Loss Of Coolant Accident Research Articles

Fracture studies on cruciform bend specimens of pressure vessel steel subjected to thermo-mechanical loading

The safety of a nuclear power plant during incidents, such as a Loss of Coolant Accident (LOCA), heavily relies on conducting a comprehensive structural integrity assessment of both the Reactor Pressure Vessel (RPV) and its components, specifically to withstand Pressurized Thermal Shock (PTS). PTS is characterized by a combination of steep temperature gradient, resulting from the injected emergency coolant during a LOCA, and internal pressure within the RPV. Majority of the reported fracture assessment studies on RPV steel, whether experimental or numerical investigations, have predominantly focused on standard uniaxial specimens at iso-thermal loading conditions. To better simulate the thermal shock scenario in RPVs, the present work aims to assess the impact of a biaxial stress field on both fracture parameters (crack mouth opening displacement and J-integral) and plastic collapse load. This assessment is conducted through experimental and numerical investigations, both with and without prior transient thermal load. Fracture experiments are performed on cruciform bend specimens, featuring two different biaxiality ratios (1:1 and 2:1). Moreover, numerical studies on cruciform specimens are conducted using finite element analysis to validate and corroborate the observations derived from the fracture experiments.

Simulation Analysis of Loss of Coolant Accidents Based on CNIFA Program

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

Investigation of quenching phenomena during the relooding phase against the FLECHT-SEASET experiment by using RELAP5/MOD3.3

The reflood model of RELAP5/MOD3.3 (patch 4) was assessed by using the FLECHT-SEASET tests. The tests were conducted to have a better understanding of the postulated loss of coolant accident in a light water reactor (LWR). The best-estimate system analysis code was used to simulate these accident scenarios, especially in Design Basic Accident such as Loss-Of-Coolant-Accident (LOCA). The primary purpose of this report was to assess the accuracy of a computer system analysis code by using RELAP5/MOD3.3 in comparison to actual test results taken from the FLECHT-SEASET tests in which the reflood model was built in. In RELAP5’s simulation cases, the various boundary and initial conditions, such as power supplied and reflooding rate were selected. As a result, the RELAP5 looked to be accurate in predicting the quenching time and rod surface temperature for this particular case. However, the RELAP5 code under-estimated the rod surface temperature in comparing with the experimental data of the FLECHT-SEASET tests. Accordingly, for this high flooding rate and particular reactor power level that the reflooding model in RELAP5 could be possible used for predicting the reflooding phenomena during the LOCA accident.

Combination of two-fluid model and delayed equilibrium model for the critical flow in a slit

Accurately predicting the mass flux, pressure profile, and velocity profile of the critical flow in a slit is essential for analyzing the breaking process of the liquid phase and calculating the aerosol source term for leak-before-break (LBB) monitoring and Loss of Coolant Accident (LOCA) risk analysis. A new critical flow model combining Two-fluid Model (TFM) and Delayed Equilibrium Model (DEM) is built to get accurate profiles while avoiding the same phase velocity in DEM and the arbitrary critical flow criterion in TFM. The new model is verified using past experiments of the critical flow in a slit. It proves to be accurate in mass flux but not in critical pressure, with maximum relative errors of around 25% in mass flux and around 80% in critical pressure. The new model is optimized for higher accuracy in critical pressure. The empirical equation of saturated phase mass flow rate fraction gradient is optimized by conducting approximate pressure profile calculation and regression analysis. The maximum relative error decreases little while the ratio of critical pressure relative errors lying in the range of ±40% increases after optimization. In contrast, the difference in the average abstract relative error of pressure between original TFM-DEM and DEM is much larger, for the maximum relative error of mass flux and critical pressure are around 25% and 110%. The comparison between the original and optimized TFM-DEM proves that the new critical flow model is accurate in mass flux and can be optimized to raise pressure calculation accuracy. The comparison between the original TFM-DEM and DEM proves that the phase velocity difference is the major source of accuracy improvement in the pressure profile.

Performances of the ITER Pressure Suppression System during unstable steam condensation regimes

Abstract The paper deals with the qualification of the main safety system of the nuclear fusion reactor ITER: the Vacuum Vessel Pressure Suppression System (VVPSS). This safety system manages the Loss of Coolant Accident (LOCA), that could occur in the Vacuum Vessel (VV), avoiding a significant increase of the internal pressure that could breach the primary confinement barrier. Steam condensation occurs at sub-atmospheric pressure. This condition is original respect the usual applications of steam direct condensation in nuclear fission reactors. An extensive experimental research program, funded by ITER Organization, on reduced scale experimental rigs (scale 1/22 and 1/10) have been carried out at the University of Pisa. Unstable condensation regimes occurred during some tests at about 500 g/s of steam mass flow rate (10 % of the maximum steam mass flow rate foreseen for the relevant accidental scenarios). In these conditions, high intensity vibrations of the whole experimental rig occurred. These events can be classified as ‘Chugging’ or Condensation Induced Water Hammer (CIWH). 
The performed tests have been analysed elaborating the images of the video-cameras located in the pool and comparing the steam jet shape with the acceleration values recorded by an accelerometer mounted on the sparger structure.
These elaborations permitted to evaluate the dynamic of bubble collapse, the volume of the developed external bubbles, the collapse frequency as function of thermal-hydraulic parameters. 
Moreover, the paper emphasizes the beneficial effects of non-condensable gases on the unstable condensation regimes. In fact, the non-condensable gas inhibits the occurrences of Water Hammer inside the sparger and eliminates the thermal stratification inside the water pool increasing the turbulence.

Study on the effects of accident-tolerant fuels on the safety of CPR1000 nuclear power plants

Abstract In this paper, CPR1000 Nuclear power plant was taken as the reference unit. According to CPR1000 Probabilistic Safety Analysis (PSA), the Large Break loss of coolant accident (LOCA), Medium Break LOCA, Small Break LOCA, Station Blackout (SBO), Total Loss of Feed water (TLOFW) and Anticipated transient without trip (ATWT) with Loss of Main Feed water (LOMF) were selected as the representative accident scenarios. The house holding deterministic programs LOCUST were used to analyze each of the above selected scenarios for five different accident tolerant fuels (ATF). The effects on accident process, reactor core damage time, system success criteria and personnel response time were analyzed under the above scenarios. The above results for the five ATF materials were compared to those for standard UO2-Zr in the same scenarios, then some useful conclusions were obtained which can be used to support ATF material selection. In the end, based on the deterministic calculations, the effects of different ATF materials on CPR1000 PSA results such as core damage frequency (CDF) were evaluated in detail with traditional PSA software Risk Spectrum. According to the results, ATFs have some advantages on core damage time, success criteria or operators’ action time window. But the CDF was changed little compared to standard UO2-Zr system. Through the studies, it is found the most important thing is that due to the much longer time windows some new plant modifications can be provided to mitigate the accidents which can contribute to the plant safety and decrease risk more.

VIKTORIA experiments in the frame of R&D project on sump filtration during a loss of coolant accident

During a Loss of Coolant Accident (LOCA), in PWR’s, water is injected by the Emergency Core Cooling System (ECCS) to ensure the long-term core coolability and by the Containment Spray System (CSS) to remove residual heat and to maintain containment integrity. After the drainage of the Refueling Water Storage Tank (RWST), water is taken from sumps in the lower part of the reactor building. A filtering system is implemented to collect debris produced by the pipe break as well as other latent materials, such as fiberglass, paint and concrete particles, and to minimize the amount of debris entering in the ECCS and CSS systems. IRSN has launched an experimental R&D project investigating the clogging of sump filters by integral tests performed in the VIKTORIA loop. The main objectives are to investigate the head loss of the filter (physical and chemical clogging) for prototypic upstream debris source term in relevant thermal hydraulic conditions (water temperature and flow velocity on the filter surface) in compliance with the temperature profile and the chemistry of the water in sumps during a LOCA transient. For that, the VIKTORIA loop was equipped successively with two types of 2 m2 filters used in 900MWe NPP’s. The tests highlight the settling of the largest particles (concrete, painted chips) and part of the fibers; the transport of debris (roughly 55 to 65 % of the injected debris source term) leads to the physical clogging of the filter. The debris carried to the filter generate at 80 °C (with chemistry) a very quick increase of the pressure drop across the filter (≈7 kPa) that could be due to rapid chemical effects further to fibers corrosion. At the end of the 80 °C plateau, a pressure drop increase was observed due to the temperature decrease in agreement with the water viscosity evolution. The two types of filters (rectangular pockets or planar types) behave very differently with rather low head losses for the second type; ≈1.5 kPa. Nevertheless, downstream debris source term appears to be more significant with the planar filter (compared to the filter with pockets) during the first hour of injection before the fibrous debris bed formation. We have also observed a more stabilized “cake”, during the (7 days) medium term evolution compared to that for rectangular pockets filter due to motion of debris inside pockets. The recent experiments performed with less amount of fibers (by replacement of part of fibrous materials by RMI metallic insulation) led to significantly reduce the head loss without any consequences on the downstream behavior.

Development of a transient analysis code for trans-critical simulation of SCWR

Supercritical water-cooled reactors (SCWR) experience pressure reductions below the critical point during a loss of coolant accident (LOCA). During the reactor depressurization, boiling crises may occur, leading to pressure fluctuations and significant increases in wall temperatures, posing a severe threat to the cladding. In this study, a one-dimensional transient analysis code has been developed to incorporate one-dimensional steady-state flow equations in the fluid domain and transient thermal conductivity equations in the solid domain. The code also includes a wall heat transfer model and a model for the velocity of the moving quench front to simulate transcritical depressurization transient processes. The simulation successfully predicts the typical experimental phenomena. It is reasonable to use the critical temperature as the demarcation point between the dry and wet conditions of the axial wall of the rod bundle under transcritical conditions. By combining the experimental data with the program calculations, the critical interface for the occurrence of boiling crisis under transcritical conditions is determined using mass flow rate, heat flow density and fluid temperature as inputs. The criterion for the occurrence of CHF under transcritical conditions is obtained, which provides a reference to the safety analysis of SCWRs.

A condition-informed dynamic Bayesian network framework to support severe accident management in nuclear power plants

In the nuclear industry, Severe Accident Management Guidelines (SAMGs) are developed with the objective of providing prescriptive actions for mitigating the consequences of Nuclear Power Plant (NPP) accidents. These guidelines are defined with respect to a set of expert-based prototypical severe accident scenarios and are symptom-based, i.e., they relate to the values of specific monitored plant parameters. Expert-based SAMGs might fail to deal with non-prototypical scenarios, i.e., scenarios not considered a priori by the experts. To address this issue, in this paper we propose a methodology based on Dynamic Bayesian Networks (DBNs) to support operators decision-making for avoiding the escalation of severe accidents to non-prototypical scenarios. The methodology is based on the integration of condition monitoring data into a DBN to allow it to i) infer the current Plant Operating Condition (POC), Initiating Event (IE), and NPP safety barriers Damage State (DS), ii) predict the grace time (GT) available to carry out mitigative actions, and iii) evaluate the effectiveness of the alternative candidate mitigative actions taking into account the safety barriers estimated DS and the predicted available GT. The proposed methodology is exemplified on a Loss of Coolant Accident (LOCA) in a WWER-1000.

Validation and parametric study of FFRD model in DRACCAR code

The recent revision of Emergency Core Cooling System (ECCS) regulations in Korea has necessitated the consideration of Fuel Fragmentation, Relocation, and Dispersal (FFRD) phenomena in nuclear reactor safety analyses. Consequently, the Korea Atomic Energy Research Institute (KAERI) has developed and integrated the FFRD model into the domestically licensed safety analysis code, SPACE. Globally, US NRC’s FRAPTRAN and IRSN’s DRACCAR are available for evaluating FFRD phenomena. The FRAPTRAN code incorporates the FFR model, developed by Quantum Technology, while the fuel dispersal model is currently not included. In contrast, the DRACCAR code functions as an integrated analysis platform capable of modeling multi-dimensional thermal, hydraulic, mechanical, and chemical phenomena during a Loss of Coolant Accident (LOCA). This study conducts a thorough examination of the FFRD model in the DRACCAR code and validates its applicability through analysis using the Halden IFA-650 tests. The results demonstrate satisfactory predictive capabilities. Furthermore, a parametric study of key FFRD model parameters enhances the understanding of the FFRD model in the DRACCAR code. The development of detailed physical models in the FFRD model could significantly enhance the performance of the DRACCAR code, warranting the establishment of a comprehensive framework for these advancements. In the future, code-to-code comparisons between the DRACCAR code and other domestically developed integrated analysis platforms will be conducted to investigate various phenomena in depth.

Development of a phenomena identification and ranking table (PIRT) for intermediate break loss-of-coolant accident in PWRs

Recent safety issues such as cladding oxidation and fuel fragmentation, relocation and dispersal (FFRD) make the loss-of-coolant accident (LOCA) acceptance criteria more difficult to be satisfied. To obtain the adequate safety margin and more economical use of the nuclear power plant, re-classification of LOCAs for Korean operating PWRs is now under consideration to exclude the large break LOCA (LBLOCA) from the design basis accidents (DBAs). Therefore, the intermediate break LOCA (IBLOCA) might become the limiting break size of concern in the LOCA analysis. To accomplish this reform of LOCA classification, an extensive understanding of IBLOCA is crucial, and the applicability of safety analysis code should be confirmed. For this purpose, IBLOCA PIRT was developed. The PIRT panel was organized and the general process of the PIRT development established by Wilson and Boyack (1998) was adopted to develop the IBLOCA PIRT. Based on IBLOCA analyses using the safety and performance analyzing code (SPACE), the PIRT panel defined the temporal phases, systems, components and possible phenomena during an IBLOCA. For all possible phenomena, the relative importance and the knowledge level were determined for each component and each phase via discussions among PIRT panel members. For phenomena having relatively high importance, a code validation matrix was also developed. The results of the IBLOCA PIRT will be used to improve the SPACE code for IBLOCA application and resolve future regulatory issues.

Validation and Application of a Code for Three-Dimensional Analysis of Hydrogen–Steam Behavior in a Nuclear Reactor Containment during Severe Accidents

In a pressurized water reactor (PWR) during a loss of coolant accident (LOCA) or a station blackout (SBO) accident, water and steam are released into the containment building. The water vapor mixes with the atmosphere, partially condensing into droplets or condensing on the containment walls. Although a significant amount of water vapor condenses, it coexists with hydrogen generated by the reactor core oxidation. As water vapor condenses, the volume fraction of hydrogen increases, raising the risk of explosion or flame acceleration. As such, water vapor’s behavior directly affects hydrogen distribution. To conservatively evaluate hydrogen safety in a PWR during a severe accident, lumped-parameter codes have been heavily used. As a best-estimate approach for hydrogen safety analysis in a PWR containment, a turbulence-resolved CFD code called contain3D has been developed. This paper presents the validation results of the code and simulation results of hydrogen behavior affected by water vapor condensation and hydrogen removal by passive autocatalytic recombiners (PARs) in the APR1400 containment. The results provide insight into the three-dimensional behaviors of the hydrogen in the containment.

Simulation study on the dynamic characteristics of natural circulation coupled phase change material for equal height difference passive containment cooling system

The passive containment cooling system (PCCS) plays a significant role in ensuring the safety of nuclear power plants. To further reduce peak pressure in the containment during a loss of coolant accident (LOCA), phase change materials (PCMs) are utilized to improve the system’s cooling effectiveness. The dynamic simulation model of PCCS has been developed using Apros software. This article focuses on examining the system’s dynamic characteristics related to various tube startup methods and comparatively analyzing the impact of different PCM thicknesses on heat transfer capacity. The results indicate that post-accident, PCM significantly suppresses peak pressure in the containment. Under the full tube startup method with 4 mm PCM, the peak pressure decreases to 1.75 MPa, whereas under the empty tube startup method, it is reduced to 1.69 MPa. With natural circulation empty tube startup method, PCM can reduce pressure change intensity in the containment but negatively affects natural circulation heat transfer. Although the full tube startup method underperforms the empty tube startup method on suppressing peak pressure, PCM can positively affect natural circulation heat transfer in the early stages of the accident. Empty tube startup offers the advantages of higher heat transfer capacity and lighter weight, making it the optimal choice for ocean nuclear platforms. Additionally, the higher peak power of PCM heat transfer under the empty tube startup method indicates that system can withstand larger thermal shocks.

A study on the impact of using a subchannel resolution for modeling of large break loss of coolant accidents

The nuclear industry is investigating the feasibility of transitioning from 18- to 24-month fuel cycles because of the positive impact it would have on the operational costs for the current fleet of light-water reactors. A challenge to making this change is the increased risk of fuel fragmentation, relocation, and dispersal (FFRD) due to the known potential for ceramic fuel to pulverize into fine particles at the higher discharge burnups. Previous work has been performed by the Nuclear Energy Advanced Modeling and Simulation program to assess FFRD risk in high-burnup cores using the BISON fuel performance code and a coarse mesh thermal hydraulics (T/H) solution for a loss-of-coolant accident (LOCA) using the TRACE system T/H code. Because of the importance of the T/H solution for FFRD assessment, this study seeks to investigate the impact of using higher-fidelity subchannel techniques for modeling of the LOCA transient. CTF was used to model a subregion of a high-burnup core that was depleted by the Virtual Environment for Reactor Applications (VERA) multiphysics core simulator. Both coarse-mesh and pin-resolved models were created in CTF, and a consistent coarse-mesh TRACE model was also developed to allow for benchmarking the code results. A large-break loss-of-coolant accident (LBLOCA) reflood transient was simulated using these three models, and results were compared. Results showed some consistent differences between the CTF and TRACE coarse models, including a higher peak cladding temperature (PCT) prediction in CTF and later quenching in CTF; however, the transient clad temperature behavior was similar, and these differences are likely due to post-critical heat flux heat transfer modeling differences and minimum film boiling temperature model differences. The pin-resolved results indicate that the PCT in the lumped model is often under-predicted by as much as 70 °C and that PCT occurs at a different location than the high-power pin in the assembly. The lumped model predicts a difference of 10 °C or less between the average and hot pins in the assembly, whereas the pin-resolved model predicts a range of over 100 °C. These results indicate that higher-fidelity T/H results may have an impact on predicted core behavior during LOCA, which may be important to consider when assessing FFRD risk.

Method of estimating break size in piping loop systems

The approach for determining the break size of recirculation loops in a multiple-loop power plant in the event of a loss of coolant accident (LOCA) is presented in this study. In this study, the MAAP5 simulation program was used. An approach to measuring the size of a crack or break in the cooling system is the temperature difference between the recirculation loops. This method does not require any additional facilities; it compares the temperatures of the cooling loops to determine which one has a rupture. The best data source was the loop monitoring system, which sends temperature data for analysis to the main control room. A real operating power reactor training simulator and the FSAR are applied to evaluate MAAP5, the methodology's engine. The results of the MAAP5 simulation code were consistent with those of the power plant simulator. Therefore, MAAP5 could produce enough analytical data to create the relationship diagram between temperature difference and break size.The study hypothesized that there exists a maximum value of temperature difference corresponding to each break size and suggested that applying the absolute maximum temperature difference can aid in identifying the break size. This approach proposes an assistive method for determining the size of a fracture or break in the recirculation system by leveraging the temperature difference between each loop.This approach eliminates the need for additional facilities, as temperature data from the recirculation loops can be transmitted to the main control room. After the reactor scram, operators can monitor the maximum temperature differences at the inlet to estimate the break size. Although the fitting curve used to preliminary estimate the Large Break Loss of Coolant Accident break size may overestimate the break size, it still provides valuable insights. This novel tool offers a rapid and comprehensive method for detecting LOCA events in the recirculation loops.

Thermomechanical analysis of SiC-based duplex claddings with varying thickness ratio for accident-tolerant nuclear fuel systems

SiC-based duplex claddings, consisting of monolithic SiC and SiC/SiC fiber composite, are emerging as a promising candidate for accident-tolerant fuel (ATF) systems in nuclear reactors. To analyze the performance of ATFs with SiC-based duplex claddings, a comprehensive computational analysis framework is presented that captures the essential properties and behaviors of the UO2-SiC fuel system. Utilizing a previously developed continuum damage model, the pseudo-ductile behavior of SiC/SiC fiber composites is accurately modelled, connecting damage evolution parameters to instantaneous stiffness matrix degradation. This framework is used to investigate the performance of UO2-SiC fuel rods under normal operating conditions and a typical Loss of Coolant Accident (LOCA) scenario. We assess the effects of the thickness ratio of the monolithic SiC and SiC-based composite layers, as well as pellet-clad cold gap thickness on the failure and leakage probabilities of the cladding. These claddings, with a thickness ratio ranging from 0.25 to 0.75, demonstrated minimal failure and leakage probabilities for both the original and reduced pellet-clad gap thickness (82.5/70 µm). When the gap thickness was further reduced to 57.5 µm, pellet-cladding mechanical interaction was observed and this greatly elevated the failure probability of the MSiC layer, thus giving rise to a loss of hermeticity. This research underscores the significant role of varying individual layer thicknesses in shaping fuel rod safety and offers potential for optimization across diverse operational conditions.

CFD simulation of thermal stratification and mixing in a Nordic BWR pressure suppression pool

Boiling Water Reactor (BWR) employs the Pressure Suppression Pool (PSP) as a heat sink to prevent overpressure of the reactor vessel and containment. Steam can be injected into the PSP through spargers in normal and accident conditions and through blowdown pipes in case of a loss of coolant accident (LOCA). There is a safety limit on the maximum PSP temperature at which such steam injection might cause dynamic loads on the containment structures. The performance of the pool can be affected if thermal stratification is developed when temperature of the hot layer grows rapidly while cold layer remains inactive. Simulation of pool behavior during realistic accident scenarios requires validated models that can sufficiently address the interaction between phenomena, safety systems and operational procedures. Direct modeling of steam injection into a water pool in long-term transients is computationally expensive due to the need to resolve simultaneously the smallest space and time scales of individual steam bubbles and the scales of the whole PSP. To enable PSP analysis for practical purposes, Effective Heat source and Effective Momentum source (EHS/EMS) models have been proposed that avoid the need to resolve steam-water interface. This paper aims to implement mechanistic approaches previously developed by authors for the simulation of transient thermal stratification and mixing phenomena induced by steam injection through spargers in a Nordic BWR PSP. The latest version of the EHS/EMS models using the ‘Unit cell’ approach has been validated against integral effect pool tests and applied to plant simulations. Several scenarios with boundary conditions corresponding to postulated accident sequences were simulated to investigate the possibility of stratification development and the effects of activation of different systems (e.g., blowdown pipes, high momentum nozzle) on the pool behavior.

Prolonged Operation of Hydro-Accumulators in VVER-1000/V446 Considering Flow Restriction Components

Hydro-accumulators, as a passive part of the emergency core cooling system, have been upgraded in different types of VVER-1000s. The number and performance of these systems are modified in newer types of VVER-1000s to operate at different pressures and flow rates. The VVER-1000/446, as a distinguished design from the old generation (VVER-1000/320), utilizes two stages of hydro-accumulators. The second-stage accumulators (SSAs) are dedicated to improving the performance of the VVER-1000/446 under design extension conditions (DECs) and delaying the onset of core damage. During the upgrading of the VVER-1000/446, some orifices have been embedded as flow restriction components (FRCs) on the injection line of the SSAs to increase the water injection time during DECs. This paper aims to assess the performance of the SSAs according to their orifice diameter changes during median-break (MB)– and large-break (LB)–loss-of-coolant accidents (LOCAs). Several cases of FRC arrangements were investigated to demonstrate the core damage times during MB-LOCA (200-mm break) and LB-LOCA (400-mm break) scenarios. The results show that although the current size of the installed FRCs on the VVER-1000/446 is quite suitable for the MB-LOCA, it is not a proper choice for the LB-LOCA. Therefore, this demonstrates that the core damage time is highly dependent on the FRC size arrangement. This paper indicates an arrangement of FRCs that can modify the plant’s response in both MB and LB LOCAs without active injections. Therefore, as an additional finding, the sequential arrangement of FRCs can increase plant resilience against a broad range of LOCAs without operating active emergency cooling systems.

Revisiting the thermodynamic properties of the ZrCr2 Laves phases by combined approach using experimental and simulation methods

A comprehensive study of ZrCr2 Laves phases is important for both fundamental research and technological applications. The ZrCr2 compound is a typical precipitate observed in Zr-based alloys including the newly developed Cr-coated Zr cladding known as Accident Tolerant Fuel cladding. The brittle behavior of the Laves phases and the low melting point at the interface zone between the precipitates and the Zr matrix are considered as governing the thermomechanical behaviors of these newly developed nuclear fuel claddings for normal operating conditions but more dramatically in case of Loss Of Coolant Accident (LOCA) conditions. The aim of the present study is to investigate the various ZrCr2 Laves polymorphs and their thermodynamic features particularly where conflicts exist in the literature. Based on diffraction experiments of annealed samples, it has been established that the transformation from the C14 high-temperature form to the C15 low-temperature form is of displacive nature without the C36 polymorph forming as an intermediate phase. The stable and metastable domains of C14 and C15 and their thermodynamic properties have been determined. The specific heat of both C14 and C15 types was measured over a wide range of temperatures from 2 to 1063 K by coupling relaxation calorimetry and DSC. The experimental data were fitted using a modified Einstein model. The room temperature entropies of the C14 and C15 phases were evaluated. The enthalpy of formation was measured for the first time using drop solution calorimetry in liquid Al at 1173 K in a Tian–Calvet calorimeter. The experimental value is in good agreement with Density Functional Theory (DFT) calculations. Finally, all these new results are discussed regarding the abundant literature on the ZrCr2 Laves phases.

NRC’s methodology to estimate fuel dispersal during a large break loss of coolant accident

Recent experimental findings indicate that under certain transient conditions, high-burnup fuel operating at sufficiently high power can fragment and escape from burst fuel cladding, as discussed in the Nuclear Regulatory Commission’s Research Information Letter (RIL) 2021–13. The escaped fuel fragments could subsequently be distributed throughout a reactor coolant system (RCS), potentially challenging core coolability during a loss of coolant accident (LOCA). Consequently, the NRC has developed a methodology to estimate the mass of fuel that could be dispersed during a large-break LOCA. The methodology involves several NRC-sponsored computer codes, including SCALE/Polaris for lattice physics, PARCS/PATHS for full-core neutronics, TRACE for core and RCS thermal hydraulics, and FAST for steady-state and transient fuel performance calculations. Output from FAST is combined with the models in RIL 2021–13 to estimate fuel dispersal to the coolant.NRC staff have exercised this methodology using a prospective Westinghouse four loop core design with high burnup and extended enrichment that was obtained from the Department of Energy’s NEAMS program. Results demonstrate that the NRC methodology can be exercised to provide fuel dispersal estimates that can subsequently be used to study the potential consequences of FFRD. However, this exercise has also identified several areas for improvements to both the codes and to the modeling approaches used here, including the pin power reconstruction methods in PARCS, the modeling approach used to connect the Cartesian and cylindrical vessel components in TRACE, transient fission gas release and fuel rod upper plenum models in FAST, and tighter coupling between FAST and TRACE.