Loss Of Coolant Accident Research Articles

Corium interface flow dynamics investigation during severe accident in pressurized water reactors using compressive advection interface capturing method

Past nuclear accident occurrences raised strong concerns which led to research on nuclear safety. One of the major causes of nuclear accidents is the impeded circulation of core coolant, leading to decay heat removal cessation and rapid temperature rise. If uncontrolled, this results in critical heat flux, loss of coolant accidents, and core dryout. Detailed melted core relocation (i.e., nuclear fuel, graphite, and zircaloy) needs to be investigated through interface capture and multimaterial flow model coupling, which have not been done in previous studies. This work aims to investigate the impacts of temperature and core material composition on the flow dynamics during core relocation. In this study, mass fraction is discretized using a streamlined upwind Petrov–Galerkin method spatially and a modified Crank–Nicolson method temporally to accurately capture fluid interfaces using a high-order accurate flux-limiter. Two core material composition cases (individual material properties case and bulk material properties case) were considered to assess the impact of temperature and core materials composition on both flow dynamics and computational time. Temperature has a significant impact on core material transport and corium flow dynamics during core relocation. Bulk materials properties case has greater impact of temperature on its corium resulting in faster materials transport, but with higher computation time.

Tailored high-temperature corrosion behavior of Cr coatings using high power impulse magnetron sputtering on ZIRLO alloys for accident-tolerant fuel application

To further enhance the oxidation resistance of Zirconium alloys during Loss of Coolant Accident (LOCA), protective Cr coatings were deposited in synchronized pulse bias mode through high power impulse magnetron sputtering (HiPIMS). The results showed the weight gain of the Cr coatings with synchronized pulse bias was reduced by approximately 61.7 % compared to that of the Cr coatings without synchronized pulse bias after 90 min of steam oxidation at 1200 °C. This was mainly attributed to the formed intense (200) preferred orientation of the Cr coatings with synchronized pulse bias during deposition, accelerating the recrystallization growth of the residual Cr layer in a high-temperature steam environment, thereby reducing the rapid diffusion pathways for oxygen. Moreover, the synchronized pulse bias induced Cr grains refinement, favoring the formation of a more compact Cr2O3 scale, and therefore efficiently inhibiting the diffusion of external oxygen to the underlying alloy.

Large pipe break opening time in pressurized heavy water reactors

In the frame of the licensing process of the AtuchaII PHWR (pressurized heavy water reactor) the BEPU (Best-Estimate Plus Uncertainty) approach has been selected for issuing Chapter 15 of FSAR (Final Safety Analysis Report) dealing with accident analysis. In case of a LOCA (Loss of Coolant Accident) the assumption of instantaneous opening time is too conservative, especially for this kind of reactor with a positive coolant void reactivity coefficient.Analysis was performed to assess the ability of Atucha II nuclear power plant’s fast boron-injection system to prevent overheating of the fuel elements in the event of a Beyond Design Basis Accident (BDBA). The goal of the Analysis was to develop a Dynamic Break Opening which was used as the input for the LOCA analyses. The results show that the fast boron injection can effectively shut down the reactor, thereby preventing the overheating of the fuel elements.

A thermogravimetric comparative study of the high temperature steam oxidation of bare and pre-oxidized Zircaloy-4, M5Framatome and Optimized ZIRLO™

For Loss of Coolant Accident (LOCA) analysis, high-temperature oxidation kinetics in steam of Zircaloy-4 and the two Nb-bearing alloys used in French Pressurized Water Reactors (PWRs), M5Framatome11M5 and M5Framatome are trademarks or registered trademarks of Framatome or its affiliates, in the USA or other countries. and Optimized ZIRLO™22Optimized ZIRLO is a trademark or registered trademark of Westinghouse Electric Company LLC, its affiliates and/or its subsidiaries in the United States and may be registered in other countries throughout the world. All rights reserved. Unauthorized use is strictly prohibited. Other names may be trademarks of their respective owners., have been studied using Thermo-Gravimetric Analysis (TGA) technique in the temperature range of 900–1200 °C. A specific test procedure allowing the use of a high steam flow rate and good steam supply to the whole specimen surface, as well as fast heating of the specimen, has been developed. Clean kinetic data, not affected by steam starvation, were obtained on the bare, as-received alloys. For as-received claddings, oxidation kinetics are comparable and close to parabolic for the three alloys at 1100 °C and above. Below 1100 °C, a significant deviation from the parabolic kinetic regime was observed, and M5 Framatome oxidizes slower than Zircaloy-4 and Optimized ZIRLO. Thanks to the fast-heating protocol, the influence of pre-oxidation, simulating corrosion in normal reactor operation, could be investigated, avoiding evolution of the pre-oxidation scale during a slow, long heating phase. Pre-oxidized Zircaloy-4 and M5 Framatome were investigated. A strong protective effect is observed below 1200 °C, while at 1200 °C, the pre-oxidation layer appears to be much less protective, possibly due to the monoclinic to tetragonal zirconia phase transformation.

A novel plasma-sprayed Cr/FeCrAl dual-layer coating on Zr alloy for potential high‐temperature applications

To better meet the requirements in loss-of-coolant accidents (LOCA) and conventional operating conditions in nuclear applications, this work prepared Cr, FeCrAl, and Cr/FeCrAl composite coatings on Zr alloys using plasma spraying technology. The microstructures, hardness, and high-temperature oxidation resistance of coatings are examined. The high-temperature oxidation of coatings was conducted at 1200 °C in air for 20, 40, and 60 min. All coatings exhibited characteristic lamellar microstructures accompanied by small pores. The inner layers of the coatings, benefiting from the “compacting effect” of subsequently deposited particles, displayed lower porosity and higher hardness. The order of high-temperature oxidation resistance among the coatings was determined as follows: Cr/FeCrAl coating > Cr coating > FeCrAl coating. Notably, the Cr/FeCrAl composite coating demonstrated exceptional durability even after a 60-min exposure, with no ZrO2 formation and negligible Fe and Cr diffusion. However, oxygen could still reach the Zr substrate due to inherent pores and the lamellar structure in the as-sprayed coatings. Therefore, post-treatments are suggested to mitigate or eliminate these pores, thereby improving high-temperature oxidation resistance.

Effect of Si on the structure and oxidation behavior of Cr-Si coatings on Zry-4 under high temperature steam environment

Under loss of coolant accident (LOCA), zirconium alloy (Zry-4) is easily oxidized by high temperature steam, which makes its performance deteriorate. To improve the oxidation resistance of Zry-4 under LOCA, Cr-Si coatings with different Si content were deposited on Zry-4 by magnetron sputtering, and the effects of Si content on the steam oxidation resistance of the Cr-Si coatings were investigated. With the fluctuation of Si content inside the coating, there appears different lattice structure. Some amorphous phases and lattice distortions in the coatings with moderate Si content were discovered by TEM, which were beneficial to limit O diffusion in the lattice. However, in the coatings with the higher Si content, the lattice distortions transform into dislocations, which contributes to the O diffusion. The oxidation kinetics of the coated samples proved that the Cr-Si coating with lattice distortion was more consistent with the parabolic law and showed excellent oxidation resistance. In addition, the results of first-principles calculations supported the above view that the O diffusion energy in the lattice increases after the introduction of Si.

Time in DNB experimental study on Cr coated zircaloy cladding

Since the Fukushima-Daiichi accident effort has been put in by the nuclear community to develop Accident Tolerant Fuel (ATF) cladding materials to reduce the risk of clad oxidation and, consequently, hydrogen generation and potential major accidents. Among designed materials, the use of a chromium coating on top of zirconium-alloy cladding showed reasonable results in preliminary tests, being a favorable cladding solution for Pressurized Water Reactors. This paper provides highly detailed temperature data obtained during time in departure from nucleate boiling (DNB) experiments under prototypical reactor conditions of temperature, pressure, mass flux, and heat flux profile on a Cr-coated zirconium-alloy cladded heater rod simulating the fuel pin. Data were obtained at mass fluxes from 1300 to 1700 kg/m2s, pressure of 16 MPa, and inlet temperature of 333 °C (subcooling degree of 14.4 °C). The change in dry time was obtained by increasing the spike in wall temperature necessary to shut down the power. These tests showed that increasing the spike in temperature to characterize the critical heat flux from 50 °C to 400 °C resulted in variations of the departure from nucleate boiling measured value within the uncertainty range of their measurements, hence, validating the use of a spike of 50 °C, a commonly used value in this kind of study, as an indicator of the critical heat flux. Within the dry region, the temperature was found to increase along the direction of the flow for a very short length, reaching a peak, from which it reduces linearly until the surface reaches a temperature that is no longer able to sustain a dry patch. It was noticed that the dry region can grow against the flow direction due to the cosine profile of the heat flux, and that small dryout events can occur without leading to departure from nucleate boiling, i.e., being fully suppressed. However, two consecutive events may sum up and cause the critical heat flux. Rewet time was found to increase with peak cladding temperature and decrease with mass flux. A maximum peak cladding temperature of 844 °C was found, with the surface remaining dry for approximately 14.1 s. The minimum film boiling temperature was measured during the rewet process, showing absolute values of around 580 °C, and difference between wall and inlet temperatures of about 240 °C. A post-inspection using microscope and X-ray images showed no damage to the cladding and heater structure, differently than what was noticed for the bare zircaloy in a previous study. This indicated that the presence of chromium coated cladding can improve the oxidation resistance of the cladding not only in short critical heat flux transients but also in longer events associated with major accidents, like a Loss of Coolant Accident.

Preliminary accident analyses of in-vessel LOCA for the chinese fusion test reactor HCCB blanket concept

For the helium-cooled ceramic breeder (HCCB) blanket concept improved in 2018 and the primary heat transfer system following China Fusion Engineering Test Reactor (CFETR) in 2019, an in-vessel loss of coolant accident (LOCA) has been investigated with the assumption of a break of the first wall (FW) coolant channels and a double-ended guillotine break of a helium cooling system (HCS) pipe in vacuum vessel (VV). The break leads to helium inject into the VV. After the accident occurs, the helium release leads a plasma disruption and pressure increase in the VV. This paper presents the analysis results of the in-vessel LOCA accidents in CFETR and gives the accident sequence compare with the break in the FW and header pipe. The accident analysis and influence compares can provide a reference for the iterative design.

Preliminary accident analysis of the loss of vacuum in vacuum vessel for the European DEMO HCPB blanket concept

Design basis accidents are investigated continuously for the European DEMO reactor accompanying its development. One selected postulated initial event (PIE) is a loss of vacuum (LOVA) in vacuum vessel (VV) with large ingress of air induced by rupture in a VV penetration. It has been investigated for the helium cooled pebble bed (HCPB) blanket concept according to the DEMO baseline 2017. The associated primary heat transfer system (PHTS) and the related systems in the tokamak building, from the VV to the PHTS vault and galleries, are considered for the investigation. The LOVA is postulated to occur at a port seal of the electron cyclotron equatorial port plug on the side of the closure plate with (i) a small leak of 1.0 × 10−3 m2, or (ii) a large break size of 1.0 × 10−2 m2. Air ingress from one port cell into the VV leads to the VV pressurization and the fusion power termination followed by an unmitigated plasma disruption. A loss of off-site power for 32 h is assumed to coincide with the disruption. An in-vessel loss of coolant accident (LOCA) is considered as a consequence if the affected first wall (FW) reaches the defined temperature of 1000 °C. The radioactive inventories in the VV (tritium, W-dust) can mobilize towards the VVPSS, the affected systems in the building and the environment due to pressurization, venting and leak conditions. MELCOR 1.8.6 for fusion is applied for this deterministic safety analysis. The resulting releases of radioactivity to the environment are then provided for dose calculation using the computer systems UFOTRI and COSYMA. Outcomes of this LOVA analysis are critically discussed: the transient evolutions of different cases are compared; hydrogen production is detected in case of aggravating FW failure; the source terms (tritium, W-dust) are transported to the connected systems; and the dose results from the environmental releases are provided.

Prediction of small-scale leak flow rate in LOCA situations using bidirectional GRU

It is difficult to detect a small-scale leakage in a nuclear power plant (NPP) quickly and take appropriate action. Delaying these procedures can have adverse effects on NPPs. In this paper, we propose leak flow rate prediction using the bidirectional gated recurrent unit (Bi-GRU) method to detect leakage quickly and accurately in small-scale leakage situations because large-scale leak rates are known to be predicted accurately. The data were acquired by simulating small loss-of-coolant accidents (LOCA) or small-scale leakage situations using the modular accident analysis program (MAAP) code. In addition, to improve prediction performance, data were collected by distinguishing the break sizes in more detail. In addition, the prediction accuracy was improved by performing both LOCA diagnosis and leak flow rate prediction in small LOCA situations. The prediction model developed using the Bi-GRU showed a superior prediction performance compared with other artificial intelligence methods. Accordingly, the accurate and effective prediction model for small-scale leakage situations proposed herein is expected to support operators in decision-making and taking actions.

Evaluations of radionuclide activity releases into environment during loss of coolant accidents using the ASTEC code in pressurized water reactors within design basis and design extension conditions

The work described in this paper was carried out within the R2CA (Reduction of Radiological Consequences of design basis and extension Accidents) project, funded in HORIZON 2020 and coordinated by IRSN (France). An increase of the level of Nuclear Power Plant (NPP) safety by consolidated and more realistic evaluations of the Radiological Consequences (RC) of Design Basis Accidents (DBA) and a strengthening of the assessments of the NPP safety levels by considering accidental situations more severe than those integrated in plant designs (i.e belonging to Design Extension Conditions domain) were the two main motivations behind this project.More specifically, the project aims at consolidating and/or refining the assessments of the radiological consequences of explicit accidental scenarios within Design Basis Accidents (DBA) and Design Extension Conditions (DEC-A conditions without significant fuel degradation) in Light Water Reactors (LWR) through the improvements of existing code predictability; the upgrading of calculation chains and methodologies; the development/refinements of models. Within the Work Package 2 of the project, coordinated by TRACTEBEL and dedicated to calculation methodologies, existing methodologies or calculation chains and simulation tools have been applied to run a first batch of calculations dealing with different reactor types: PWR (Pressurized Water Reactor), BWR (Boiling Water Reactor), VVER (Water-Water Power Reactor) and EPR (European Pressurized Reactor). Loss Of Coolant (LOCA) and Steam Generator Tube Rupture (SGTR) accidents have been selected for the exercise and bounding scenarios of the DBA and DEC-A domains have been analysed. The results of this first set of calculations will be used as a reference to quantify the gains obtained by the updated methodologies/simulation tools developed within the project.This paper describes the results of the first batch of calculations, performed with the ASTEC integral code, simulating LOCA scenarios (DBA and DEC-A categories) in a PWR 900 MWe with a focus on the predicted number of failed fuel rods and Source Term (ST) in the environment governing the RC. Limitations of the used approach are outlined, as well as the needs for further upgrading the calculation chains are proposed in the light of the improvement that was planned within the project in Work Package 3 (WP3: LOCA – Loss of Cooling Accidents), coordinated by IRSN and dedicated to the improvement of code models dealing with LOCA scenarios.

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.

Enhancing cladding mechanical modelling during DBA/LOCA accidents with FRAPTRAN: The TUmech one-dimensional model

The analysis of the cladding mechanical performance is essential for predicting rod failure during the heat-up phase of a loss-of-coolant accident (LOCA). Previous modelling and simulation exercises have shown that there is room for improvement in the accuracy of fuel performance simulations under such conditions. In this work, TUmech, a simplified standalone version of the mechanical model in TRANSURANUS, has been developed and coupled with FRAPTRAN-2.0. After highlighting the most significant differences between both approaches, particularly in their mechanical treatment and material property correlations, simulations have been performed for IFA-650.2 (fresh fuel) and IFA-650.10 (high burnup fuel) of the Halden LOCA test series. The results obtained from both FRAPTRAN versions have been compared with experimental data and benchmarked against TRANSURANUS estimates, focusing on rod inner gas pressure, post-test cladding outer diameter, and burst time. In both scenarios, FRAPTRAN-TUmech has proven to enhance the accuracy in predicting these three variables compared to the FRAPTRAN default version. However, further validation is required before drawing definitive conclusions based on the encouraging results.

Research on thermal hydraulic characteristics of reflooding process based on core multi-channel model in system analysis code

There are four main stages in PWR Loss of Coolant Accident (LOCA): blowdown, re-irrigation, re-flooding and long-term cooling. The re-flooding stage is a critical phase to avoid core meltdown and achieve long-term cooling. RBHT reflooding experiment was simulated with the RELAP5 and COSINE program respectively.Thermal-hydraulic characteristics in re-flooding phase were analyzed using multi-channel model. Quench time is mainly influenced by system pressure and coolant subcooling. Later quench will result in a higher cladding temperature. Comparatively, the peak cladding temperature is less affected by the power of the heating rod and the re-flooding rate, which fluctuates within 7 %. The higher re-flooding rate allows quenching to occur as fast as possible and improves cooling efficiency. However, higher coolant sub-cooling and re-flooding rates may cause greater thermal stresses on the core and endanger core integrity. Based on the analysis of the re-flooding rate on the peak cladding temperature and quench time, the optimal re-flooding rate interval of 0.099–0.105 m/s was obtained.

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.

Time-Series Forecasting of a Typical PWR Undergoing Large Break LOCA

In this work, a machine learning (ML) metamodel is developed for the time-series forecasting of a typical nuclear power plant response undergoing a loss of coolant accident (LOCA). The plant model of choice is based on the APR1400 nuclear reactor. The key systems and components of APR1400 relevant to the investigated scenario are modelled using the thermal-hydraulic code, RELAP5/MOD3.4, following the description published in the design control document. The model is tested under a spectrum of initial and boundary conditions via propagation of key uncertain parameters (UPs) which are derived from the phenomena identification and ranking table (PIRT). This is achieved by loosely coupling RELAP5/MOD3.4 with the statistical tool, Dakota. The most probable nuclear power plant (NPP) response was calculated using the best estimate plus uncertainty (BEPU) approach. Next, the database generated from the NPP system response was used as an input for the ML model. The NPP system response was represented by peak cladding temperature (PCT), safety injection system (SIT), mass flow rate, reactor power, and primary system pressure. In this research, two regression models were tested with reasonably good performance, namely, the gated recurrent unit (GRU) and the long short-term memory (LSTM).

An FE-MCS based modeling method for event sequence development for the multi-module probabilistic risk assessment

The Multi-Unit Probabilistic Risk Assessment (MUPRA) has been a research hotspot since the Fukushima Daichi accident in 2011. Among the technical elements, event sequence modeling is the most challenging one in MUPRA implementation. During the PRA development for a 600 MW HTR-PM nuclear power plant (HTR-PM600) which consist of 6 reactor modules, we find that the complexity of the obtained event sequence model increases rapidly and massively with the number of reactors, therefore it is quite difficult to guarantee the modeling correctness. To address this issue, in this paper, we propose a multi-module event sequence modeling method based on Minimal Cut Sets at the level of Functional Event (FE-MCS). The proposed method is illustrated with two cases in this paper, namely, Loss of Coolant Accident caused by the Isolatable Large breach in primary pressure boundary (LLOCA-I) and Loss of vital Direct Current bus (LSDC).

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.

Evaluating direct vessel injection accident-event progression of AP1000 and key figures of merit to support the design and development of water-cooled small modular reactors

The passive safety systems (PSSs) within water-cooled reactors are meticulously engineered to function autonomously, requiring no external power source or manual intervention. They depend exclusively on inherent natural forces and the fundamental principles of reactor physics, such as gravity, natural convection, and phase changes, to manage, alleviate, and avert the release of radioactive materials into the environment during accident scenarios like a loss-of-coolant accident (LOCA). PSSs are already integrated into such operating commercial reactors as the Advanced Pressurized Reactor-1000 MWe (AP1000) and the Water-Water Energetic Reactor-1200 MWe (WWER-1200) are adopted in most of the upcoming small modular reactor (SMR) designs. Examples of water-cooled SMR PSSs are the passive emergency core-cooling system (ECCS), passive containment cooling system (PCCS), and passive decay-heat removal system, the designs of which vary based on reactor system-design requirements. However, understanding the accident-event progression and phases of a LOCA is pivotal for adopting a specific PSS for a new SMR design. This study covers the accident-event progression for direct vessel injection (DVI) small-break loss-of-coolant accident (SB-LOCA), associated physics phenomena, knowledge gaps, and important figures of merit (FOMs) that may need to be evaluated and assessed to validate thermal-hydraulics models with an available experimental dataset to support new SMR design and development.

Enhancing Resilience through Nuclear Emergency Preparedness at El Dabaa Site

The research utilized advanced PCTRAN and RASCAL software to evaluate the potential radiological impacts of hypothetical accidents, specifically loss-of-coolant accident (LOCA) and long-term station blackout (LTSBO), at the El Dabaa Nuclear Power Plant. Over a span of ten years, comprehensive meteorological data were meticulously analyzed to assess the dispersion of radioactive substances within a 40-kilometer radius across all four seasons. The outcomes revealed that only in the case of LTSBO did the radiological levels surpass the limits set by the Environmental Protection Agency (EPA). Notably, during spring, LTSBO exhibited a maximum total effective dose equivalent (TEDE) value of 13 millisieverts (mSv) at a distance of 3.2 kilometers, and the highest thyroid dose (TD) recorded was 63 mSv at 8 kilometers. These significant findings play a crucial role in shaping strategies related to the distribution of potassium iodide (KI) and further enhance the overall preparedness and evacuation planning protocols.