Accident Conditions Research Articles

Mist-control of polyalphaolefin (PAO) lubricants using long pairwise end-associative polymers

Accidental release of pressurized hydrocarbon fuels and lubricants are a major fire hazard due to the formation of small droplet mists that can readily evaporate and ignite. Mist control through increasing droplet size and suppressing droplets has been previously demonstrated with high molecular weight polymer additives, but traditional long polymer additives do not survive the pumping that would usually precede accidental release. This constraint inspired associative polymer additives that can transiently form the high molecular weights needed for mist control, while reversibly breaking during pumping. A prior study demonstrated the efficacy of such a system in fuel: long telechelic polycyclooctadiene (PCOD) with pairwise associating acid and base end-groups. Here, we address an obstacle to applying this same polymeric system in a polyalphaolefin (PAO) solvent—its poorer solvent quality for PCOD than fuel. We measured the effects of the end-associative PCOD compared to a non-associative control on the rheological properties of solutions in both PAO (a common lubricant and heat transfer fluid) and decahydronapthalene (decalin, a solvent with PCOD solubility similar to fuel) in shear and extension, and connect those rheological modifications to observed changes in PAO spray under simulated accidental release conditions. The PCOD additives demonstrated substantial mist control in PAO, both in terms of reduced spray angle and droplet suppression. Despite the worse solubility in PAO and thus smaller effective coil size, these associative PCOD additives are effective at the low concentrations (¡0.1 wt %) necessary for practical use as a safety measure.

Effects of boric acid on volatile tellurium in severe accident conditions

Boric acid is used in light-water nuclear reactors to control the reactor and is expected to be present as part of the chemistry of a severe accident. Therefore, its influence on other prominent species expected in an accident must be investigated. One such species is tellurium. In the present study, tellurium is volatized, and boric acid is dissolved and injected into the system as a means of studying the interaction between it and tellurium. The experiments were evaluated with ICP-MS and XPS. Results suggest that while there is no direct interaction, boric acid still affects the tendency for tellurium to oxidize. In general, less oxidation was detected in the presence of boric acid than in its absence, especially at high temperatures. The species formed upon oxidation was determined to be TeO2. Since tellurium metal is more volatile than TeO2, this may have implication in a wider severe accident context.

On the behavior of a concrete-based dry cask for spent nuclear fuel in off-normal and accidental conditions

The safe storage of Spent Nuclear Fuel (SNF) within a dry cask system must be guaranteed for any operating condition. In particular, the Peak Cladding Temperature (PCT) is currently employed as a measure of the fuel integrity in normal conditions. The same fuel integrity has to be ensured also in off-normal and accidental conditions by keeping the PCT below the regulatory limits. In this regard, the work proposed in this paper aims at evaluating the performance of a concrete-based dry cask in off-normal and accidental conditions. More specifically, the influence of a partial to complete blockage of the air inlet openings on the PCT is assessed by means of a MELCOR model of the HI-STORM 100S cask. Attention is also paid to the temperature of the external concrete cask in order to check the integrity of the concrete itself.

Multiphysics analysis of a metal hydride moderated megawatt heat pipe reactor with burnable poisons

With the development of nuclear energy, microreactors have received increasing interest among researchers in recent years. In this paper, a megawatt heat pipe reactor with metal hydride moderators and burnable poisons is proposed. The hydrogen stability of the reactor under accident conditions, including reactivity insertion accidents, loss of power conversion unit heat sink accidents, and heat pipe failure accidents are analyzed. In this work, Gd2O3 is introduced as a burnable poison in the form of mixing with the UO2 fuel. According to the results of the burnable poison design, the 0.1% mass fraction of Gd2O3 is selected as the burnable poison loaded in the core. Safety analysis indicates that the introduction of burnable poisons can be beneficial during the positive reactivity insertion accident as it can reduce the excessive reactivity at BOL, thus reducing power peak and preventing hydrogen dissociation in ZrH1.4 rods. However, during the loss of PCU heat sink accident, ZrH1.4 rods will dissociate regardless of the presence of burnable poisons, whereas YH1.7 rods show better hydrogen stability. Moreover, in the event of the heat pipe failure accident, ZrH1.4 rods are more susceptible to dissociation than YH1.7 rods. As a result, the YH1.7+BP core is a better choice compared with other designs proposed in this paper as it provides a relatively high temperature margin.

Reaction Capsule Design for Interaction of Heavy Liquid Metal Coolant, Fuel Cladding, and Simulated JOG Phase at Accident Conditions

High temperature corrosion of fuel cladding material (15-15Ti) in high burn-up situations has been an important topic for molten metal-cooled Gen-IV reactors. The present study aims to investigate the simultaneous impact of liquid lead (coolant side) and cesium molybdate (fuel side) on the cladding tube material. A capsule was designed and built for experiments between 600 °C and 1000 °C. In order to simulate a cladding breach scenario, a notch design on the cladding tube was investigated pre- and postexposure. Material thinning by corrosion and leaching at temperatures ≥ 900 °C caused breaches at the notches after 168 h exposure. The temperature dependent cladding thinning phenomenon was used for kinetic interpretation. As the first of a two-part study, this paper will focus on the exposure capsule performance, including metallographic cross-section preparation and preliminary results on the interface chemistry.

Investigation of Heat Exchanger Performance in The Heating Tank Section of Loop FASSIP 03 NT

The Passive System Simulation Facility (FASSIP) loop is an experimental test facility for a passive cooling system to recover the residual heat from decay produced by the reactor core during accident conditions. The Heating Tank Section (HTS) is one of the components of the FASSIP 03 NT facility. This component is equipped with a heat exchanger, 3 types of heat exchangers can be applied to HTS. Namely heat exchangers of the Straight Pipe Heat Exchanger (SPHE) type, the Straight Pipe Fins Heat Exchanger (SPFHE) type, and the Helical Pipe Heat Exchanger (HPHE) type. A modification was made to increase the efficiency of HTS, namely replacing the electric heater on the HTS from a ceramic band heater type to an immersion heater type. With this modification, it is necessary to know the performance of the heat exchanger on HTS and its speed in reaching operational temperature. The HPHE-type heat exchanger is more efficient than the SPHE-type and SPFHE-type heat exchangers. The HPHE-type heat exchanger has a much larger length of 5.5 m, so the thermal resistance (Rth) is very small, namely 0.003926 ℃/W. To reach the working fluid temperature in the range of 50 – 90 ℃, the HPHE-type heat exchanger requires 35 – 86 minutes.

Application of Level 3 Probabilistic Safety Analysis in UK HPR1000

Abstract A Level 3 probabilistic safety assessment (L3 PSA) is required in United Kingdom (UK) generic design assessment (GDA) to demonstrate that a new nuclear power plant is suitable to be built in UK. L3 PSA is used to assess the individual and societal risk and compare the results against the offsite radiation protection targets (RPTs) for fault and accident conditions. Its results can be used for risk evaluation and environmental impact assessment and provide useful information for alternative design features, rulemaking, and regulatory procedures. It takes into account atmospheric dispersion, demography, dosimetry, pathways to man, and plant/site characteristics. The radioactive source terms and their frequencies often are passed on from Level 1/2 PSA analyses. The framework for demonstrating the acceptability of the new plant design at GDA is presented for the assessments against RPTs 7, 8, and 9. There is little relevant good practice and mature standard for L3 PSAs that have recently been implemented worldwide. In this study, a pilot L3 PSA is performed for UK Hua-long Pressurized Water Reactor (UK HPR1000) to reflect the UK context and relevant good practices. It introduces the methodology and the processes to be followed to perform conditional consequence calculations for the faults and accident scenarios. All radiation sources are considered and analyzed. The radiological risks to a potential UK site are analyzed and compared against RPTs. A widely used code - PC COSYMA, is selected as a primary tool for the accident consequence calculations. The strengths and limitations of the code are identified based on the project situation, and either qualitative arguments or supplementary analyses are subsequently proposed to overcome the limitations. The final L3 PSA results are derived to support the demonstration that the offsite radiological risks for UK HPR1000 have been achieved as low as reasonably practicable (ALARP) and have met the UK regulatory expectation.

CFD Modeling and Analysis of a Heat-Generating Melt Pool Under External Vessel Cooling and Top Flooding

In some nuclear reactors, under accidental conditions, core debris forms a molten pool, which is later located in a core catcher. The core catcher proposed by the authors uses special refractory material to absorb enthalpy of corium so that temperatures are within 1500 K, which is possible to cool with side cooling and top flooding. Since performing a full-scale prototypic experiment is extremely challenging and complex because of the involvement of very high temperatures and the presence of radioactive materials, it is important to develop a Computational Fluid Dynamics (CFD) model capable of simulating coolability of the melt pool with the above cooling strategy. In the present work, a CFD model was developed for the above purpose and was benchmarked with experiments conducted under simulated conditions by the authors. The experiment involved the melting of about 25 L of sodium borosilicate glass at about 1473 K and cooling it in a scaled-down core catcher model. In the presence of decay heat inside the melt pool, turbulent natural convection plays an important role in the temperature distribution inside the melt pool and on the vessel walls. For this, we used different turbulence models. Comparisons among the Standard k-ε, Shear Stress Transport (SST) k-ω, and two-dimensional (2D) Large Eddy Simulation (LES) turbulence models show that SST k-ω and 2D LES turbulences are found to be in good agreement with the experimental results for the temperature distribution in the melt pool, and SST k-ω is found to be computationally less expensive than 2D LES. In general, the CFD model is capable of simulating heat transfer with phase changes inside the heat-generating melt pool. In view of this, the model can be further extended to include cooling of the melt pool in the prototype core catcher. The evolution of crust formation has been investigated in detail using a CFD model.

Comparison of accelerated and non-accelerated post-LOCA simulation

Cables important to safety used in nuclear power plants are required to withstand environmental, operational and system conditions to which they are exposed during their design life. This may include harsh conditions resulting from a design basis event (DBE), such as a Loss-of-Coolant Accident (LOCA). The testing procedure includes accelerated thermal and ionizing radiation ageing which brings the tested samples to their end-of-life condition. Accelerated pre-ageing is then followed by accident simulation (LOCA test). The latter part of accident simulation (post-LOCA), which may take up to 1 year, is commonly accelerated based on the Arrhenius approach, however the adequacy of this approach remains a topic of ongoing discussion, lacking a clear scientific or engineering consensus. While numerous works have focused on testing various conditions during accidental and post-accidental conditions in nuclear power plants, there is notable absence of direct assessment of acceleration effects.The aim of this work was to assess the effects of accelerating post-accident conditions on electrical cables using non-destructive electrical measurements. The cables were subjected to thermal and radiation pre-ageing, LOCA test, and then were divided into two groups for post-LOCA tests: one group was tested for one year in non-accelerated post-accident conditions, while the other was accelerated using the Arrhenius approach. Diagnostic methods, including insulation resistance, capacitance, voltage withstand tests, and high-frequency parameters such as attenuation, characteristic impedance, and relative permittivity, were employed for comparing sample conditions. Results from both accelerated and non-accelerated post-LOCA simulations were consistent, indicating that the acceleration was adequate.

Contribution of IAEA coordinated research projects to light water reactors advanced technology fuel testing and simulation

The Fukushima-Daiichi accident in 2011 highlighted the pressing need for enhanced design and safety analysis of nuclear fuels, especially under accident conditions in nuclear power plants (NPPs). To address this need, the international nuclear fuel community has developed several concepts of Accident Tolerant and Advanced Technology Fuels (ATFs) for light water reactors. The International Atomic Energy Agency (IAEA), at the request of its Member States, has initiated a series of three Coordinated Research Projects (CRPs) to aid in the development and testing of these ATFs. These projects, known as FUMAC, ACTOF, and ATF-TS, focus on the experimentation and simulation of ATFs under various accident conditions, including design basis accidents and design extension conditions. This paper provides a comprehensive overview of the main activities, as well as the achieved or anticipated results of these three CRPs. The insights and advancements gained from these projects will enable IAEA Member States to further refine nuclear fission technology, thereby contributing to efforts aimed at mitigating climate change.

The effect of nanometre-scale kinetic competition on the phase selection in Zr/Si superstructure

Nano structured coating system demonstrates outstanding performance stemming from unique structure evolution. Zr/Si coating was proposed as an ideal candidate for the protection of Zr alloy cladding. However, the phase selection remains a challenge to coordinate the complex requirement under the service and accident conditions of reactor core. In this work, a phase selection mechanism for amorphous Zr-Si-O (abbr. ZSO) was elucidated by the competitive consumption of Zr layer by silicidation and oxidation, which was also scale-dependent. The energy-loss near-edge fine structure and nanoindentation were also employed to characterize the unique scale effect in the two processes. Both the oxides kinetics and elastic modulus gain indicated an immediate passivation process under the subcritical condition after the in situ phase selection of the amorphous ZSO. Ab initio molecular dynamics (AIMD) was performed to disclose the origination of depressed O migration in ZSO. The multilayer structure also shows accident tolerance potential. The interfacial phase selection provided a strategy to regulate mesoscale structure originating from nanomter interface and meets the diversified requirement for “structure-performance relationship” under the service and accident conditions.

Thermal-Mechanical performance analysis of a petal-shaped fuel rod under RIA condition

This paper establishes the performance analysis model of a petal-shaped fuel rod (PSFR) under Reactivity Initiated Accident (RIA) condition based on thermal elastic–plastic theory and finite element method. The results show that the temperature changes of fuel and cladding are basically synchronized, and the temperature extremes appear in the central region, with the distribution characterized by a large middle and two small sides. The stress is concentrated in the center region of the cladding. At the beginning of operation, the stress at the lobe root of cladding is the largest. With the increase of operation time, the stress concentration at the lobe tip is gradually obvious. At 150th days of operation, a sudden RIA occurred and the power factor of the PSFR rapidly increased over 5.38 times and the cladding stress exceeded the yield strength. The axial and circumferential center cladding lobe roots and nearby may be at greater risk of plastic deformation. These results will provide an important theoretical basis for the determination of the deformation and accident operation risk of PSFRs.

In-situ determination of strain during transient burst testing and the temperature dependence of Zircaloy-4 claddings

Understanding fuel system behavior during postulated loss-of-coolant accidents is pertinent for continued safe and efficient operation of light water reactors, particularly as higher burnups are being pursued and safety margins re-evaluated. Conventional mechanical models for the incumbent Zr alloys typically rely on the assumption that steady-state creep is the dominant fuel cladding response during transient accident conditions. To investigate this assumption, simulated accident burst testing was performed on Zircaloy-4 claddings with balloon behavior measured in-situ. Two distinct loading conditions were utilized during burst testing: (1) constant-gas-inventory where pressure was allowed to increase with temperature and (2) constant pressure. In-situ strains and strain rates were measured via 2-dimensional digital image correlation techniques and synchronized with temperature to determine deformation dependencies. The temperature dependence of strain rate was characterized by a two segment Arrhenius relationship, with a distinct transition between the high and low temperature/strain regimes. The average activation energy of the lower temperature/strain regime was 328 ± 25 kJ/mol, in agreement with the ∼320 kJ/mol used for conventional LOCA models. However, the higher temperature/strain segment, which encompassed most of ballooning, showed increased activation energies as well as a dependence on whether the burst region was in view. For tests that burst away from the camera view, the average high temperature/strain segment activation energy was 635 ± 150 kJ/mol. For samples where the rupture opening formed in view, the average activation energy was 1015 ± 179 kJ/mol. This observed shift in temperature dependence indicates a transition in deformation mechanism at the end of life, possibly to time independent failure mechanisms, which has not yet been visualized in the literature for Zr alloys. Parameters at the transition points were analyzed to determine thresholds for this change in behavior, which occurred at an average hoop strain of 6.9 ± 2.1 %.

Dynamic Analysis of Three Alternative Mooring Systems for a Semi-Submersible Wind Turbine in Intermediate Water Depth

Abstract Three forms of the mooring system in 60 m water depth are proposed for semi-submersible with partially inclined columns (SPIC) concept floating wind turbine (FWT). One is a simple form with only catenary lines, and the other two are hybrid forms including clump weights. The clumps are attached to the suspended section for Hybrid form1 and the bottom section for Hybrid form2. Hybrid form2 achieves the smallest line length and chain weight. Three alternative proposals can be evaluated through mooring line characteristics, dynamic responses, utilization factors, and simple cost analysis. Hybrid form2 allows for smallest pretension, and largest stiffness and nonlinearity only at large offsets. Under operational conditions, the mean surge for Hybrid form1 and Hybrid form2 is similar, but the fairlead tension is significantly smaller for Hybrid form2. Under the survival condition, the clumps of Hybrid form2 are lifted up and put down, leading to small mean offsets of FWT but large wave-frequency components of line tension. Among the three forms of the mooring system, the Hybrid form2 can limit the FWT to the smallest offset range while also controlling the mean mooring line tension to a level similar to the other two forms. Under normal working conditions and accidental conditions with single line broken, the maximal surge motions of FWT under the restraint of three mooring systems all meet the design requirements. The mooring line strength of the three mooring systems meets the requirements in ultimate limit state (ULS) and accidental limit state (ALS) analyses. Among them, the utilization coefficient of Hybrid form2 is closest to 1, demonstrating its best economic performance.

Blowout Fluid Velocity Measurement Method based on High-speed Sequence Images

Given the high-temperature and high-pressure environment after the blowout, the staff usually cannot calculate and analyze various blowout parameters at close range and thus take effective measures to control accidents quickly. In this paper, a blowout fluid velocity measurement method is proposed based on high-speed sequence images, and simulation experiments are conducted. The scene conditions of blowout accidents are simulated and analyzed. On this basis, cameras are adopted to shoot a large number of simulated blowout fluid videos from different angles under outdoor conditions. Afterward, the video is converted into a multi-frame image, and the blowout fluid image feature point detection and extraction model are established based on the orb (oriented fast and rotated brief) algorithm. On the feature points extracted by the model, the gms algorithm is used to match the features of the blowout fluid images in the front and back frames. Additionally, the wrong matching points are eliminated using the ransac (random sample consensus) algorithm, and the obtained feature point pairs are employed into the flow velocity calculation model for flow velocity calculation. The results demonstrate that the measured fluid velocity is more accurate. The method proposed in this study can provide reliable data support for early blowout emergency rescue.

Impact of LWR assembly structural features on cladding burst behavior under LOCA conditions

This provides an initial scoping study on clad balloon and burst behavior for burnup extension of reactor fuel. The associated issues with burnup extension are fuel fragmentation, relocation, and dispersal in the event of cladding failure. The general finding of this work is that the structural features, spacer grids and mixing vanes, locally suppress cladding deformation but have little impact on the overall clad performance during loss-of-coolant accidents. The work detailed in a previous report by Capps et al. focused on core optimization via neutronics, thermal hydraulics and thermomechanical analysis for burnups beyond 62 GWD/tU and enrichments above 5%. Uncertainty of fuel fragmentation relocation and dispersal in high-burnup rods during accident conditions was also investigated. The dispersal aspect of fuel fragmentation depends on cladding rupture. Thus, assessing uncertainties in the rupture behavior is helpful in estimating the dispersal of high-burnup fuel. This study builds on the previous work by assessing the impact of assembly structural features on cladding balloon and burst behavior in a full-length fuel rod.In this work, the BISON fuel performance code was used to generate 2D radial and height meshes containing structural features commonly used in nuclear fuel assemblies. First, meshes were generated with spacer grids. Results were then compared to the cladding burst temperature and balloon strain results from the previous work. A mesh sensitivity study was performed to ensure that mixing vanes and spacer grid effects were appropriately considered, resulting in a more refined mesh than the previous study. The balloon deformation and burst times of the cladding were compared to the original case. Consideration was also given to the effect of rod initial pressure. 3D quarter rod simulations were also performed and found good agreement with the 2D simulations in clad deformation and reasonable agreement in burst times.

Analysis of Heat Flux Quenching of a SS-316L Rod Using Nanofluids

Quenching is one of the rapid cooling processes; however, the jet impingement technique is an efficient method for cooling in constrained space. In the nuclear reactor, to prevent the loss of coolant accident condition, de-ionized water is stored in the repository of the emergency core cooling system. Nanofluids exhibit better heat removal capacity, thus, the present article investigates experimentally an optimum oxide based nanofluid under various parameters such as set wall temperature, mass flow rate, and concentration. Quenching curves with different working fluids at a particular heat flux condition have been recorded at different locations of the heater surface. In case of heat flux quenching, limiting heat flux is defined as the boundary heat flux after which the quench front is not able to cool the heater surface. Quench front velocity calculation is the criteria to decide the limiting case. In case of nanofluids, it was observed that there was enhancement in the value of limiting heat flux under various parameters as compared to de-ionized water.

Hybrid Heat Pipe Shutdown Rod as a Novel Concept of Passive Safety System for Microreactor

Ensuring the structural integrity of the monolithic core, which houses critical components such as heat pipes and fuel rods, is crucial for the safety of a microreactor. This research introduces a hybrid heat pipe shutdown rod as a novel passive safety system to address potential temperature rises in the monolithic core during accidents. It is designed to perform a dual function under accident conditions. It simultaneously absorbs neutrons and removes heat through structural modifications to the existing shutdown rod. Specifically, this system provides a heat transfer path within the monolithic core without increasing the size of the microreactor. By selecting cesium as the working fluid, we aimed to achieve rapid operation of the heat pipe to quickly reduce the temperature gradient in the monolithic core under accident conditions. The hybrid heat pipe was fabricated and evaluated and found to develop continuous flow even at temperatures around 205.1°C. However, its unique structure causes a pronounced converging–diverging effect, resulting in a temperature drop from about 70–170°C in the evaporator region, followed by a slight recovery to below 50°C in the condenser. This effect arises from changes in the cesium vapor mass flow rate due to phase changes and variations in the flow area between the evaporator and condenser. Despite these effects, the use of liquid cesium as the working fluid enables the hybrid heat pipe to operate under natural convection, avoiding startup problems that could cause flow blockage.

Sensitivity and uncertainty analysis in predicting hygroscopic growth of aerosols for atmospheric and nuclear relevance

Sensitivity and uncertainty analysis serves as a tool for modeling physical phenomenon and evaluating the model validity. It finds applications in simulating the dynamical behavior of aerosols in different contexts. Hygroscopic growth of aerosol particles is one such domain where it can optimize the accuracy of model output greatly. Accurate prediction of hygroscopic growth characteristics depends on the intrinsic parameters and the size-dependent evolution of growth factors. In reactor accident scenarios, growth characteristics of fission/activation product aerosols play a significant role in determining the airborne or deposited fraction. Köhler theory is fundamentally used to predict the growth behavior of aerosols with a detailed physical explanation of critical size and supersaturation characteristics. Alternative approaches including modified Köhler theories and semi-empirical models rely on the linkage of physical and chemical composition-based parameters. The variation of involved parameters and the choice of working range influence the outcome of the theory greatly. This study is focused on sensitivity analysis of the modified Köhler theory given by Brechtel and Kreidenweis and the evaluation of sensitivity and associated uncertainty in the parameters. Monte Carlo approach has been employed, and sensitivity indices were obtained. Identified sensitive parameters were dry particle diameter, solute density, molecular weight, and chemical composition dependent factor (b0). These parameters require careful attention while designing experimental measurements and/or preprocessing model inputs. All sensitive parameters were subjected to uncertainty analysis and the overall output of the aerosol growth code thus obtained has been observed to follow log-normal law with median size at 335.2 nm with a geometric standard deviation of 1.54. This finds applications in the case of studies aiming to predict the growth of aerosols in the reactor component systems during accident conditions. In such conditions, there lies no choice of selection of input parameters and one cannot be sure about their variation range. This provides a conservative estimate for such a random scenario as uncertainty analysis performed in this study covers all the random combinations of input parameters and the overall most probable grown droplet size is estimated.

Numerical simulation of indoor air distribution and wall heat storage under accident conditiohs in a main control room

The main control room serves as a refuge for staff in a nuclear power plant in the event of an accident. To determine whether the control room can meet habitability requirements under accident conditions, this study utilized CFD methods to establish a habitability model of the main control room under such conditions. The changing trends of temperature and velocity near personnel were analyzed. The numerical results indicate that the building envelope significantly contributes to reducing the indoor temperature. The average wall temperature measured at 72 hours was 27?C, with an average temperature increase of 4?C. The highest average ambient temperature in the main control room occurred at 24 hours. At this time, the indoor ambient temperature reached approximately 34.62?C, with a temperature increase of 11.62?C, while the wind speed near the personnel remained below 0.2 m/s, meeting habitability requirements for personnel under extreme conditions.