Accident Conditions Research Articles

An integrated assessment method of real-time source term for high temperature gas-cooled reactor

To address high temperature gas-cooled reactor (HTR) nuclear power plant (NPP) safety analysis and emergency response, this paper proposes an integrated source term assessment (ISTA) method, guided by the principles of reality, rapidity and rationality. The fundamental concept of ISTA is to establish a connection between the theoretical model of source term (ST) calculation and actual conditions of HTR NPPs. According to the ISTA method, STs can be evaluated by mapping key parameters through monitoring variables during any abnormal operational conditions or accident conditions, resulting in a more realistic representation. Furthermore, the output of STs in various typical spaces within an HTR NPP further expands the application scope beyond the development of nuclear emergency protection strategies. In addition, a corresponding tool (ISTAT) for engineering application is developed and the case is given to verify the feasibility. This research highlights the technological advancement affecting radiological release scenarios and avoiding unnecessary conservatism, which also can give reference for other advanced reactor NPPs.

Speciation and mobility of uranium isotopes in the Shu River: impacts for river to sea transfer

Abstract Fresh water outflows to sea are a source of marine contamination by radionuclides, most originating from nuclear industry operations. Usually, these lead to small amounts of radionuclides entering the sea, for example, because of the discharge of cooling water. However, under accident conditions large amounts of radionuclides may enter river systems. The extent of radionuclide transport along freshwater systems and their subsequent dispersal in the ocean depends upon their speciation. This manuscript examines the speciation of uranium, with particular reference to sulphate complexes, along the Shu River in Central Asia with a view to predict its transport along such rivers to receiving seas. The speciation of uranium isotopes in the Shu River (at the border of Kazakhstan and Kyrgyzstan) was determined. Calculations were based on the measured concentrations of Ca2+, Mg2+, Na++K+, HCO3 −, Cl−, SO4 2−, and of uranium isotopes. These calculations were determined by either titrimetric methods, gravimetric methods or, for uranium isotopes, by alpha-particle spectrometry. Almost all uranium was present as one of three soluble complexes, either [UO2OH]+, [UO2(CO3)2(H2O)2]2− or [UO2(CO3)3]4−. However, despite the significant concentration of sulphate-ions in the water, especially for the Lower Shu in Kazakhstan, the probability of uranium being present as of sulphate complexes in the analyzed water was very low. The presence of soluble ions is consistent with their mobility along freshwater systems and their transfer to the marine environment. No evidence of an effect from influxes of sulphate was identified.

Simulation Analysis of Influence of Accident Conditions on Hydraulic Reliability of Gas Pipe Network

Taking reliability as the starting point, based on Pipeline Studio software and combined with the actual situation of medium-pressure gas pipe network in Weidu District, Xuchang City, the operation conditions of gas pipe network under design conditions and accident conditions were simulated, and the accident conditions were assumed to be the shutdown of important pipe segments for reasons or the failure of the pipe segment with the maximum flow rate. According to the simulated hydraulic calculation results, the overall gas supply capacity of the pipeline network is comprehensively evaluated and analyzed, and the conclusion is reached that the pipe network can meet the requirements of normal and accident conditions by properly enlarging the pipe diameter closer to the gas source point.

Simulation of nuclear fuel clad high-temperature ballooning under loss-of-coolant accident conditions considering anisotropic creep

Zirconium (Zr) alloys are standard clad material in commercial light water reactors. However, it is vulnerable to ballooning and burst under high-temperature conditions such as LOCA (Loss-Of-Coolant Accident), due to large internal/external pressure difference. Considering that clad ballooning is detrimental in terms of nuclear reactor safety especially for high-burnup fuel rod, high-fidelity numerical simulation of Zircaloy clad ballooning is desirable. This paper presented a finite-element based implementation for the simulation of Zircaloy clad ballooning. The developed simulation capability predicts progressive clad ballooning up to failure given empirical burst limits. The material models coupled together are phase transformation and high-temperature creep. Especially, anisotropic creep was considered for α phase Zircaloy. The implemented algorithm for stress update is implicit in time integration, which allows relatively large time step size to efficiently simulate the process that can last for hundreds of seconds. An analytic benchmark, the PUZRY ballooning and burst test, and the Hardy stress-rupture test were utilized to verify and validate the simulation capability. The anisotropic creep calculation conforms to the analytical solution. The comparison with PUZRY and Hardy test data shows that considering anisotropic creep in the α phase region is essential. 3D ballooning was also simulated by considering circumferential temperature difference for both conventional and novel annular clad designs. Interestingly, it is the first time to note that the risk of buckling failure is predicted for the double-cooled annular nuclear fuel clad during the accident condition, only after considering the anisotropic creep algorithm.

Release Behavior of Gaseous Ruthenium Tetroxide During Heating of High-Level Liquid Waste Simulant During Simulated Accident Conditions

A malfunction of the cooling system of high-level liquid waste (HLLW) and the failure of countermeasures may lead to the evaporation to dryness due to the loss of cooling functions (EDLCF) of the HLLW. In the EDLCF, ruthenium (Ru) can be released at a greater fraction to the initial amount than the other elements in HLLW by forming gaseous Ru. It is important to identify the chemical form of the released gaseous Ru to achieve a comprehensive understanding of the events impacting the source term assessment of Ru in this accident, such as particle formation, gas absorption, and deposition on migration pathways. In this study, we observed the ultraviolet/visible spectroscopy of the off-gas generated during the heating of a HLLW simulant. Employing a program that allows for the separation and quantification of known components within the spectrum [ruthenium tetroxide (RuO4), nitrogen dioxide, and nitric acid], we attempted to analyze the composition of gaseous Ru within the generated off-gas. Our findings revealed RuO4 as the main component of the gaseous Ru in the off-gas after comparing the total amount of released Ru and the RuO4 released amount obtained via spectroscopic analysis.

Assessment of the relevancy of ENEA Water Loop facility with respect to ITER WCLL TBS Water Cooling System by considering their thermal-hydraulic performances

The Water-Cooled Lead-Lithium (WCLL) is one of the two candidate concepts for the Breeding Blanket (BB) of DEMO. A Test Blanket Module (TBM) together with its Water Cooling System (WCS) is going to be installed and tested in the ITER reactor. The WCS acts as primary cooling circuit of the TBM module, and it is designed to reproduce the water thermodynamic conditions expected at the DEMO BB inlet.During last years, ENEA and the DIAEE of Sapienza University of Rome have carried out the conceptualization of the Water Loop (WL) facility, belonging to the W-HYDRA experimental platform planned at C.R. Brasimone. The W-HYDRA platform is composed by three individual facilities called: Water Loop, Steam, and LIFUS5/Mod4. Water Loop replicates the salient thermal-hydraulic features of the ITER WCLL WCS, and it is equipped with a test section placed inside a Vacuum Vessel (VV) to investigate mock-ups of the whole TBM or its individual parts.This paper assesses the relevancy of the WL facility with respect to ITER WCLL TBM System comparing their thermal-hydraulic performances during selected operational and accidental conditions. Two RELAP5/Mod3.3 models were developed, and the outcomes of the simulations showed a good agreement, thus demonstrating the effectiveness of WL facility to support the ITER TBM program.

Impacts on Crash Cushions—Analysis of the Safety Performance of Passenger Cars with Improved Safety Equipment Compared with Test Vehicles Based on Assessment Criteria as Defined in EN 1317

To assess the safety performance of crash cushions, guidelines or standards are used. Real-life accident conditions might deviate substantially from the approval test conditions. The objective of this study is to evaluate occupant safety in passenger cars in the event of an impact against a crash cushion. Real-life accident configurations deviate significantly from the impact configurations used in the approval test EN 1317. In four different tests, two vehicles regularly used in EN 1317 and two vehicles with improved safety equipment (airbag, pretensioner, and load limiter) are used. The impact speed is 100 km/h, whereas the crash cushion is designed for an impact speed of 80 km/h. One configuration is defined as a full overlap, and one has a 50% offset. The ASI (Acceleration Severity Index), THIV/OIV (Theoretical Head Impact Velocity/Occupant Impact Velocity), and PHD/ORA (Post Head Deceleration/Occupant Ride down Acceleration) are calculated from the acceleration signals. The offset impact was more serious for both the regularly used vehicle and the vehicle with improved safety equipment. Vehicles with improved safety equipment do not have any influence on these criteria. It is apparent that new occupant safety technologies will not have any influence on occupant safety performance. The criteria currently in use are more likely to be of use for assessing vehicle performance rather than occupant safety.

Fragility and Leakage Risk Assessment of Nuclear Containment Structure under Loss-of-Coolant Accident Conditions Considering Liner Corrosion

The steel liner plays a key role in ensuring the leak-tightness of nuclear containment structures (NCSs). Once the liner is subjected to corrosion, its corrosion effects can severely impact the leak-tight function of the NCS, especially in severe accident scenarios. Therefore, evaluating the corrosion effects of the liner is essential to guaranteeing the safe operation of nuclear power plants. This paper presents a probabilistic safety analysis of the NCS under liner corrosion conditions. Firstly, the corrosion mechanism of the liner is elucidated, and a refined simulation method is developed to investigate the localized corrosion effects of the liner. Utilizing a probabilistic finite element method, the fragility of the NCS under loss-of-coolant accident (LOCA) conditions is evaluated. Finally, the leakage risk of the NCS at different degrees of liner corrosion is discussed. The results indicate that liner corrosion has a significant impact on the median pressure capacity and high-confidence–low-probability pressure capacity of the NCS. With the aggravation of liner corrosion, the safety margin of the NCS decreases, and the total probability of failure increases. Within the scope of this paper, the NCS can fulfill the probabilistic safety requirements.

Data-driven model order reduction for sensor positioning and indirect reconstruction with noisy data: Application to a Circulating Fuel Reactor

Sensor positioning and real-time estimation of non-observable fields is an open question in the nuclear sector, especially for advanced nuclear reactors. In Circulating Fuel Reactors (CFR), liquid fuel and coolant are homogeneously mixed, and thus these reactors will not have internal structures, making sensor positioning in the primary circuit, including the core, an unresolved problem, making most of the core blind to sensors. Thus, the possibility of estimating the system state in the whole domain using a few local measurements has important implications for safety, monitoring, and control both in nominal and accidental conditions. In this context, the integrated Model Order Reduction and Data Assimilation framework offers intriguing opportunities to reliably combine experimental data and background knowledge from a reduced mathematical model. This work discusses and applies innovative methods within this framework, based on the Generalized Empirical Interpolation and the Indirect Reconstruction algorithms, to a proposed concept of CFR. This work aims to identify the optimal sensor positioning within the core and assess the feasibility of reconstructing the quantities of interest starting only from transient sparse data on fuel temperature, possibly noisy, and testing the predictive capabilities of the discussed methods.

Formulation of high-temperature strength equation of 9Cr-ODS tempered martensitic steels using the Larson–Miller parameter and life-fraction rule for rupture life assessment in steady-state, transient, and accident conditions of fast reactor fuel

This paper discusses the applicability of Straalsund et al.’s technique for combining the Larson–Miller parameter (LMP) and life-fraction rule to form a single high-temperature strength equation for 9Cr-oxide-dispersion-strengthened (ODS) tempered martensitic steels (TMS). It uses the extensive dataset on creep rupture, tensile, and temperature-transient-to-burst tests of 9Cr-ODS TMS cladding tubes in the α-phase, α/γ-duplex, γ-phase matrices, which are accumulated by the Japan Atomic Energy Agency so far. The technique is adequately applicable to 9Cr-ODS TMS cladding tubes. A single high-temperature strength equation expressing the mechanical strength in different deformation and rupture modes (creep, tensile, temperature-transient-to-burst) is derived for 9Cr-ODS TMS cladding tubes. This equation can predict the rupture life of the cladding tubes under various stresses and temperatures over time. The applicable range of the high-temperature strength equation is specified in this study and the upper limit temperature for the equation is found to be 1200 °C. At temperatures higher than 1200 °C, the coarsening and aggregation of nanosized oxide particles and the γ to δ phase transformation are reported in previous studies. The high-temperature strength equation can be well applied to the creep and tensile strength in the α-phase matrix, the creep strength in the γ-phase matrix and the temperature-transient-to-burst strength in both phases except for the low equivalent stress (43 MPa) at temperatures exceeding 1050 °C. The mechanism of the notable consistency between creep and tensile strength in the α-phase matrix is discussed by analyzing the high-temperature deformation data in the light of existing deformation models. The study suggested the thermal activation process is dominant even for the high-temperature deformation modes exceeding yield stress in the 9Cr-ODS TMS with the fine-grained matrix.

ORCA: A Tool for Radiological Consequences for Accidental Releases

Evaluating atmospheric dispersion and radiological doses in the vicinity of buildings is required for small modular reactors (SMRs) because of the reduced size of their exclusion area boundary. The current Canadian nuclear industry tool for these calculations implements the methodology defined in CSA Standard N288.2-M91, which was written to support large Canada Deuterium Uranium (CANDU) nuclear reactors as opposed to SMRs. The ORCA (On/offsite Radiological Consequences of Accidents) code has been developed to address this technical concern in addition to evaluating atmospheric dispersion and doses in the far field. The code calculates worker and public doses following an airborne release of radioactive material into the atmosphere under postulated accident conditions at a nuclear facility. The current paper presents the key assumptions and methods utilized in ORCA and discusses qualification of the software to the requirements of CSA Standard N286.7-16. The new model is applicable to SMRs and existing reactor designs and reduces conservatisms in the near field (i.e., <1 km from the source) relative to the methods in CSA N288.2-M91.

Radiological safety assessment for transportation of reactor pressure vessel during decommissioning of a nuclear power plant in Korea

In Korea, decommissioning of nuclear power plants and transportation of the decommissioning waste are expected to expand in the near future. It is necessary to confirm that radiological risks to the public and workers are not significant through radiological safety assessment. The objective of this study is to assess the radiological safety for transportation of RPV waste, which is a major decommissioning waste with relatively high level of radioactivity. It was assumed that the waste would be transported to the Gyeongju disposal facility by land transportation. First, the source term and transportation method of the RPV waste were determined, and the external dose rates from the waste were calculated using MCNP. Then, transportation scenarios were assumed under both normal and accident conditions. Under the scenarios, radiation doses were calculated using the RADTRAN. Under normal operation scenarios without a transportation accident, assuming 40 shipments per year, the average individual doses for the public ranged from 6.56 × 10−6 to 2.18 × 10−2 mSv yr−1. The maximum individual doses for only a single shipment ranged from 2.43 × 10−6 to 3.14 × 10−1 mSv. For cargo handlers and vehicle crew members, the average doses were 2.26 × 101 mSv yr−1 and 2.95 mSv yr−1, respectively. Under transportation accident scenarios, average individual radiological risks which are product of the radiation doses and the annual accident rates ranged from 1.14 × 10−11 to 1.61 × 10−10 mSv yr−1 by transportation route segment when considering the transportation accident rate. Average individual doses assuming transportation accident occurrence ranged from 2.62 × 10−4 to 1.42 × 10−3 mSv. The maximum individual dose under accident conditions was 7.99 × 10−2 mSv. The calculated doses were below the regulatory limits in Korea. However, relatively high doses were observed for cargo handlers and vehicle crew members because of conservative assumptions. This study results can be used as basic data for the radiological safety assessment for the decommissioning waste transportation in the future.

Bayesian statistical model for cladding high-temperature burst under loss-of-coolant accident conditions

ABSTRACT This study developed a probabilistic determination model with respect to cladding high-temperature burst conditions based on the Bayesian statistical method to reasonably evaluate fuel behaviors under loss-of-coolant accident conditions, including fuel fragmentation, relocation, and dispersal. The candidate models (correlations) were based on the widely accepted empirical model established based on nonirradiated fuel cladding data. Explanatory variables were added to improve the applicability of these models with respect to irradiated materials and generalization performance. The posterior predictive distribution of each candidate model was evaluated using Bayesian estimation comprising 238 sets of high-temperature burst test data. The generalization performance was evaluated using information criteria. The results of model evaluation showed improved predictive performance by considering the effect of hydrogen content. A comparison with burnup as an alternative explanatory variable confirmed that hydrogen content was the better parameter and other burnup-associated effects, such as irradiation hardening of the metal matrix and oxide growth (reduction of the metal matrix), were less dominant under burst conditions.

A comprehensive examination of the linear and numerical stability aspects of the bubble collision model in the TRACE-1D two-fluid model applied to vertical disperse flow in a PWR core channel under loss of coolant accident conditions

The one-dimensional Two-Fluid concept uses an area-average approach to simplify the time and phase-averaged Two-Fluid conservation equations, making it more suitable for addressing difficulties at an industrial scale. Nevertheless, the mathematical framework has inherent weaknesses due to the loss of details throughout the averaging procedures. This limitation makes the conventional model inappropriate for some flow regimes, where short-wavelength perturbations experience uncontrolled amplification, leading to solutions that need to be physically accurate. The critical factor in resolving this problem is the integration of closure relations. These relationships play a crucial function in reintroducing essential physical characteristics, thus correcting the loss that occurs during averaging and guaranteeing the stability of the model. To improve the accuracy of predictions, it is essential to assess the stability and grid dependence of one-dimensional formulations, which are particularly affected by closure relations and numerical schemes. The current research presented in the text focuses on improving the well-posedness of the TFM, specifically within the TRACE code, which is widely utilized for nuclear reactor safety assessments. Incorporating a bubble collision model in the momentum equations is demonstrated to enhance the TFM's resilience, especially in scenarios with high void fractions where conventional TFMs may face challenges. The analysis presents a linear stability analysis performed for the transient one-dimensional Two-Fluid Model of system code TRACE within the framework of vertically dispersed flows. The main emphasis is on evaluating the stability characteristics of the model while also acknowledging its susceptibility to closure relations and numerical techniques.

Validation of transient SYSTEM analysis code GINKGO based on a SLB TEST

Transient System analysis code is an important tool for reactor thermal hydraulic analysis. GINKGO is a transient system analysis in-house code developed by China General Nuclear Power Corporation (CGN). The code is primarily used for the analysis of Non-Loss of Coolant Accident (Non-LOCA)transients. Steam Line Break (SLB) accident is one of the representative accident condition of Non-LOCA transients. To validate the capability of GINKGO to simulate SLB accident scenario, a test of small non-isolatable main steam line break (10 %) at hot standby conditions, PKL III B5.1 test is selected. The characteristic of PKL facility, the PKL III B5.1 test sequence and the validation process of PKL III B5.1 test by GINKGO code are briefly introduced in the paper. In validation process, GINKGO simulates the steady state first and then simulates the transient process by setting reasonable initial conditions and boundary conditions. Then, the calculation results of key parameters including temperature in hot leg, temperature in cold leg, Steam Generator (SG) secondary side pressure and Pressurizer (PZR) pressure are compared with the experimental results. Through the rationality of the calculation results, it shows that GINKGO has the capability to simulate the key phenomena in a typical SLB accident scenario.

The critical influencing factors responsible for the particle cracking in UMo/Zr dispersion fuel plates during post-irradiation anneal tests

The post-irradiation anneal tests for dispersion fuel plates are commonly employed to determine the blistering threshold temperature in order to evaluate their resistance to accidents. In this study, a three-dimensional simulation method for UMo/Zr dispersion fuel plates is developed to predict the normal stresses in the fuel skeleton and some related variables during in-pile irradiation and out-of-pile anneal tests. The effects of the intrusion of extra fission gas atoms into gas bubbles are investigated, and the contributions of the creep performances of fuel particles and the matrix are examined. The predictions reveal the following insights: (1) the thermo-mechanical behaviors during anneal tests are strongly dependent on the intrusion of extra fission gas atoms into gas bubbles and the additional fission-gas induced swelling during post-irradiation anneal tests; the maximum tensile stress in the fuel skeleton increases significantly, from 57.4 MPa to 346.0 MPa, as the fraction of generated fission gas atoms in gas bubbles shifts from approximately 23.4 % to 100 % in the annealing process; (2) the thermo-mechanical responses during anneal tests are greatly influenced by the creep performances of fuel particles and the matrix; the significantly weakened creep ability of fuel particles is found to be the critical factors responsible for particle cracking in the annealing process; the fuel particle fracture will hardly occur when the irradiation creep rate model is maintained during the considered annealing scheme, because the skeleton tensile stress does not appear. It is important to note that the particle cracking and the subsequent blister behaviors under in-pile neutron irradiation conditions are possible to occur if a rapid temperature increase takes place under accident conditions. Consequently, the out-of-pile annealing tests are of significance.

Oxide Layers in Ni-doped FeCrAl Alloy in 320°C Radioactive Hydrogenated Water

Iron-chromium-aluminum (FeCrAl) alloys have been the interest of the international nuclear energy industry as an accident tolerant fuel cladding replacement to the Zircaloy cladding alloys due to their enhanced corrosion resistance in both normal operation and accident conditions in light water reactors. It is important to understand the corrosion behaviors of FeCrAl before deploying it to nuclear power plants. This study aims to investigate if the addition of a small amount of nickel (Ni) could stabilize the oxide layer by forming the Ni-bearing spinel oxide, such as NiFe2O4, and increase the stability of the oxide layer under irradiation conditions. Current results show that the outer oxide crystals formed on the irradiated FeCrAl-Ni sample surface became larger than that on the FeCrAl sample, which indicates that the Ni addition facilitates the growth of oxide layers and that the FeCrAl-Ni sample exhibited a protective inner oxide layer in the harshest irradiation environment, indicating that Ni improves the corrosion resistance of FeCrAl in radioactive hydrogenated water.

Degradation of the Cr2O3 layer protectiveness caused by grain refinement during high-temperature oxidation of Cr coatings

To enhance the understanding of the failure mechanism of Cr coatings under simulated loss of coolant accident (LOCA) conditions in pressurized water reactors, an investigation into the isothermal oxidation behavior of Cr coatings in 1200 °C steam was conducted. The structure evolution of Cr coatings was characterized by scanning electron microscope (SEM) and electron backscatter diffraction (EBSD) methods. The obtained results reveal the formation of a three-layer structure during the 1200 °C steam oxidation process. This structure comprises a Cr2O3 layer, a residual Cr layer, and a ZrCr2 layer. Over time, the grains in the Cr2O3 layer undergo refinement, leading to an increase in grain boundaries. This phenomenon promotes the inward diffusion of oxidizing medium, contributing to understanding of the degradation of Cr2O3 layer protectiveness after prolonged oxidation. These findings offer valuable insights into the failure mechanism of Cr coatings in high-temperature steam environments.

A study on the fracture pattern change of high-burnup fuel cladding failed by pellet-cladding mechanical interaction failure under reactivity-initiated accident conditions

ABSTRACT In a part of the historical reactivity-initiated accident (RIA) simulated tests on high-burnup fuels performed at the Nuclear Safety Research Reactor, the fuel failure caused by pellet-cladding mechanical interaction (PCMI) led to splitting into upper- and lower-part pieces or even fragmentation of the cladding tube. A massive release of fuel fragments accompanied this fracture pattern change from previously known axial cracks and thus identified as a potential concern in safety evaluation regarding core coolability. Dedicated out-of-pile mechanical tests were performed with unirradiated Zircaloy-4 cladding specimens to clarify the condition of such fracture pattern change. The specimens were pre-hydrided and subjected to loading with axial-to-hoop strain ratios of ℇz/ℇθ = 0.5–1.25, simulating the effects of hydrogen embrittlement and pellet-cladding mechanical bonding of high-burnup fuels, respectively. The results indicate that higher biaxiality of the loading and lower ductility (failure strain level) assist the fracture pattern change. This study proposes a conservative criterion that a PCMI failure splits the cladding tube into more than two pieces when strain ratio ℇz/ℇθ >0.75 and a concurrent hoop strain < 1.7% at the failure instant.

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.