Loss Of Coolant Accident Event Research Articles

Fuel integrity evaluation for the shutdown BWR-4 plant during LOCA event

Two BWR units in the Chinshan stopped operation since 2017. However, the spent fuel assemblies still reside in the reactor core, and the plant staffs defined this configuration as the Mode 5 condition. If a Loss-of-Coolant Accident (LOCA) occurs in this condition, the water level will drop and then maintain at approximately 2/3 of the height of the reactor core. The upper 1/3 of the fuel will be exposed to air, and then the fuel cladding temperature will increase significantly. This study use RELAP5 to investigate the effect of different Emergency Core Cooling System (ECCS) to mitigate the LOCA. It is found that the core spray (CS) system can sufficiently suppress the fuel temperature. However, the temperature increase of the uncovered fuel becomes larger if the low pressure coolant injection (LPCI) system which does not provide direct core injection is operable. Based on the results, the peak cladding temperature of both cases are still below the acceptance criteria of 600 °C.

Quenching Experiments with CrAl-coated Zircaloy Cladding in Reflooding Water Flows

A quenching experiment is performed to investigate the heat transfer characteristics and cooling performance of CrAl-coated Zircaloy (Zr) cladding in a water flow. The CrAl-coated Zr cladding is one of the accident tolerant fuels for light water reactors. The uncoated Zr cladding is also used in this quenching experiment for comparison. This experiment simulates reflood quenching of fuel rod during loss of coolant accident (LOCA) in nuclear power plant. The test conditions were determined to represent the peak cladding temperature, the coolant subcooling and the reflood velocity in the event of LOCA. The flow visualization showed the film boiling during early stage of reflood quenching and the transition to nucleate boiling. The film layer decreases as the coolant subcooling increases and becomes wavy as the reflood velocity increases. The CrAl-coated Zr cladding showed more wavy and thinner film than the uncoated Zr cladding. The rewetting temperature increases as the initial wall temperature and/or the coolant subcooling increases. The quench front velocity increases significantly as the coolant subcooling increases. The reflood velocity has a negligible effect on rewetting temperature and quench front velocity.

Transient System Thermal-Hydraulic Assessment of Advanced Uranium- and Thorium-Based Fuel Bundle Concepts for Potential Use in Pressure Tube Heavy Water Reactors—II: Full-Core Analyses

The use of advanced uranium-based and thorium-based fuel bundles in pressure tube heavy water reactors (PT-HWRs) has the potential to improve the utilization of uranium resources while also providing improvements in performance and safety characteristics of PT-HWRs. Earlier lattice physics and reactor core physics studies have demonstrated the feasibility of using such advanced fuels; however, thermal-hydraulic (T-H) studies are required to confirm that these advanced fuels will have adequate T-H safety margins. Preliminary system T-H transient simulations have been carried out for a 700-MW(electric)–class PT-HWR in a postulated loss-of-coolant accident (LOCA) using the CATHENA code. One purpose of this work was to demonstrate that such simulations of a PT-HWR filled entirely with advanced fuels could be set up and executed successfully in a CATHENA transient simulation model. The other purpose was to evaluate the peak sheath and peak fuel centerline temperatures during a LOCA to perform an analysis that compares the relative performance of each of the proposed advanced fuels. System T-H simulations with CATHENA were performed to model a postulated LOCA event with a 20% inlet header break in a typical 700-MW(electric)–class PT-HWR using two types of advanced uranium-based and thorium-based fuel bundles in modified 37-element and 35-element geometries. Calculations were also performed for a PT-HWR using conventional natural uranium fuel in 37-element fuel bundles for comparison. In the event of a LOCA, there is a drop in the primary circuit pressure. It is assumed that there is a 2-s delay between the signal of the low primary pressure and the tripping of the reactor. When the reactor trips, the shutdown rods are inserted. The reactor trip is followed by the activation of the emergency core cooling system, which occurs 30 s after the LOCA starts, with a trip signal on the boiler crash cooling. Simulation results for the LOCA demonstrated that the peak fuel centerline temperatures (ranging from 1822°C to 2183°C) were several hundred degrees below the expected melting point of UO2 (~2865°C). Simulations also demonstrated that the peak sheath temperatures for the advanced fuel concepts ranged from 1177°C to 1204°C, which are lower than that with conventional NU fuel in 37-element fuel bundles. Thus, the system T-H analysis of the relative results provides confidence in the proposed advanced uranium-based and thorium-based fuel concepts for potential use in PT-HWRs.

Transient System Thermal-Hydraulic Assessment of Advanced Uranium- and Thorium-Based Fuel Bundle Concepts for Potential Use in Pressure Tube Heavy Water Reactors—I: Two-Channel Analyses

The use of advanced uranium-based and thorium-based fuel bundles in pressure tube heavy water reactors (PT-HWRs) has the potential to improve the utilization of uranium resources while also providing improvements in performance and safety characteristics of PT-HWRs. Previous lattice physics and core physics studies have demonstrated the feasibility of using such advanced fuels; however, thermal-hydraulic (T-H) studies are required to confirm that these advanced fuels will have adequate T-H safety margins. Preliminary system T-H transient simulations have been performed for a 700-MW(electric)–class PT-HWR in a postulated loss-of-coolant accident (LOCA) using the CATHENA code. One purpose of this work was to demonstrate that such simulations of a PT-HWR with advanced fuels could be set up and executed successfully in a CATHENA transient simulation model. The other purpose was to evaluate the peak fuel sheath and fuel centerline temperatures in two designated fuel channels containing advanced uranium-based or thorium-based fuel during a LOCA transient event. In the CATHENA simulation models, a PT-HWR core is fueled with conventional 37-element natural uranium fuel bundles in 378 out of 380 fuel channels while two designated fuel channels, the channel with the highest total power and the channel containing the bundle with the highest power level, are filled with various types of advanced fuels. Results indicate that setting up these models is feasible and that the predicted peak fuel centerline temperatures and peak sheath temperatures for the advanced fuel channels are well below the fuel melting points and the rapid oxidation temperature for the Zircaloy-4 sheath/clad (~1200°C), respectively. These preliminary results provide confidence that the advanced fuels will likely have adequate T-H safety margins in a transient LOCA event in a PT-HWR. These results set the stage for more detailed and comprehensive system T-H models of PT-HWRs fueled entirely with advanced uranium-based or thorium-based fuels.

Thermal-Hydraulic and Engineering Evaluations of New LOCA Testing Methods in TREAT

Design evaluations and thermal-hydraulic predictions were undertaken to compare three candidate options for loss-of-coolant accident (LOCA)–testing capabilities at the Transient Reactor Test facility (TREAT). These options included a small water capsule, a large natural circulation capsule, and a forced-convection water loop. All options were found to create reasonable approximations of radial temperature gradients in fuel specimens prior to blowdown to a secondary tank. Given its minimalist form factor, the small water capsule was predicted to perform remarkably well in mimicking the thermal response of a 20-pellet fuel rodlet during a postulated LOCA event, while sensitivity studies showed versatility in adjusting conditions for specific test phenomena priorities. The large natural circulation capsule predictions showed representative capsule pressure evolution but indicated it would likely struggle to capture the desired fuel temperature response in the blowdown transition, primarily due to nonrepresentative cooling as its tall water column was discharged. These results were compared to previous water loop evaluations that had shown a capability for representing ideal test conditions, especially for longer rodlet test specimens in small bundles, while being achievable in a self-contained form factor to facilitate test logistics. Considering the broader cost and capability spectrum needed to support fuel safety research, these efforts resulted in a recommendation to pursue the small water capsule for near-term data needs specific to high-burnup fuel fragmentation, while continuing detailed development of the water loop to address ultimate data needs.

Analysis of stress applied to a ruptured cladding tube under horizontal vibration

ABSTRACT After the termination of a loss-of-coolant accident (LOCA), the reactor continues to be cooled for a long term until fuel assemblies are withdrawn from the reactor core. The fuel cladding tube degrades in strength due to high-temperature oxidation during a LOCA event. It is important to confirm that fuel rods exposed to LOCA conditions can withstand earthquakes during the long-term cooling in terms of preserving the coolable geometry of the reactor core. Finite element method analyses were performed to estimate the deformation of fuel rods in a fuel assembly under vibrations simulating an earthquake as well as the stress applied to the fuel cladding tube with a rupture opening. The localized stress at the rupture opening in the analyses was compared with the strength assessed through bending tests of the cladding tube samples that were ruptured and oxidized to less than 15% equivalent cladding reacted (ECR) in advance. As the result, the fuel rods are expected to be prevented from fracture due to bending at earthquakes during the post-LOCA cooling unless the oxidation of cladding tubes exceeds the limit defined in the current Japanese LOCA criteria, 15% ECR and a deflection of the fuel rodexceeds approximately 40 mm.

The impact of emergency operating safety procedures on mitigation the nuclear thermal power plant severe accident

The most important public concerns is the safety of nuclear power reactors. It is extremely important to protect the public from the nuclear radiation exposure in the undesired case of a nuclear power plant (NPP) accident. The proper performance of the safety systems of nuclear power plants is one of the most highly important concerns which enhances the use of nuclear energy options. This research paper focuses on the response and performance of the emergency safety systems of a pressurized water reactor (PWR) during loss of coolant accident (LOCA) for a selected reference PWR. The simulation depicts the reactor behavior during hot leg loss of coolant accident event and the leakage area is 200 cm2. This was accomplished by modeling the PWR and the passive safety systems employed using PCTran code which is designed to describe the emergency core cooling system (ECCS) response and core behavior under severe accident conditions. The ECCSs are designed to cool down the reactor core and to prevent core damage. Results of the core temperature distribution, reactor coolant inlet and outlet temperatures, reactor coolant mass flow rate, and other parameters have been compared with the reference data and demonstrated to be in good agreement with each other. This study demonstrates that the PCTran model is capable of contributing to the process of safety analysis of the NPP.

Activity emission model for VVER 440/213 reactor LOCA

A VVER-440/213 activity emission model has been developed in order to describe the activity release process in a case of a LOCA (Loss Of Coolant Accident) event. The model uses calculated and measured data. The pressure of the hermetic rooms in a LOCA event is calculated, the leakage from the hermetic rooms at certain pressure is measured, the model also considers the starting time of the fuel damage. For the estimation of the fission products release from the fuel damage the EUR conservative approach was used. The model also deals with filtering of the emitted fission products and considers the activity changes for the given nuclides by time. Three nuclides were used for our calculations. One noble gas 133Xe the behaviour of which is supposed to be similar to all other noble gases. The other two nuclides are very important from radiation protection point of view. For learning about short term contamination the 131I was chosen, for long term consequences the 137Cs was taken.

Full Field Measurements used for Assessing Industrial Issues-Two Examples

This paper presents two case studies in which the use of full field measurements could provide valuable information to allow the development of R&D methods having the potential of assessing industrial issues.The first example is related to the study of the fracture properties of the nuclear fuel rods after a postulated Loss of Coolant Accident (LOCA) event. It will be shown that 3-D Digital Image Correlation strain measurements performed on a non-conventional mechanical test (expansion due to compression) can help to assess failure properties and derive fracture criteria of highly oxydized and hydrided Zircaloy cladding. These fracture criteria are therefore used in the Safety Demonstrations due conducted by EDF in reply to the French Safety Authority.The second example is related to the investigation of potentially brittle failure of an irradiation aged nuclear reactor pressure vessel (RPV) during a Loss of Coolant Accident event. As a matter of fact, the reactor pressure vessel is one of the non replaceable components of a nuclear power plant. Consequently, the study of the RPV ageing is of major importance for assessing plants lifetime. One consequence of the RPV ageing is the increasing risk of brittle fracture during postulated LOCA events. As a first step to better characterize experimentally the embrittlement of RPV steel, the second example presents in-situ strain heterogeneities in the plastic response of bainitic-ferritic 16MND5 steel, with respect to crystalline phases and orientations. These local observations are then used by a specific identification procedure, in order to obtain experimentally the crystal plasticity hardening flow behaviour of RPV steel.

ICONE19-43390 INVESTIGATIONS ON PRESSURE DROP CHARACTERISTICS AND DISPOSAL OPTIMIZATION OF CONFLUX CHANNELS OF CONTAINMENT SUMP STRAINERS

In the event of LOCA (Loss-of-Coolant Accident) or HELB (High Energy Line Break) within the containment of a light-water reactor (LWR), some debris will be transported to and accumulate on the containment sump strainers by the steam/water flows induced by the break and by the containment sprays. And pressure loss will be introduced by the accumulated debris, which will decrease the available net positive suction head (NPSH) of circulating pumps of emergency core cooling system (ECCS) and containment spray system (CSS). Then cavitations may occur to induce the failure of ECCS and CSS pumps. Consequently it is necessary to carry out calculation and experiment to investigate the pressure drop characteristics of containment sump strainers. In the event of LOCA or HELB, the pressure drop of containment sump strainers is mainly composed of the pressure drop introduced by debris bed, strainer screen and the conflux channels. The authors established 3-D computational simulations to conduct investigations on the velocity field distribution and the pressure field distribution of water flow in the conflux channels. By analyzing pressure drop characteristics, the design optimization and disposal optimization were brought forward to minimize the pressure loss.

Control rod drop analysis by finite element method using fluid–structure interaction for a pressurized water reactor power plant

The control rod drop analysis is very important for safety analysis. For seismic and loss of coolant accident event, the control rod assemblies shall be capable of traveling from a fully withdrawn position to 90% insertion without any blockage and within specified time and displacement limits. The analysis has been executed by analytical method using in-house code. In this method, several field data are needed. These data are obtained from nuclear, thermal–hydraulic and mechanical design groups, peculiar codes, those work groups need to cooperate together. Following the enhancement of a computer and development of the multi-physics analysis code, a new method for the control rod drop analysis is proposed by finite element method. This analysis model incorporates the structure and fluid parts, termed as a fluid and structure interaction (FSI). Because a control rod is submerged inside a guide tube of a fuel assembly, the FSI boundary condition is applied. In this model, it is assumed that the fluid is incompressible laminar flow. The structures are modeled with the solid elements because there is no deformation due to the fluid flow. The analysis two-dimensional plane model is created in the analysis with considering an axi-symmetric geometry. Therefore, the proposed analysis model will be very simple and the design data from other fields will be unnecessary. The analysis results are compared with those of the in-house code, which have been used for a commercial design. After validation, it is found that the present analysis gives a useful tool in the design of the control rod and fuel assembly.