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[en] Highlights: • A safety culture framework and a quantitative methodology to assess safety culture were proposed. • The relation among Norm system, Safety Management System and worker's awareness was established. • Safety culture probability at NPPs was updated by collecting actual organizational data. • Vulnerable areas and the relationship between safety culture and human error were confirmed. - Abstract: For a long time, safety has been recognized as a top priority in high-reliability industries such as aviation and nuclear power plants (NPPs). Establishing a safety culture requires a number of actions to enhance safety, one of which is changing the safety culture awareness of workers. The concept of safety culture in the nuclear power domain was established in the International Atomic Energy Agency (IAEA) safety series, wherein the importance of employee attitudes for maintaining organizational safety was emphasized. Safety culture assessment is a critical step in the process of enhancing safety culture. In this respect, assessment is focused on measuring the level of safety culture in an organization, and improving any weakness in the organization. However, many continue to think that the concept of safety culture is abstract and unclear. In addition, the results of safety culture assessments are mostly subjective and qualitative. Given the current situation, this paper suggests a quantitative methodology for safety culture assessments based on a Bayesian network. A proposed safety culture framework for NPPs would include the following: (1) a norm system, (2) a safety management system, (3) safety culture awareness of worker, and (4) Worker behavior. The level of safety culture awareness of workers at NPPs was reasoned through the proposed methodology. Then, areas of the organization that were vulnerable in terms of safety culture were derived by analyzing observational evidence. We also confirmed that the frequency of events involving human error decreases when the level of safety culture is high. It is anticipated that the causality between the safety culture awareness of worker and the state of safety at NPPs can be verified using the proposed methodology.
[en] Highlights: • Fuel performance codes are limited by empirical materials models correlated to burnup. • We propose mechanistic materials models based on the evolving microstructure. • Multiscale simulation is used with experimental data to inform model development. • The approach’s completion will require researchers working together around the world. - Abstract: Fuel performance codes are critical tools for the design, certification, and safety analysis of nuclear reactors. However, their ability to predict fuel behavior under abnormal conditions is severely limited by their considerable reliance on empirical materials models correlated to burn-up (a measure of the number of fission events that have occurred, but not a unique measure of the history of the material). Here, we propose a different paradigm for fuel performance codes to employ mechanistic materials models that are based on the current state of the evolving microstructure rather than burn-up. In this approach, a series of state variables are stored at material points and define the current state of the microstructure. The evolution of these state variables is defined by mechanistic models that are functions of fuel conditions and other state variables. The material properties of the fuel and cladding are determined from microstructure/property relationships that are functions of the state variables and the current fuel conditions. Multiscale modeling and simulation is being used in conjunction with experimental data to inform the development of these models. This mechanistic, microstructure-based approach has the potential to provide a more predictive fuel performance capability, but will require a team of researchers to complete the required development and to validate the approach.
[en] Highlights: • An emergency shutdown system for the TRR is carried out based on a heavy water tank. • The performance of the heavy water tank are carried out based on “first and equilibrium cores”. • Heavy water discharging flow rate is also studied in the current research. • Thermal flux in the radioisotope channel with and without the heavy water tank are studied. • A core with and without the heavy water tank for the cases of 5 × 6, 5 × 5, 5 × 4, and 4 × 4 fuel assemblies are investigated (for two types of fuel loading—first and equilibrium cores). - Abstract: In this paper, a neutronics design of the secondary (i.e., emergency) shutdown system for the Tehran Research Reactor (TRR) is carried out based on a heavy water tank design. The heavy water tank in a cylindrical shape is around the core, and calculations for the optimized radius and height of the tank are performed. The performance of the heavy water tank calculations are carried out based on two types of fuel loading, which are called the “first and equilibrium cores” of the TRR. For both cases, neutronics and standard safety analysis are taken into account, benchmarked, and described herein. Heavy water discharging flow rate is also studied in the current research, and the results are compared with the IAEA criteria. Moreover, thermal flux in the radioisotope channel with and without the heavy water tank (as the reflector) are studied herein. Specifically, a core with and without the heavy water tank for the cases of 5 × 6, 5 × 5, 5 × 4, and 4 × 4 fuel assemblies are investigated (for two types of fuel loading—first and equilibrium cores). Based on our optimization, the 5 × 5 fuel assembly, which is called “B configuration,” has better performance and efficiency than that of the other described layouts.
[en] Highlights: • BEPU analysis were performed with a scenario of PCS pumps fail simultaneously. • The results from BEPU and conservative analysis were compared. • The comparing result shows the applicability and advantages of a BEPU safety analysis. - Abstract: Best estimate plus uncertainty (BEPU) is a promising approach to the safety analysis of nuclear reactors, and the uncertainty calculation is a very important concern about it. BEPU ensures realistic safety margins and secures higher reactor effectiveness by adopting best-estimate codes and realistic input data with uncertainties, whereas the previous conservative analysis generates excessive conservatism by considering each input parameter separately. A loss of flow accident (LOFA) of a 5 MW open-pool type research reactor was selected as a sample problem for a BEPU uncertainty assessment. We selected the failures of all primary cooling system (PCS) pumps, which would cause the abrupt reduction of flow and the reversal of core flow. The significant contributors to the reactor safety were identified and then input sets were sampled. For the uncertainty evaluation, 124 calculations were performed. This is the number of code runs required for a 95%/95% level with the 3rd order Wilk’s formula. The MOSAIQUE software developed by Korean Atomic Energy Research Institute (KAERI) was used for automated sampling of the uncertainty parameters, a global uncertainty calculation, and post processing of the results. The critical heat flux ratio (CHFR) and the fuel centerline temperature (FCT) were calculated at the 95%/95% level and were compared with those from conservative analyses. In addition, the impact of each design variables on the safety parameters was estimated by sensitivity analysis.
[en] Highlights: • A fully-implicit high-order system thermal-hydraulics model for advanced non-LWR. • Applied fully-coupled JFNK solution scheme to avoid the operator-splitting errors. • The effects of different spatial and temporal discretization schemes are investigated. • High-order convergence rates of discretization schemes are verified. - Abstract: An advanced system analysis tool is being developed for advanced reactor safety analysis. This paper describes the underlying physics and numerical models used in the code, including the governing equations, the stabilization schemes, the high-order spatial and temporal discretization schemes, and the Jacobian Free Newton Krylov solution method. The effects of the spatial and temporal discretization schemes are investigated. Additionally, a series of verification test problems are presented to confirm the high-order schemes. It is demonstrated that the developed system thermal-hydraulics model can be strictly verified with the theoretical convergence rates, and that it performs very well for a wide range of flow problems with high accuracy, efficiency, and minimal numerical diffusions.
[en] Highlights: • A new methodology to demonstrate compliance with the new ECCS acceptance criteria is described. • A wide spectrum of fuel rod initial burnup states can be analysed in the design phase. • The coupled suite PHISICS/RELAP5-3D has been used in the analyses. • A demo simulation of the equilibrium cycle, load-following and a LOCA analysis has been performed. - Abstract: The U.S. Nuclear Regulatory Commission is currently proposing rulemaking to revise the Loss Of Coolant Accident (LOCA) and therefore the Emergency Core Cooling System (ECCS) acceptance criteria, to include the effects of higher burnup on cladding performance as well as to address other technical issues. As motivated by the new rule, the need to use advanced cladding designs may be a result. A loss of operational margin may result due to the more restrictive cladding embrittlement criteria. Initial and future compliance with the rule may significantly increase vendor workload and licensee cost, as a spectrum of fuel rod initial burnup states may need to be analyzed to demonstrate compliance. Consequently, there will be an increased focus on licensee decision making related to LOCA analysis to minimize cost and impact, and to manage margin. The study here presented has been part of a big project used to investigate technical issues and approaches for future industrial applications within the Risk-Informed Safety Margin Characterization (RISMC) Pathway. Specifically, the primary aim of this study is to lay out a roadmap to demonstrate the application of the new methodology. The present analysis shows a simplified version of the methodology of an industrial application on the Core Design and the Multi-Cycle Analysis.
[en] Highlights: • A model for handling large number of frictional-contact constraints is developed. • The nonlinear problem formulated to be a Linear Complementarity Problem (LCP). • Natural frequencies of the CANDU fuel string with 12 and 13 bundles are obtained. • Different boundary conditions at the location of the downstream bundle are examined. - Abstract: A CANDU fuel string consists of 12 or 13 fuel bundles that laid horizontally inside a pressure tube. Under the large amount of drag load induced by the coolant flow the behaviour of individual bundles are coupled together through friction and contact. Presence of large number of contact and friction constraints causes the dynamical behaviour of many mechanical systems, like the vibration of CANDU fuel string, to be very complex. In this paper a numerical method based on linear complementarity problem (LCP) is developed to simulate vibrational behaviour of such systems. Then the presented method is employed and for the first time, natural frequencies of a CANDU fuel string are obtained numerically. Knowing the natural frequency of the string is very beneficial and can help to mitigate and decrease the damages, improve new fuel designs and develop new safety standards. All the solid components are discretized in space domain by the means of finite element method. Bozzak-Newmark integration scheme is employed to discretize the system equations of motion in the time-domain. With the computational power available today, frictional contact among fuel elements via spacer pads, between fuel elements and the pressure tube via bearing pads, and between neighbouring fuel bundles via endplates are modelled and the response of the string is obtained. Two different fuel string consist of 12 and 13 bundles are studied in this paper. FFT analyses are performed and natural frequencies of the systems are extracted. Results show great agreement with experimental values. The effect of boundary conditions in the last endplate of the downstream bundle on the natural frequencies is also investigated.
[en] Highlights: •This paper reports an experimental study of air-water pool entrainment with side exit. •Total 150 sets of test data cover all sub-regimes of momentum controlled region. •The validity of existing correlation in low and intermediate gas flux regimes is proved. •A new correlation of pool entrainment with side exit in high gas flux regime is proposed. -- Abstract: Pool entrainment in upper plenum is an important safety related phenomenon in SBLOCA transient for advanced PWR plants like AP1000. Due to the unique geometric characteristics, the application of standard Kataoka model in the upper plenum/hotleg entrainment phenomena requires significant caution. Based on a careful review of recent pool entrainment studies, this research carried out an experimental study of air-water pool entrainment with and without side exit/outlet to better capture the prototypic geometry of upper plenum/hotleg arrangement. Total 150 sets of test data are obtained at various combinations of gas velocities and water levels. The test data covers several entrainment regions from low gas flux region to near saturation region. The results show that the side exit/outlet will reduce the entrainment rate in high gas flux region and near saturation region. The mechanism is analyzed using visualization and CFD simulation. Based on the experiment data, a new correlation of pool entrainment with side exit/outlet in high gas flux region is proposed.
[en] To meet the energy requirement for a remote and isolated region, a concept of supercritical CO2 (S-CO2) cooled micro modular reactor (MMR) has been developed, which is easy to transport by fully modularizing the nuclear power system. From the previous work, only the on-design performances of MMR were presented. However, since the nuclear system needs to meet high safety standard, the design can be finalized only after the evaluation of the nuclear system response under multiple transient conditions. Such transient conditions may include part load operation and postulated accidents. The system response can be evaluated only with a high fidelity computer simulation, since it is almost impossible to construct a nuclear facility to test at such conditions. The simulation is performed to guarantee its autonomous operability and safety. Unfortunately, a system analysis computer code available for the S-CO2 cycle analysis is yet to be developed for the MMR safety evaluation. Description of a S-CO2 system analysis code platform is first presented in this paper. After that, the developed code is validated with experimental data. The steady state of MMR is first modeled with the code, which is followed by simulations of part load operation and postulated accidents of MMR. All simulation results show that the current design of MMR has the ability to maintain its safety and structural integrity even under extreme conditions to protect the public from radiation hazard at all times.
[en] Highlights: • A sub-channel code and point kinetic model are coupled. • The concept of “equivalent assembly” has been proposed and tested. • Reactor inherent safety can be enhanced by enlarge the P/D ratio. - Abstract: In this research paper, a sub-channel thermal hydraulic analysis code is coupled with the point reactor neutron kinetics model with six group delayed neutron. The coupling code is mainly used to perform the transient calculation of ADS/lead based alloy cooled fast reactor. The thermal hydraulic model is used for calculating temperature distribution profile and the feedback temperature information, providing input parameters for point kinetic model. This sub-channel analysis model can provide a new approach to solve the problem of one-dimension thermal hydraulic model and simulate the temperature distribution accurately. Furthermore the accuracy and reliability of calculated results are tested by another coupled code named FLUENT/PK and good agreements are achieved. To improve computational speed, one equivalent assembly is used to replace the whole core and the study shows that using of equivalent assembly which has the same average outlet temperature with the core obtained more reasonable results. The effects of fuel rods pitch diameter P/D ratio on simulation results are discussed. The code is capable to the quick calculations and safety analysis for reactivity accidents.