After years of operation, nuclear power plants accumulate a substantial amount of spent fuel stored in spent fuel pools. Prior to the completion of dry storage facilities, this spent fuel remains in ...the spent fuel pools for several decades. The progression of accidents in spent fuel pools leading to fuel damage typically spans several days or even weeks, which may lead people to overlook the importance of spent fuel pool safety. However, accidents involving spent fuel pools have a characteristic of low probability but high consequences. Their significance has been reemphasized, especially after the Fukushima accident. This study conducts an analysis of the scenario where a spent fuel pool experiences a complete loss of water. A plant equipped with a BWR-4 reactor and a MARK-I containment structure is used as the reference plant. The analysis utilizes the MAAP5 code to assess the impact of different fuel cooling times and ventilation flow rate of secondary containment building on the fuel temperature variation, considering only air circulation for cooling. Based on the analysis results, the plant utilizes the normal ventilation system of the reactor building, which can maintain the fuel temperature below 565 °C with a cooling time of 1 year. Additionally, it was found in the study that adequate ventilation flow rates and sufficient cooling time can ensure that the fuel temperature remains below 565 °C, thereby preserving fuel integrity.
Recent simulations by Scanlon et al. showed seemingly spectacular results for the Waymo self-driving vehicle in simulations of real accident situations. In this paper, it is argued that the selection ...criteria for accident situations must be modified in accordance with the relevant policy alternatives. While Scanlon et al. compare Waymo with
old
human-driven vehicles, it is argued here that the relevant policy question is whether we ought to use self-driven vehicles or human-driven vehicles
in the future
, which means that we need to consider whether other technological solutions, which are available but not broadly used in human-driven vehicles, could result in human-driven vehicles managing to avoid the same accidents. In this article, a proposal for a new standard of selection criteria is made.
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EMUNI, FIS, FZAB, GEOZS, GIS, IJS, IMTLJ, KILJ, KISLJ, MFDPS, NLZOH, NUK, OILJ, PNG, SAZU, SBCE, SBJE, SBMB, SBNM, UKNU, UL, UM, UPUK, VKSCE, ZAGLJ
•The impact of the SFP fuel assembly arrangement is more limited than expected.•The offload power does not present a crucial role in fuel damage progression.•User effects on the building modelling ...greatly impacts the SFP response.
This work performs an analysis of the inherent nuclear spent fuel pool response to a loss of pool cooling accident. The influence of fuel offload times, fuel assembly arrangement, fuel assembly binning for modelling, and building compartmentalization are evaluated. These impacts are assessed through sensitivity calculations performed with the MAAP code. Results are compared in terms of selected significant safety quantities such as fuel uncovery time, hydrogen generation, building temperature, relocated mass to the floor, and integrated total radiological dose at 1 km distance from the spent fuel pool.
According to the results, the impact of the fuel assembly arrangement on the output safety quantities is more limited than initially expected. Also, the offload power (as a function of elapsed time between shutdown and removal from the reactor pressure vessel to the spent fuel pool), relevant in the pre-uncovery phase, does not present a crucial role in fuel damage progression.
Somewhat unexpectedly, user effects on the building modelling hosting the pool significantly affect its inherent response. This is due to the large influence that natural circulation currents of gas have on mitigating fuel damage as such currents significantly rely on the modelling of the building in terms of nodalization, thermodynamic imposed conditions and heat sinks configuration. Follow-on work is identified to address uncertainties in the calculation of natural convective flows and heat transfer.
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GEOZS, IMTLJ, KILJ, KISLJ, NUK, OILJ, PNG, SBCE, UILJ, UM, UPUK
Interest in evaluation of reactor response and off-site consequences following beyond design basis accidents has significantly increased after Fukushima. MELCOR is an advanced computational aid that ...has been widely used and adapted to various reactor designs for severe core damage analyses. This paper summarizes the core thermal hydraulic response for a hypothetical severe accident caused by station blackout with failure of the steam generator safety relief valve at a Chinese pressurized reactor 1000-MW power plant – CPR1000 using MELCOR. Analytical results for a) temperature distribution of the fuel pellets, b) the fuel cladding, c) flow rate of the coolant, and, d) hydrogen mass changing over time are presented. The analyses are focused on safety assessment of the reactor core for severe accidents and are part of the overall evaluation of safety features of the CPR1000 reactor for residual risk posed by severe accidents.
•MELCOR simulation of core thermal response during a severe accident CPR1000 was conducted.•An original investigation in severe core damage accident is presented.•Consequences assessments lead to qualification of the mitigation strategies.
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GEOZS, IJS, IMTLJ, KILJ, KISLJ, NUK, OILJ, PNG, SAZU, SBCE, SBJE, UL, UM, UPUK
This study aims to analyze egress routes depending on, among other factors that influence evacuation in the event of an outbreak of fire, the characteristics of occupants and the initial points of ...fire, using the GongEgress simulation program. The simulation result shows that the evacuation time of the vulnerable users is found to take 18% longer than that of ordinary people, and the transfer passageways have lower survival probability compared to that of the platforms. Through the analysis of the results, the structural features of the underground subway station and the points of fire are proved to be the major factors that determine the survival probability of the occupants. Therefore, safety training for passengers through conducting fire drills at the station or fire accident simulations is considered necessary for prevention and proper response to underground subway fires.
Side-sweep accidents are one of the major causes of loss of life and property damage on highways. This type of accident is caused by a driver initiating a lane change while another vehicle is ...blocking the road in the target lane. In this article, we are trying to quantify the degree to which different implementations of vehicle-to-vehicle (V2V) communication could reduce the occurrence of such accidents. We present the design of a simulator that takes into account common sources of lack of driver awareness such as blind-spots and lack of attention. Then, we study the impact of both traditional, non-technological communication means such as turning signals as well as unidirectional and bidirectional V2V communications.
Children are increasingly transported in cars or other modes of road transportation. With this increased travel comes the higher risk of children becoming involved in a vehicle accident as an ...occupant. Among the injuries included in the Abbreviated Injury Scale (AIS), the head is one of the most frequently injured parts of the body. For a 3-year-old child (3-YOC), the most common injury observed during a road accident is skull fracture. This study proposed to develop a 3-YOC head finite element model (FEM) based on the geometrical three-dimensional reconstruction of two-dimensional slices obtained using X-ray computed tomography scanning geometries. In order to reproduce the skull fracture and brain injury with the head FEM, 13 domestic accident reconstructions involving fall cases of 3-YOC were collected from the paediatric emergency departments of different hospitals. After the numerical reconstructions of these fall cases, some mechanical parameters were extracted and then correlated with the observed injuries.
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BFBNIB, GIS, IJS, KISLJ, NUK, PNG, UL, UM, UPUK
•Severe accident codes are capable of performing simulations of stable steady-state and DBA conditions for iPWR designs, including natural circulation between the RPV and containment.•The impact of ...the safety system on circulation patterns should be further investigated.•Experiments in the area of natural circulation in integrated designs under various conditions would be useful to further validate integral codes such as MELCOR.
One of the coming Small Modular Reactor (SMR) designs is the integrated Pressurized Water Reactor (iPWR), which merges years of knowledge and experience in light water reactors with new demands from the market, such as flexibility, construction optimization, and reliability. Generally, SMRs are considered to be inherently safe and, in many cases, safer than generation III designs. Most of the SMR designs rely on passive safety systems at different levels of passive operation.
One of the ways to examine new concepts is a numerical simulation using a reliable tool. However, when new features come up, there are also new challenges for existing tools, even though they are continuously updated following the technology’s evolution. To investigate if implemented new features give expected results and to increase knowledge about SMRs modeling and simulations, accident transients of iPWR were calculated using MELCOR 2.2 code and analyzed.
The work aims to understand better numerical simulations of the analyzed iPWR accident scenarios and code reliability. To do so, the paper is divided into two parts; in Part 1, the authors present MELCOR 2.2 iPWR input deck description, including nodalization and steady-state conditions as well as the analysis of an iPWR LOCA-type scenario in the design basis accident domain. In addition, design basis accident progression is compared to beyond design basis scenario in which safety systems are postulated partly failing. The analyses of beyond design basis scenarios in which core degradation is observed are published separately in second part of that paper.
The obtained results indicate that MELCOR2.2 can simulate the thermal–hydraulic response of an iPWR in accident scenarios. Values received in this study in steady state and accident simulations are mostly expected and coherent with general knowledge of accident progression.
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GEOZS, IJS, IMTLJ, KILJ, KISLJ, NLZOH, NUK, OILJ, PNG, SAZU, SBCE, SBJE, UILJ, UL, UM, UPUK, ZAGLJ, ZRSKP
•Large validation task has been performed on the severe accident integral software ASTEC V2.2.•All physical phenomena are covered (except steam explosion), from thermal-hydraulics to Fission Products ...behaviour in the circuits and in the containment.•This task has been conducted by IRSN and partners of the recent NUGENIA ASCOM project.
Significant efforts are being continuously put for many years into the assessment of the severe accident integral code ASTEC developed by IRSN, through comparison with the results of most of the experiments developed internationally or through benchmarks with other severe accident simulation codes. For this assessment process, the IRSN code developers are supported by international partners, notably in the frame of the recent SNETP-NUGENIA ASCOM collaborative project.
This paper relates to the 3rd major version of the ASTEC V2 series, V2.2, that was released in 2021 to the ASTEC community. It aims at providing an overview of the ASTEC V2.2 validation by comparison to experimental data. After a reminder of the ASTEC validation strategy, the ASTEC V2.2 validation matrix is depicted, including more than 300 experimental tests conducted at various scales in more than 50 different facilities worldwide. Then some V2.2 results are discussed for a few representative applications. These calculation examples are selected in a way to cover diverse aspects of severe accident phenomenology in order to provide a good picture of the ASTEC V2.2 modelling status for both in-vessel and ex-vessel processes. Finally, the main lessons drawn from this quite large validation task are summarized, along with an evaluation of the current physical modelling relevance and how it relates to the current state-of-the-art. Based on those outcomes, the ASTEC V2.2 validity domain is specified and some prospects for further improvements are put forward.
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GEOZS, IJS, IMTLJ, KILJ, KISLJ, NLZOH, NUK, OILJ, PNG, SAZU, SBCE, SBJE, UILJ, UL, UM, UPUK, ZAGLJ, ZRSKP
Transporting blended hydrogen natural gas through existing natural gas pipeline networks is an important strategy for meeting the growing demand for hydrogen energy. However, gas leakage poses a ...serious safety concern. This study conducts a large-scale simulation of gas leakage and explosion accidents in a hydrogen-doped natural gas station and aims to evaluate the impact of pipeline pressure and leakage direction on the accident consequences. Simulation results indicated that the high pipeline pressure led to more accumulated flammable gas, which caused severer explosion disasters. Within the 100s simulation period, the cumulative volume of the flammable gas cloud reached 2329.80 m3 under 4 MPa pipeline pressure, and the peak explosion overpressure increased by 69.64% compared with the 1 MPa case. Leakage directions also significantly affect the evolution of accidents. Suppose the leakage direction is towards living areas, flammable gas is more likely to accumulate in confined spaces up to 2363.20 m3. The gas explosion will severely damage the building structure with a 17 kPa explosion overpressure and 7 kPa shock wave intensity. Interaction with wind and the disturbance of obstacles also contributed to the flammable gas cloud accumulation, which increased peak values of explosive overpressure.
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•Large-scale simulation of accidents at a hydrogen-doped natural gas station.•Analysis of flammable gas dispersion after leakage under various conditions.•Prediction and assessment of potential gas explosion consequences.•Evaluation of dangerous areas generated during the accidents.
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GEOZS, IJS, IMTLJ, KILJ, KISLJ, NLZOH, NUK, OILJ, PNG, SAZU, SBCE, SBJE, UILJ, UL, UM, UPUK, ZAGLJ, ZRSKP
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