Abstract: To investigate the seismic performance and isolation effect of a high-temperature gas-cooled reactor, a 1/20 scale model including a reactor, a spent-fuel plant, and a nuclear auxiliary plant was fabricated. In addition, 220 mm lead-rubber bearings were designed and produced for use in the shaking table test, which included both isolated and nonisolated conditions. Two historical earthquake records and three artificial earthquake motions were used to input the ground motion in the tests. The results demonstrated that the seismic performance of the plant was better and that the structure was in an elastic state, under a safe shutdown earthquake event. Isolation bearings were found to effectively reduce the dominate frequency of the structure. The acceleration amplification factor of the superstructure was found to be less than 1. The isolation test results showed that the peak of the floor response spectrum at the pressure vessel support was less than 0.1 g. In the nonisolation test, the peak of the floor response spectrum was greater than 1 g. In the isolation test, the relative displacement of the structure was less than 1.1 mm, which was relatively small. The structure maintained a good isolation performance and exhibited improved safety under extreme ground motion. PubDate: Tue, 31 Oct 2023 07:05:01 +000
Abstract: This paper introduces the utilization of the International Atomic Energy Agency’s toolkit for reactor technology assessment (RTA) application to deploy small modular reactors (SMRs) in the Czech Republic, Egypt, and Poland. The increasing demand for clean energy has led to the prominence of small modular reactors (SMRs) in addressing global energy challenges. The successful integration of SMRs into national energy systems necessitates comprehensive evaluations that take into account each country’s specific characteristics and energy requirements. RTA application represents significant progress towards innovative nuclear solutions, advancing a cleaner and more resilient energy future plan. The aim of this study is assessing the feasibility and advantages of SMR implementation in these countries, focusing on energy security, emission reduction, and long-term sustainability. Various SMR technologies, including NuScale, SMART, HTR-PM, BWRX-300, SMR-160, and RITM-200, are comparatively analyzed based on safety, scalability, efficiency, and economic viability. The findings reveal that BWRX-300 suits the needs of the Czech Republic and Poland, while RITM-200 is the optimal choice for Egypt. Moreover, NuScale also stands as a strong alternative for all three countries. This article emphasizes the importance of informed discussions and evidence-based decisions, promoting sustainable energy development and global advancements in nuclear technology. By utilizing SMRs, the Czech Republic, Egypt, and Poland can enhance energy security, reduce emissions, and meet rising energy needs sustainably. PubDate: Sat, 21 Oct 2023 09:20:00 +000
Abstract: In the failure analysis and safety assessment of dissimilar metal welded joints, the mechanical heterogeneity of local regions is usually ignored and limited sampling locations are selected. The mechanical behavior of the crack tip region is the main variables affecting the environmentally assisted cracking behavior, and it is crucial for understanding the impact of mechanical heterogeneity on the local stress-strain state at the crack tip in welded joints. In this study, the effect of mechanical heterogeneity on the local mechanical behavior at the crack tip and on the stress-strain condition at the crack tip front for different crack sizes was investigated through finite-element simulations based on user-defined material subroutines. The local mechanical behavior of an interface region and crack propagation direction with mechanical heterogeneity and a series of initial crack locations were analyzed. The results show that mechanical heterogeneity has a significant effect on the mechanical condition and growth path of cracks at different sampling locations. The interaction between the mechanical heterogeneity around the crack and the crack depth determines the stress and plastic strain in front of the crack tip, which causes a substantial change in the crack growth path. The interface cracks have high stress and plastic strain; thus, the interface is often the weak position where damage occurs. To guarantee a reliable integrity assessment of cracks in mechanically heterogeneous interface regions, local mechanical properties related to crack locations should be determined and utilized. PubDate: Wed, 09 Aug 2023 06:05:16 +000
Abstract: This paper introduces a controller unit for reactivity monitoring and automatic power control that was designed and constructed for the 500 kW Dalat Nuclear Research Reactor (DNRR). For power control and reactivity calculations, frequency signals from neutron measurement channels of starting and working ranges of the reactor are used. Two abovementioned independent functions were combined in an Artix-7 FPGA board for determining reactivity values by solving the point reactor kinetics equations with six delayed neutron groups and for stabilizing the reactor power at preset levels by determining the unbalance voltage signal to control the automatic control rod. With real-time calculations, the newly developed controller can monitor the reactor reactivity and control the reactor power online. The developed controller unit’s reactivity measuring and power stabilizing capabilities were assessed using the DNRR in normal operation and assumed emergency conditions and compared with those of the preexisting imported BNO-102R1 module of the DNRR control and protection system, known as ASUZ-14R. The results of the experiments show that the produced FPGA-based unit and the BNO-102R1 unit have the same technical characteristics and features, with the disparities being less than 5% and 1%, respectively, in reactivity measurement and power stabilization. The experimental data of reactivity measurements by the FPGA-based unit and the calculation results were also compared and found that the relative deviations between those are also less than 10%. The developed controller unit is capable of carrying out a variety of training and operational experiments on the DNRR. PubDate: Mon, 07 Aug 2023 10:50:01 +000
Abstract: The neutronics and thermal-hydraulics (N/TH) coupling behavior analysis is a key issue for nuclear power plant design and safety analysis. Due to the high-dimensional partial differential equations (PDEs) derived from the N/TH system, it is usually time consuming to solve such a large-scale nonlinear equation by the traditional numerical solution method of PDEs. To solve this problem, this work develops a reduced order model based on the proper orthogonal decomposition (POD) and artificial neural networks (ANNs) to simulate the N/TH coupling system. In detail, the POD method is used to extract the POD modes and corresponding coefficients from a set of full-order model results under different boundary conditions. Then, the backpropagation neural network (BPNN) is utilized to map the relationship between the boundary conditions and POD coefficients. Therefore, the physical fields under the new boundary conditions could be calculated by the predicated POD coefficients from ANN and POD modes from snapshot. In order to assess the performance of an ANN-POD-based reduced order method, a simplified pressurized water reactor model under different inlet coolant temperatures and inlet coolant velocities is utilized. The results show that the new reduced order model can accurately predict the distribution of the physical fields, as well as the effective multiplication factor in the N/TH coupling nuclear system, whose relative errors are within 1%. PubDate: Mon, 10 Jul 2023 03:05:00 +000
Abstract: In this paper, a nanosecond voltage comparator with PECL logic for a photon-counting radiation imaging system is presented. To realize a high-speed comparison of four gamma detector channels in a limited board space, quad comparators MAX9602 with PECL logic are chosen. Each of the four channels is coupled with a PECL to CMOS converter ICS508, which exports CMOS logic data for later use in an FPGA. Simulated findings for cobalt-60 with intensities ranging from 30 Ci to 300 Ci show little count loss caused by using a comparator and indicate ideal propagation delays at all source intensities. While in the laboratory test using a PCB-level system, signals with pulse width less than 3 ns might be dropped, and dispersion of propagation delay occurs. Despite these, the performance is still satisfactory and can meet the requirements of practical applications, as demonstrated by an improved result of 0.9% in the contrast indicator model. Further studies to optimize the circuit design can be conducted to gain improvement. PubDate: Sat, 08 Jul 2023 06:05:01 +000
Abstract: When the feedwater valve at the outage loop of the floating nuclear power plant leaks, thermal stratification occurs in the steam generator. It causes lower water temperature in the outage loop. The extent of hazard of this phenomenon cannot be directly determined by the existing measurement parameters, which poses a threat to the operational safety of the reactor. Therefore, this study adopts two routes: data-driven combined with safety analysis system (DSAS) and mechanism model-driven combined with safety analysis system (MSAS), to propose the prediction methods for the minimum temperature of the outage loop and the maximum power caused by the low-temperature coolant. Then, the actual data are used to verify these methods and the prediction results under different initial conditions are analyzed. The results show that both the DSAS method and the MSAS method can predict the minimum temperature of the steam generator in the outage loop and the maximum power when the outage loop is put into operation, but the DSAS method has better performance under certain conditions. These methods can provide guidance to the operators to avoid reactivity insertion accident. PubDate: Mon, 19 Jun 2023 10:20:01 +000
Abstract: Probabilistic risk assessment (PRA) is an effective methodology that could be used to improve the safety of nuclear power plants in a reasonable manner. Dynamic PRA, as an advanced PRA, allows for more realistic and detailed analyses by handling time-dependent information. However, the applications of this method to practical problems are limited because it remains in the research and development stage. This study aimed to investigate the possibility of utilizing dynamic PRA in risk-informeddecision-making. Specifically, the author performed an accident sequence precursor (ASP) analysis on the failure of emergency diesel generators that occurred at Unit 1 of the Tomari Nuclear Power Plant in Japan using dynamic PRA. The results were evaluated by comparison with the results of simplified classical PRA. The findings indicated that dynamic PRA may estimate lower risks compared with those obtained from classical PRA by reasonable modeling of alternating current power recovery. The author also showed that dynamic PRA can provide detailed information that cannot be obtained with classical PRA, such as uncertainty distribution of core damage timing and importance measure considering the system failure timing. PubDate: Wed, 07 Jun 2023 06:35:01 +000
Abstract: Establishing a dynamic model that accurately describes a realistic pressurized water reactor (PWR) fuel assembly is crucial to precisely evaluate the mechanical properties of the fuel assembly in seismic or loss of coolant accidents (LOCAs). The pluck test combined with the logarithmic decrement method has been widely applied in previous studies to extract fundamental modal parameters to calibrate dynamic models. However, most previous investigations focused on the first cycle of free vibration, which is strongly affected by stiction, baseline shift, drop conditions, and high-order mode interference, leading to inaccurate results. Moreover, these traditional methods cannot be used to extract high-order modal parameters. In this work, a novel experimental method for identifying the nonlinear modal parameters of a PWR fuel assembly is proposed. First, two algorithms are adopted to decompose the free vibration. Second, the local linearized modal parameters are extracted by a single-degree-of-freedom fitting method with a sliding window. Finally, these local linearized modal parameters are summed to obtain the nonlinear relationships between the modal parameters and amplitude. The new method makes more effective use of experimental data, obtains more accurate modal parameters than the logarithmic decrement method, and is capable of extracting high-order modal parameters. In the end, the test results are fitted by a fractional polynomial, which is of great value for numerical simulations. PubDate: Thu, 25 May 2023 08:05:05 +000
Abstract: A two-dimensional numerical model incorporating solid mechanics, electrochemistry, mass diffusion, and ion migration processes is developed to investigate the load effect on the crevice corrosion. The model is a transient model of crevice corrosion occurring in cracks of 304 stainless steel in a dilute NaCl solution, and the interaction between stress and electrochemical corrosion was considered. By solving the multiphysical coupling model in COMSOL, the effect of applied load on electrochemical corrosion in the crack tip region was calculated, and the local corrosion current density in the crack tip region with stress concentration within the crack was also calculated by using the Tafel relationship. The distribution of Fe2+ ion, Na+ ion, CL− ion, and H and O2 substance concentrations within the crack cavity is predicted by the equation analysis of substance transport. The results show that metal oxidation is more clearly affected by plastic deformation, the rate of hydrogen evolution reaction increases with stress enhancement, and the oxygen absorption reaction is not affected by stress strain. The distribution of iron ions, hydrogen, and oxygen within the crack is affected by the electrochemical reaction rate, and the distribution of iron ions, sodium ions, and chloride ions is affected by the electrolyte potential. PubDate: Wed, 03 May 2023 06:50:01 +000
Abstract: This research determines the Acceptable Level of Acceptance (ALA) based on the countries with active Nuclear Power Plant (NPP). The ALA is a particular value of public acceptance of NPP, indicating public support and participation in the program. If the public acceptance level is lower than the ALA, then the probability of public resistance against the program is relatively high and would harm the NPP. There is no correlation between the number of populations. This research uses four categories to classify public acceptance: (1) low, (2) moderate, (3) high, and (4) very high. Based on these categories, this research suggests that the moderate ALA is 27.5% of the acceptance level. PubDate: Sat, 11 Mar 2023 08:20:02 +000
Abstract: In CANDU reactors, shim operation is used when the online refuelling capability becomes temporarily unavailable. Adjuster rods, normally in-core to provide flux flattening, are withdrawn in sequence to provide additional reactivity as the fuel is depleted. In a CANDU 900 reactor, up to three of the eight adjuster banks may be withdrawn, with the power derated accordingly. In this study, the shim operation was modelled using a combination of TRACE_Mac1.1, PARCS_Mac1.1, and scripts modelling the reactor regulating system, all running as a single coupled simulation. A driver script simulated the operation as a sequence of steady-state, depletion, and transient models. The results were compared to operational data from a nuclear power plant, evaluating the key figures of merit. The simulation was extended beyond the operational data by reducing the power to 59% FP and withdrawing the remaining adjusters, to observe the behaviour of the simulated reactor for a deeper reactivity-driven transient. Sensitivity cases, including adjuster rod depletion and nuclear data uncertainty, were also evaluated. This study was able to successfully reproduce the general results of the shim operation. Some discrepancies were observed between the simulation and dataset for the duration of the shim, particularly for the one bank out phase of the shim. Several potential causes for the early phase behaviour were identified. When the simulation was extended, the model predicted that a power reduction below 60% FP would lead to xenon poison out when the adjusters were depleted, with the timing sensitive to the adjuster depletion. Nodalisation of the PARCS model also had a significant impact, due to the effect on adjuster nodalisation and its area-of-effect with respect to the actual adjuster locations. Nuclear data uncertainty had a lesser but still noticeable effect. Other parameters, such as the distribution of fuel burnups in the core, only had a small effect on the shim operation. PubDate: Wed, 08 Mar 2023 11:35:00 +000
Abstract: Owing to pipe thinning, fatigue damage, and aging, pipes, valves, and devices installed in the primary and secondary systems of nuclear power plants may leak high-temperature/high-pressure reactor coolant. Thus, a system must be developed to determine if the leakage is exceeding the operating limit of the nuclear power plant, thereby mitigating any loss of life or economic loss in such cases. In this study, a validated numerical analysis method was established to initially simulate the leakage behavior and subsequently to evaluate the small amount of leakage in the compartment. For this purpose, a vapor-jet collision test in the compartment and a vapor-jet test in the pipe were performed; numerical analysis was conducted, and comparative analysis was performed to verify the validity of the established method. The evaluation results suggested that the proposed numerical analysis method could optimally simulate the flow characteristics of the steam jet. Notably, compared to the existing evaluation method, the proposed approach simulated a more detailed behavior of the jet formed at the leakage point. In future research, the results of this study (data) will be used to inform the design of the second phase of the leak-capture system and will be served as the foundation for a performance-optimization study on the capture system. PubDate: Tue, 07 Mar 2023 14:05:01 +000
Abstract: Common cause failures (CCFs) may lead to the simultaneous unavailability or failure of numerous components in the nuclear power plant because of the existence of a shared cause when an initiating event disrupts the normal functioning of nuclear power plants. The presence of common cause failures (intra-unit and inter-unit) can be recognized in a multi-unit probabilistic safety assessment (MUPSA) as a crucial dependency factor that can influence accident scenarios and the core damage frequency (CDF), as CCF may affect the availability and proper operation of mitigating systems. Since such failures are likely to significantly undermine the benefits of the concept of redundancy in nuclear power plant systems, it is necessary to identify the CCFs that contribute to the core damage in a multi-unit site and analyse their overall quantitative magnitude and qualitative proportions. In this study, a twin-unit generic pressurized water reactor (PWR) nuclear plant is modeled using the AIMS-PSA software. For the loss-of-offsite-power (LOOP) and station blackout (SBO) events, the site CDF was calculated, and the cut-sets produced by this quantification were examined for the modeled CCF basic events in the fault trees. The quantitative and qualitative contributions of the CCFs to the frequency of site core damage were examined. CCFs in the modeled fault trees contributed to 4.58% to the site CDF of the combined LOOP followed by SBO event. In the LOOP event alone that leads to core damage, the CCF contributed 4.58% to the site CDF while CCFs contributed 17.19% to the site CDF in the SBO event alone that leads to core damage. With CCF events considered in the modeling process, the site CDF estimated with CCF events increased by 7.53% in the combined LOOP followed by SBO event. In the LOOP event alone that leads to core damage, inclusion of CCF events in the modeling increased the site CDF by 7.42%. A 15.66% increase in site CDF was recorded in the SBO event alone that leads to core damage as compared to modeling without CCF events. The results show how crucial the common cause failure contribution is to site CDF. The safety of the nuclear plant at a site is impacted by an increase in site CDF when common cause failures are considered. The various CCF fundamental event compositions and their percentage contributions were explicitly examined by the minimal cut-sets which leads to core damage in the units. In conclusion, this study’s findings can help us better understand how CCFs increase multi-unit site risk and can also act as a starting point for future studies on the qualitative and quantitative categorizations of CCF effects within MUPSA. PubDate: Fri, 03 Mar 2023 11:35:02 +000
Abstract: A deep-learning model was proposed for predicting the remaining time to automatic scram during abnormal conditions of nuclear power plants (NPPs) based on long short-term memory (LSTM) and dropout. The proposed model was trained by simulated condition data of abnormal conditions; the input of the model was the deviation of the monitoring parameters from the normal operating state, and the output was the remaining time from the current moment to the upcoming reactor trip. The predicted remaining time to the reactor trip decreases with the development of abnormal conditions; thus, the output of the proposed model generates a predicted countdown to the reactor trip. The proposed prediction model showed better prediction performance than the Elman neural network model in the experiments but encountered an overfitting problem for testing data containing noise. Therefore, dropout was applied to further improve the generalization ability of the prediction model based on LSTM. The proposed automatic scram prediction model can provide NPP operators with an alert to the automatic scram during abnormal conditions. PubDate: Fri, 03 Mar 2023 07:50:01 +000
Abstract: As an integrated computer code development for severe accident sequence analysis in Korea, CINEMA has been developing from an initiation event to a containment failure. The CINEMA computer code is composed of CSPACE, SACAP, and SIRIUS, which are capable of simulating core melt progression with thermal hydraulic analysis of the RCS (reactor coolant system), severe accident analysis of the containment, and fission product analysis in the vessel and the containment, respectively. The severe accident progression in TMI unit 2 has been analyzed as a part of a validation of the CINEMA computer code. This analysis has been performed to validate CINEMA models on the core melt progression, in particular, RCS thermal hydraulic behavior during core melt progression, fuel cladding oxidation with hydrogen generation, and fuel melting with relocation to the lower part of the core. The CINEMA results on main parameters, such as RCS pressure and an integrated hydrogen generation mass are compared with the TMI-2 data. The CINEMA results have shown that the RCS pressure is very similar to the TMI-2 data. The CINEMA results and measured total hydrogen production are very similar, which were approximately 465 kg and 460 kg, respectively. PubDate: Mon, 27 Feb 2023 11:20:00 +000
Abstract: In recent years, much attention has been dedicated to finding techniques to reduce exposure doses. This work examines the effectiveness of using serpentine concrete to shield a neutron source using a 241Am-Be neutron source facility at the National Nuclear Research Institute (NNRI) as a case study. The results obtained for both neutrons and gamma indicate that serpentine concrete provides better shielding as compared to ordinary concrete. At a distance of 100 cm from the Am-Be source, when shielded with serpentine concrete, it was found that personnel will receive an average gamma dose of 4.395.395 ± 0.122 μSv/h while a dose of 10.399 ± 0.083 μSv/h will be received for ordinary concrete shield. The average neutron dose equivalent at 100 cm, for ordinary concrete and serpentine concrete were 32.189 ± 0.277 and 9.276 ± 0.505, respectively. All dose equivalents obtained were also within internationally accepted limits. PubDate: Sat, 18 Feb 2023 07:35:02 +000
Abstract: In this study, an integrated human error simulation model in nuclear power plant (NPP) decommissioning activities (HEISM-DA) that can integrate and manage various factors affecting human errors is developed. In the HEISM-DA, an error probability input method suitable for the characteristics of each performance shaping factors (PSFs) was presented. Because each PSF has different importance on human error, the relative importance of decommissioning PSF Levels 1 and 2 and influential factors is considered. A multiplier was selected for each PSF and then used for human error evaluation. To calculate the human error probability (HEP) for the NPP decommissioning activity, the relationship between each PSF is identified and linked to develop a human error evaluation model. Using the HEISM-DA, HEP for reactor pressure vessel internal cutting work is evaluated based on the experience data. HEP is calculated to be approximately 1%. As a result of HEP calculation, it is found that the “operation” factor has a significant influence on the HEP of NPP decommissioning activities. Therefore, if the dismantling work is conducted by supervising the “operation” factors in a detailed and systematic approach, it is believed that the HEP will be reduced as other factors are also affected. PubDate: Thu, 12 Jan 2023 09:20:01 +000
Abstract: At present, the allowed outage time (AOT) of an M310 unit emergency diesel generator (EDG) was 3 days, which can be extended to 14 days through replacement of additional diesel units; although it provides a certain online maintenance time, it cannot meet the needs of ten-years overhauls. In order to avoid stopping the reactor for maintenance of NPP due to insufficient of EDG AOT, based on risk-informed method analysis feasibility of extending AOT for EDG to 28 days, we quantitatively calculate the impact of extension of AOT on risk level of nuclear power plants (NPPs). Analysis shows that extension of EDG AOT to 28 days has less impact on NPPs, and safety of NPPs can be further ensured through temporary risk control measures, so the extension of AOT to 28 days is acceptable. By using risk-informed technology to extend AOT for EDG, unnecessary shutdown and maintenance is avoided and the economy of NPPs and flexibility of maintenance work arrangement is greatly improved while ensuring safety, which is of great significance to operation and maintenance of NPPs. PubDate: Thu, 05 Jan 2023 12:05:02 +000
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