other systems and the requirement related to the LiPb loop functions. An initial layout of the pipework and the position of the main components have been defined on the basis of the following design requirements: (I) gamma radiation shielding of the components and the pipework; (II) target flow velocity of the LiPb; (III) thermal expansion of the pipes; (IV) possibility to drain the entire loop; (V) redundancy of the loops; (VI) remote maintenance; (VII) position in the building and dimensions of the storage tanks. The 3D model of the entire loops has been provided and integrated in DEMO Tokamak building pointing out the issues related to the interfaces with the other systems and with the building itself.

New highly radioactive particles derived from Fukushima Daiichi Reactor Unit 1: Properties and environmental impacts

A contaminated zone elongated toward Futaba Town, north-northwest of the Fukushima Daiichi Nuclear Power Plant (FDNPP), contains highly radioactive particles released from reactor Unit 1. There are uncertainties associated with the physio-chemical properties and environmental impacts of these particles. In this study, 31 radioactive particles were isolated from surface soils collected 3.9 km north-northwest of the FDNPP. Two of these particles have the highest particle-associated 134+137Cs activity ever reported for Fukushima (6.1 × 105 and 2.5 × 106 Bq per particle after decay-correction to March 2011). The new, highly-radioactive particle labeled FTB1 is an aggregate of flaky silicate nanoparticles with an amorphous structure containing ~0.8 wt% Cs, occasionally associated with SiO2 and TiO2 inclusions. FTB1 likely originates from the reactor building, which was damaged by a H2 explosion, after adsorbing volatilized Cs. The 134+137Cs activity in the other highly radioactive particle labeled FTB26 exceeded 106 Bq. FTB26 has a glassy carbon core and a surface that is embedded with numerous micro-particles: Pb–Sn alloy, fibrous Al-silicate, Ca-carbonate or hydroxide, and quartz. The isotopic signatures of the micro-particles indicate neutron capture by B, Cs volatilization, and adsorption of natural Ba. The composition of the micro-particles on FTB26 reflects the composition of airborne particles at the moment of the H2 explosion. Owing to their large size, the health effects of the highly radioactive particles are likely limited to external radiation during static contact with skin; the highly radioactive particles are thus expected to have negligible health impacts for humans. By investigating the mobility of the highly radioactive particles, we can better understand how the radiation dose transfers through environments impacted by Unit 1. The highly radioactive particles also provide insights into the atmospheric conditions at the time of the Unit 1 explosion and the physio-chemical phenomena that occurred during reactor meltdown.

Modelling the work of the transport complex

The modelling of the transport complex is made using an automated system and an example of the enterprise that uses the dangerous materials. The example of such enterprise is a nuclear power station. The automated system used for modelling is a web-based intranet application that realizes monitoring and control functions. The automated system uses a system of objects included into the model of the enterprise, namely: gates, roads, transportation devices and controllers, airtight zone, describing the reactor building, gateway which allows to move cargo into and out from the building, non-airtight zone, which describes the outside area, and the zone of loading and unloading the cargo. The system receives input signals of discrete or analogue origin. Basing on the results of data collection, the algorithms of forecasting the crucial changes in states of main elements of the transport complex were developed. The automated system models the workflow of elements and the whole transport complex.

Direct Method for Generating Floor Response Spectra considering Soil–Structure Interaction

ABSTRACT A methodology is developed for generating floor response spectra (FRS) considering the effect of dynamic soil–structure interaction (SSI) based on the substructure technique and a recently-developed direct spectra-to-spectra method. Dynamic stiffness matrix of a three-dimensional structure with a rigid foundation subject to three translational and three rotational base excitations is derived, and is expressed in terms of the modal information of the structure. A transfer matrix, depending on the dynamic stiffness matrices of the structure-foundation system and generalized soil springs, is developed to modify the tri-directional response spectra at the foundation level of free-field (FIRS) in soil medium. The modified response spectra, called foundation level input response spectra (FLIRS), are used as the input to the fixed-base structure to generate FRS using the direct spectra-to-spectra method. Numerical examples of a reactor building and a service building in nuclear power plants demonstrate the accuracy and efficiency of the proposed method. FRS with % and % Non-Exceedance Probability (NEP) given by the proposed method agree well with the FRS obtained from a large number of time history analyses. Furthermore, the effect of SSI on FRS is studied, the physical meaning, and the advantages of the proposed method are highlighted.

Real-time Simulation Methods for Robots with a Flexible Arm Based on Computer Graphics Technology

A robot with a flexible arm that is controlled with a remotely operated water-pressure mechanism has been developed for dismantling objects that are heavily contaminated by radioactive materials during decommissioning of nuclear power plants. The objective of this research is to develop a process planning support system and method that can improve the accuracy of estimating the time required by the robot to complete its dismantling activity, support recovery from delays, and determine the feasibility of conducting a dismantling process in the reactor building. Since the flexible arm has a more complex mechanism and shapes those that of multi-axis robots, unique movable arm structures were modeled with a tool for three-dimensional (3D) computer graphics (CG) technology. The 3D CG model was used to make valid operation sequences for planning the motion of the robot. With the help of a prototype system, motion planning can perform to calculate the duration needed for the robot to complete its operation. The calculated duration is then used for updating the planned duration of a specific activity in a dismantling schedule. To plan the robot behaviors for complex dismantling processes, a simplified planning method with few virtual controllers based on a spline inverse kinematics (IK) with a non-uniform rational B-spline (NURBS) curve for an arm comprised of pistons and cylinders pressurized by water was studied. A prototype system for planning the behaviors of the robot was evaluated, and it was confirmed that the movement trajectory of the robot and the three-dimensional isometric display could be visualized using the mesh model generated from point-cloud data used to make the environment model of the robot. It was also confirmed that the operations involved in a specific activity of the robot could be completed within the duration determined in the simulation.

Simulation of hydrogen deflagrations with the lumped parameter code COCOSYS

In case of a postulated loss-of-coolant accident (LOCA) in a light water reactor, core degradation might occur if emergency measures fail. During this process a large amount of hydrogen might be generated by the oxidation of zirconium in the fuel rod cladding with steam. If the hydrogen leaks into the containment of a pressurized water reactor or the reactor building of a boiling water reactor and if the design concept of passive autocatalytic recombining is not adequate or fails due to adverse atmospheric conditions, it might form a combustible mixture with air. In case of ignition, pressure peaks can occur that are relevant for the integrity of the containment or the reactor building and safety systems. In the past 20 years 45 hydrogen deflagration tests (HD tests) have been performed in the THAI/THAI+ test facility with a free volume of 60–80 m3 which is operated by Becker Technologies GmbH in Germany. The lumped parameter code COCOSYS V3.0 (Containment Code System) which is part of the software package AC2 2019 developed by Gesellschaft für Anlagen-und Reaktorsicherheit (GRS) gGmbH offers the models FRONT and COMB for the simulation of hydrogen deflagrations in the containment or reactor building of a nuclear power plant (NPP). COCOSYS simulations of select THAI+ HD tests are shown, and model development potential is deduced.

ENHANCEMENT OF VALIDATION STUDIES FOR REACTOR DOSIMETRY AND ACTIVATION PREDICTIONS WITH THE NUCLEAR DATA SAMPLING METHODOLOGY

The CASMO/SIMULATE/MCNP/FISPACT-II calculation route has been established at the Paul Scherrer Institute (PSI) for reactor dosimetry and activation studies. Furthermore, the in-house tool NUSS is in use at PSI for nuclear data (ND) related uncertainties quantifications with Monte Carlo neutron transport calculations. The use of randomly sampled ACE-formatted ND files not only allows propagation of the ND uncertainties, but also can serve for assessing the applicability of different types of experimental data for validation of calculation predictions of parameters of interest. In the present work an application of the PSI calculation scheme for analysis of activation reaction rates and the fast neutron fluence (FNF), throughout the Swiss pressurised water reactor (PWR) and simplified containment building models, is demonstrated. As particular examples of potentially available experimental data, two kinds of the neutron flux monitors are considered: a) the reactor pressure vessel scraping samples and detectors placed in the dosimetry channels, mounted at the core barrel and designed for validation of FNF calculations, and b) the ex-vessel dosimeters, specifically used by the Swiss waste management organisation (NAGRA) for validation of bio-shield activation predictions. The calculations were done with the ENDF/B-VII.1 library. The obtained results demonstrate importance of the ND uncertainties for the dosimetry evaluations. The assessment of the applicability of the selected experimental information for validation of the bio-shield irradiation calculations was done based on evaluation of the ND-related Pearson correlation coefficients.

Labor costs management in the large-block construction of NPP

The article continues the research of the features of the use of prefabricated cast-in-place structures in the construction of nuclear power plants (NPP), the organizational and economic feasibility of the use of prefabricated cast-in-place structures in the implementation of new projects for the construction of nuclear power plants. The estimation of labor costs and resource costs of installation of large-block structures for the reactor building of the NPP was made. The dependence of labor costs for the installation of reinforced blocks, depending on the share of using reinforced blocks during the construction of the reactor building of a nuclear power plant, was obtained. The dependence of the overall labor costs for the industrial production and installation of reinforced blocks, depending on the share of using reinforced blocks during the construction of the reactor building of a nuclear power plant, was obtained. Graphical representations of dependencies are illustrated. The main daily resource consumption for the installation of reinforced blocks during large-block construction has been analytically calculated. The data on the required shares of labor costs at industrial enterprises and at the construction site in percentage terms were obtained. The conclusion is made about the rational use of large-block construction in the implementation of NPP construction projects.