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The De Gruyter Series in Mathematics and Life Sciences is devoted to the publication of monographs in the field. They cover topics and methods in fields of current interest that use mathematical approaches to understand and explain, model and influence phenomena in all areas of life sciences. This includes, among others, theory and application of biological mathematical modeling, complex systems biology, bioinformatics, computational biomodeling stochastic modeling, biostatistics, computational evolutionary biology, comparative genomics, or structural bioinformatics. Also, new types of mathematical problems that arise from biological knowledge shall be covered.

A bstract Charge Coupled Devices (CCD) are being used for reactor neutrino experiments and have already demonstrated their potential in constraining new physics models. The prospect of a Skipper-CCD experiment looking for standard and beyond standard model (BSM) physics in a nuclear reactor has been evaluated for different benchmark scenarios. Here, we report the first installation of a 2-g Skipper-CCD inside the containment building of a 2 GW th nuclear power plant and analyze its performance throughout its first 18 months of operation. The sensor was successfully deployed at Atucha II, in Argentina, 12 meters away from the center of the reactor core. We discuss the challenges involved in the commissioning of the detector and present data acquired during reactor ON and reactor OFF periods, with the sensor functioning with a sub-electron readout noise of 0.17 e − . Based on an exposure of 56.8 g day reactor ON and two reactor OFF data sets with a total exposure of 118.1 g day we characterize the system and evaluate the sensitivity to CEvNS. We achieved a background rate of 33 kdru and a low threshold of 45 eV ee . The ongoing efforts to improve sensitivities to CEvNS and BSM interaction are also discussed.

Here we report the results of multiple analytical techniques on sub-mm particulate material derived from Unit 1 of the Fukushima Daiichi Nuclear Power Plant to provide a better understanding of the events that occurred and the environmental legacy. Through combined x-ray fluorescence and absorption contrast micro-focused x-ray tomography, entrapped U particulate are observed to exist around the exterior circumference of the highly porous Si-based particle. Further synchrotron radiation analysis of a number of these entrapped particles shows them to exist as UO 2 —identical to reactor fuel, with confirmation of their nuclear origin shown via mass spectrometry analysis. While unlikely to represent an environmental or health hazard, such assertions would likely change should break-up of the Si-containing bulk particle occur. However, more important to the long-term decommissioning of the reactors at the FDNPP (and environmental clean-upon), is the knowledge that core integrity of reactor Unit 1 was compromised with nuclear material existing outside of the reactors primary containment. The larger particulates from reactor Unit 1 of the Fukushima Daiichi Nuclear Power Plant have received sparse attention compared to the Unit 2 particulate. Here the authors perform the higher-resolution and 3-dimentional analysis of several high-density micron-scale fragments, from within a larger Unit 1-derived representative ejecta particle.

The integrity of containment buildings in nuclear power plants is crucial for preventing the release of radioactive materials during severe accidents. This study investigates the effect of the uncertainty in temperature-dependent strength of concrete on the internal pressure capacity of prestressed concrete containment vessel (PCCV). To this end, a high-fidelity finite element model is developed and the uncertainty in concrete material properties due to temperature variations is taken into account for a finite element analysis with internal pressure and temperature histories. In addition, two limit states of PCCV, such as onset of leakage and functional failure, are defined to investigate internal pressure capacity depending on the different damages to PCCV. The results provide insights into the behavior of PCCV under severe accident conditions and the impact of the uncertainty in the concrete material due to temperature on their performance, in terms of the leakage of PCCV. This research enhances the understanding of PCCV’s response to internal pressure and temperature and contributes to the safety assessment of nuclear power plants.

The loss of coolant accident in the primary pipeline of a reactor is a typical accident in the safety analysis of nuclear power plants, and it is one of the accidents with a high probability of occurrence. Due to the extremely complex structure of the lower space of the containment vessel, it increases the complexity of fluid flow during a loss of coolant accident. This article is based on the internal water flooding analysis simulation software CNIFA of a nuclear power plant to simulate the water flooding of a rupture in the main pipeline heat pipe section of a power plant, and to conduct a comprehensive water flooding scenario simulation for this working condition. The quantitative information collected during the simulation process, including dynamically displayed water volume, water level, and their changes over time, facilitate emergency response personnel to choose the correct response plan in advance, it can also provide more accurate basis for various specialties in design analysis.

The new designs of Nuclear Power Plants are increasingly emphasizing the implementation of passive systems to fulfill several safety functions. Such systems operate with low driving forces, and often employ components that require a “change state” signal for actuation. These features may lead to complete or partial actuation failures, altering the driving force conditions and consequently reducing the operational ranges and reliability of the systems themselves. However, even though the operational modes differ from those of active systems, the safety standards remain equivalent. To study these possible situations, the computer system codes must be able to simulate the phenomena related to the specific system, and they should be assessed using data both from separate and integral experimental facilities. Within the framework of the EU2020 PASTELS (Passive Systems: Simulating the Thermal hydraulics with Experimental Studies) project, a series of natural circulation experiments have been conducted in the PASI facility, an open loop Containment Wall Condenser (CWC), varying initial and boundary conditions to support the validation of system codes. As project partner, ENEA developed the PASI input deck for RELAP5/mod3.3 to assess the code’s capability in simulating complex loop components, such as the sparger, and related thermohydraulic phenomena like water flashing and geysering within the open loop. This paper presents the post-test calculations conducted with RELAP5 for test 2 (quasi-static test) and test 6 (replicating the drain phase of the water storage tank).

Probabilistic evaluation of the internal pressure capacity before excessive leakage is essential for ensuring preemptive response in case of a severe accident in a prestressed concrete containment vessel (PCCV) subjected to internal pressure in a nuclear power plant. Therefore, in this study, we conducted a fragility analysis to evaluate the probabilistic internal pressure capacity for damage states before excessive leakage. Three damage states were classified, and a finite element (FE) model was developed based on the experimental test results of a 1 : 4 scale PCCV under the overpressurization condition at Sandia National Laboratory. The three damage states were determined as the onset of inelastic behavior (DS1), onset of leakage (DS2), and excessive leakage (DS3). Fragility analysis was performed using a validated FE model to consider uncertainties in material properties and prestressing force. The Latin hypercube sampling method was used to perform uncertainty quantification, and normality tests verified that the samples exhibited normal distributions. The internal pressure capacity variability in DS1 and DS2 was higher than that in DS3. DS1 is the point at which the compressive deformation caused by the prestressing force was converted into tensile deformation because of the increase in the internal pressure, which resulted in tensile concrete cracking. Furthermore, DS2 is the state in which localized damage of the liner occurred and cracks in the concrete spread. Thus, the internal pressure capacity for DS1 and DS2 exhibited high variability because of the uncertainty of the prestressing force and material properties. Moreover, the median capabilities of DS1, DS2, and DS3 were ~1.85, 2.61, and 3.45 times the design pressure, respectively. Therefore, material and structural uncertainties in nuclear power plants should be appropriately considered, and a conservative assessment of the damage state before excessive leakage should be performed for preemptive response in the case of severe accidents.

The paper provides a detailed overview of the development of a containment pressure and temperature (P/T) analysis methodology for advanced power reactor 1400 (APR1400) nuclear power plants (NPPs). The study addresses the restrictions on exporting independent nuclear power plants by utilizing the containment analysis package (CAP) computer code developed in Korea. One of the key aspects highlighted in the paper is the comparison of results obtained from the CAP code with those from the CONTEMPT-LT/028 code which is used P/T analysis. The analysis focuses on two types of accidents: loss of coolant accidents (LOCA) and main steam line break (MSLB) accidents, which are considered as design basis accidents. By comparing the outcomes of both codes, the paper evaluates the performance and effectiveness of the CAP code in predicting the P/T behavior within the containment during these accidents. The paper also discusses the criteria and technical standards for the containment P/T analysis. It emphasizes the importance of ensuring that the peak P/T remain within the safety related systems, equipment, and structures of NPP. The design pressure is identified as a critical factor in achieving these objectives. In conclusion, the study presents the successful development of a containment P/T analysis methodology using the CAP computer code for APR1400. The methodology considers the specific characteristics of Korean NPPs and new code, CAP. The paper emphasizes the applicability and effectiveness of the CAP code in this context. However, further research and validation efforts are recommended to enhance the accuracy and reliability of the methodology for various design basis accidents. The developed methodology is expected to contribute to the safe and efficient operation of APR1400 NPPs and support Korea’s ambitions in exporting NPPs’ technology to other countries.

Frequently, cracks are found in welded joints of the insert plate assembly of a nuclear reactor containment vessel. To study the residual stress state of the dissimilar steel welded joint of the insert plate assembly and ensure that there are no defects after welding, the ABAQUS finite element analysis software was used to simulate the temperature field and the post-weld residual stress field of the dissimilar steel inserted into the plate assembly in the containment of the nuclear reactor. The insert plate and sleeve material are SA738GR.B and A350, respectively; the heat treatment process after welding is simulated to predict the residual stress state of the welded joint after heat treatment. The simulation results show that the weld zone is the most sensitive to the circumferential cracks, and the maximum residual stress appears near the weld toe near the tube side.

Safety is always the primary concern for designing and analyzing nuclear reactor systems. The requirements for the safety margin for advanced small modular reactor (SMR) systems are targeted even higher than the conventional commercial large-scale nuclear reactors incorporating the passive and inherent safety systems. The SMR systems are designed with the condensation passive containment cooling system (PCCS), which plays a critical role in removing reactor heat during a steam release accident case. However, the presence of non-condensable gas (NCG), like air, reduces the heat transfer performance. This physics phenomenon becomes multifactorial for nuclear reactor containment during a fuel failure accident case that releases hydrogen gas. Besides, the mixture component of steam-air-hydrogen varies in reactor accident cases, which needs simulation and validation keeping parameters of importance. Reviews showed that previous studies for SMR’s PCCS did not cover the condensation heat transfer (CHT) in the presence of multicomponent NCG mixture parametric computational fluid dynamics (CFD) simulation and validation, making a research gap in the SMR design safety. A comprehensive CHT parametric CFD study was performed for SMR PCCS to fill this research gap. This study used experimental data as simulation 3D physics domain inlet and outlet boundary conditions. However, the wall boundary conditions were constant temperature, curve-fit, and annular coolant for verifying the turbulence models. Parametric simulations were performed, verified, and optimized for steam-NCG mixtures. The multicomponent gases, multiphase mixtures, and fluid film condensation models were applied with associated turbulence models. The results of the parametric study were evaluated for realistic reactor conditions. Results showed that parametric study provided critical insight about the dependency of multicomponent gas mixture parameters that supports reactor safety design, analysis, and licensing.