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[en] These proceedings present the outcome of an international conference, at which the nuclear community had the opportunity to reflect on the pivotal role that human and organizational aspects play in assuring nuclear safety. Held 30 years after the Chernobyl accident which led to the international adoption of the concept of safety culture, the conference provided distinguished experts and practitioners with a unique opportunity to share insights from the past and visions for a safer future. The publication contains the conference opening and closing addresses, summaries of all conference sessions as well as the fully edited papers produced for the conference plenary sessions. The papers presented at the parallel sessions and dialogue sessions of the conference are included in their original form in the CD-ROM accompanying the publication.
[en] Highlights: • Transient scaling distortion of a single phase natural circulation is analyzed. • A Dynamical System Scaling (DSS) method is applied to assess the dynamic process. • The transient mass flow rate and temperature difference are compared and evaluated. - Abstract: Scaling analysis is widely used in the design of nuclear reactor passive safety systems to ensure those scaled-down test facilities can accurately capture important phenomena in a full-scale prototype. In this study, the transient scaling distortion of a single-phase natural circulation system was evaluated using the new Dynamical System Scaling (DSS) method. For convenience of comparison, the conventional Hierarchical Two-Tiered Scaling (H2TS) method, based on the initial static characteristic values, was applied first to determine the system scaling ratios. The different scaled-down cases based on the two methods were calculated with the Relap5 computational code. The results show that two different scaling number groups can be obtained based on the traditional H2TS method and the new DSS identity method, and both of the methods can effectively model the single phase natural circulation in a simple loop. The relative scaling distortion of the transient mass flow rate fluctuated sharply at the initial stage, when the power input increased step-wise, but gradually grew afterwards. In addition, with a smaller power ratio, the DSS identity method was more helpful for the scaled-down facility design.
[en] JSC TVEL has carried out a technical and economic study with the involvement of the National Research Centre Kurchatov Institute in the use of nuclear fuel enriched above the current limit of 5 wt% for VVER-1000/1200. The article presents neutronic characteristics of developed 18- and 24-month fuel cycles based on fuel enriched above 5 wt% and assessment of nuclear safety for fabrication and handling with high enriched fuel.
[en] Highlights: • This paper reviews the evolution of the use of probabilistic risk assessment in the U.S. regulations. - Abstract: This paper provides historical perspectives and insights on the early development of the U.S. nuclear regulatory process and its subsequent evolution towards risk-informed processes. After the landmark Reactor Safety Study (WASH-1400) and the TMI-2 accident, the U.S. Nuclear Regulatory Commission (NRC) began to use probabilistic risk assessment (PRA) methods and insights in regulatory applications as deemed necessary or useful. In 1995, the NRC adopted a policy that promotes increasing the use of probabilistic risk analysis in all regulatory matters to the extent supported by the state of the art to complement the deterministic approach. The NRC then started moving toward a much expanded use of PRAs in what is termed risk-informed regulatory approach. This paper discusses the challenges and the success stories of the use of probabilistic assessment of the risk to support and inform regulatory decisions.
[en] Highlights: • Autonomous control algorithm for safety functions was modeled with a FHF and an LSTM. • LSTM network was trained using a simulator and validated to demonstrate the effectiveness of the algorithm. • Autonomous control could manage the plant safety better than the current automation plus human control. - Abstract: With the improvement of computer performance and the emergence of cutting-edge artificial intelligence (AI) algorithms, an autonomous operation based on AI is being applied to many industries. An autonomous algorithm is a higher-level concept than conventional automatic operation in nuclear power plants (NPPs). In order to achieve autonomous operation, the autonomous algorithm needs to include superior functions to monitor, control and diagnose automated subsystems. This study suggests an autonomous operation algorithm for NPP safety systems using a function-based hierarchical framework (FHF) and a long short-term memory (LSTM). The FHF hierarchically models the safety goals, functions, systems, and components in the NPP. Then, the hierarchical structure is transformed into an LSTM network that is an evolutionary version of a recurrent neural network. This approach is applied to a reference NPP, a Westinghouse 930 MWe, three-loop pressurized water reactor. This LSTM network has been trained and validated using a compact nuclear simulator.
[en] Highlights: • This study presents a LBB leakage analysis code (LEABLE) for various conditions. • Criteria were employed in LEABLE to select the corresponding model. • Single-phase and two-phase critical flow models were proposed. • The LEABLE code was validated by experimental data. • The influences of different parameters on LBB leakage were studied by LEABLE. - Abstract: The leak before break (LBB) concept plays an important role in reactor safety. A code (LEABLE) for leak before break (LBB) leakage estimation was developed in this paper. Complicated crack morphologies and different conditions were taken into account, covering subcritical and supercritical fluids, single phase and multi-phases, choked and unchoked flow. Precise and reliable criteria were proposed for proper model selection. A modified two-phase critical flow model was presented and a single-phase critical flow model for superheat gas was established. Besides, modifications on flashing inception, flow resistance and particulate plugging were conducted in the LEABLE code. Moreover, the code was validated and verified by experimental data for both artificial and natural cracks. Reasonable agreements between experiment values and code predictions are shown for all the conditions. And in the present study the LEABLE shows a higher precision than existing models and commercial softwares (PICEP and SQUIRT). The parameter sensitivity study of LBB leakage was performed as well. The mass flux increases with crack opening crack opening displacement (COD), crack length and the entrance resistance coefficient, and decreases with global and local roughness, number and angle of turns. However, the entrance resistance coefficient and local roughness have limited effects on LBB leakage.
[en] VVER-440 reactors have been utilized in Slovakia since 1978. So far, the vast majority of their core loadings were designed in VUJE institute. This paper presents a description of the methods and procedures, which have been used for this purpose in the last decade. Main attention is focused on the calculating tools for core refuelling scheme optimization.
[en] Highlights: • Identification and selection of important initiating events. • Approaches in listing and methods in screening and grouping of IEs. • Focus on internal IEs due to the random failure of components and human error. - Abstract: A key element in the safety of any Nuclear Research Reactor design is the evaluation of the reactor's ability to withstand events that could reasonably be postulated to occur and, if unmitigated, could lead to core damage or radionuclides releases to the atmosphere. A first step to ensuring that the reactor design is sufficiently robust to withstand accidents is to identify a comprehensive list of IEs that might lead to core damage or radionuclide releases. This work seeks to present as comprehensive as possible the results obtained from identifying possible important initiating events (IEs) applied in the development of PSA Level-1 study for a 10 MW Water-Water Research Reactor (VVR). The methodology involves the listing approach and the IE screening and grouping methodologies and the focus was on internal IEs due to random failures of components and human errors with full power operational conditions and the reactor core was the radioactivity source. The results provided a set of IEs that were as systematic and as representative as possible, providing confidence to the completeness of PSA study. This study is one of the first few to address comprehensive steps to identify important IEs used in Level-1 PSA study.
[en] This Safety Requirements publication takes into account and incorporates developments relating to site evaluation for nuclear installations since the publication of IAEA Safety Standards Series No. NS-R-3 in 2003. It applies IAEA Safety Standards Series No. SF-1, Fundamental Safety Principles. Requirements for site evaluation are intended to contribute to the adequate protection of site personnel and the public and protection of the environment from harmful effects of ionizing radiation arising from nuclear installations. It is recognized that there are steady advances in technology and scientific knowledge, in nuclear safety and in what is considered adequate protection. Safety requirements evolve with these advances and this publication reflects the present consensus among States.
[en] Highlights: • Temperature fluctuations of lead-based reactor core outlet model were simulated. • Effect with the gap size between adjacent fuel assemblies was analyzed. • Effect with the opposite edge width of fuel assembly was analyzed. - Abstract: The temperature fluctuations induced by incomplete mixing of coolants with different temperature may cause thermal fatigue at the components of the lead-based reactor core outlet. Thus the accurate analysis of the phenomenon is very crucial for reactor safety operation. In this paper, the temperature fluctuations of the lead-based reactor core outlet were simulated by using large eddy simulation (LES) method in the simplified core outlet models. In order to analyze the temperature fluctuation sensitivity for the fuel assembly design parameters, such as the fuel assembly size and the gap between two adjacent fuel assemblies, five geometry models were constructed with different fuel assembly design parameters. The time histories of temperature fluctuations at different monitoring points on the center of three fuel assemblies were obtained. Then the amplitudes and the power spectrum density (PSD) of temperature fluctuations were analyzed, in order to compare temperature fluctuations of different geometry models at the same locations of core outlet. Finally the distribution characteristics of core outlet temperature fluctuations were obtained in axial directions, and the temperature fluctuation sensitivity with fuel assembly parameters was also analyzed based on the amplitudes, PSD and the normalized root-mean square temperature analysis. It is found that the temperature fluctuation intensity is enhanced with the increase of the gap size between adjacent fuel assemblies and the opposite edge width of each fuel assembly. The analysis results could provide important references for optimized design and engineering guidance of lead-based reactor.