Simulated Loss Of Coolant Accident Research Articles

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.

Oxidation characteristics of Fe22Cr5Al3Mo-xNb alloys in high temperature steam

The oxidation behaviors of Fe22Cr5Al3Mo-xNb (x=0.5, 1.0, 2.0 wt%) alloys under simulated loss of coolant accident (LOCA) conditions at 1000 and 1200 ℃ were investigated. The results revealed that fine granular oxides and a ridged oxide film were formed on the FeCrAl alloys at 1000 and 1200 ℃, respectively. The oxide film was composed of α-Al2O3, α-Cr2O3 and a trace amount of FeO at 1000 ℃ but only consisted of α-Al2O3 at 1200 ℃ due to volatilization of Fe and Cr oxides. Nb addition decreased the weight gains and improved the oxidation resistance of FeCrAl alloys under simulated LOCA condition.

Coupling Gothic and RELAP5 for Passive Containment Cooling Modeling

GOTHIC and RELAP5 have been coupled to model the containment and passive containment cooling system (PCCS) for the Korean advanced containment designs. In the coupled system, GOTHIC models the containment and the outer shell (mid tube wall to outside surface) of the PCCS heat exchanger tubes and RELAP5 models the inner shell (inside surface to mid tube wall) and the coolant loop to the external heat sink. The coupling approach leverages the modeling capabilities of RELAP5 for piping load analysis and the capabilities of GOTHIC for containment and heat/mass transfer with noncondensing gases. With the coupled model, it is possible to apply the thermal-hydraulic load analysis on the PCCS supply and return piping considering the containment conditions predicted by GOTHIC during a loss-of-coolant accident (LOCA). This paper describes the coupling approach, a coupling dynamic linked library for GOTHIC, modifications to RELAP5, and verification of the coupling. Last, demonstration results from a LOCA simulation with four PCCS trains is provided and the results of the GOTHIC/RELAP5–coupled model are compared to a GOTHIC-only result, where GOTHIC was used to model both the containment and the PCCS.

RELAP5 Modeling of a Siphon Break Effect on the Brazilian Multipurpose Reactor

This work presents the thermal-hydraulic simulation of the Brazilian Multipurpose Reactor (RMB) using the RELAP5/Mod3 code. The RMB will provide Brazil with a fundamental infrastructure for the national development on activities of the nuclear sector in the areas of social, strategic, industrial applications and scientific and technological development. A RELAP5/Mod3 code model was developed for thermal-hydraulic simulation of the RMB to analyze the phenomenology of the Siphon Breakers device (four flap valves in the cold leg and one open tube for the atmosphere in the hot leg) during a Loss of Coolant Accident (LOCA) at different points in the primary circuit. The Siphon Breaker device is an important passive safety system for research reactors in order to guarantee the water level in the core under accidental conditions. Different simulations were carried out at different location in the Core Cooling System (CCS) of the RMB, for example: LOCA before the CCS pumps with and without pump trip and LOCA after the CCS pumps and the heat exchanger. In all RELAP5/Mod3 code simulations, the Siphon Breaker device's performance after a LOCA was effective to allow enough air to enter the outlet pipe of the CCS in order to break the siphon effect and preventing the pool level from reaching the riser (chimney) and the RMB core discovering. In all cases, the reactor pool level stabilized at about 5.5 m after the end of the LOCA simulation and the fuel elements were kept underwater and cooled.

Axial strength of Zircaloy-4 samples with reduced thickness after a simulated loss of coolant accident

To investigate wall-thinning impact on axial load resistance of Zircaloy-4 cladding rods after a LOCA transient, axial tensile samples have been machined on as-received tubes with reduced thicknesses between 370 and 580 μm. After high temperature oxidation under steam at 1200 °C with measured ECR ranging from 10 to 18% and water quenching, machined samples were axially loaded until fracture. These tests were modeled using a fracture mechanics approach developed in a previous study. Fracture stresses are rather well predicted. However, the slightly lower fracture stress observed for wall-thinned samples is not anticipated by this modeling approach. The results from this study confirm that characterizing the axial load resistance using semi-integral tests including the creep and burst phases was the best option to obtain accurate axial strengths describing accurately the influence of wall-thinning at burst region.

A study on the robustness of neural network models for predicting the break size in LOCA

The diagnosis of loss of coolant accidents in nuclear reactors has attracted a great deal of attention in condition monitoring of nuclear power plants given that the health of the cooling system is crucial to the nuclear reactor's stable operation. Many different types of neural networks have commonly been applied to loss of coolant accident diagnosis. It is important to select a suitable architecture for the neural network that delivers robust results, in that the predicted break size is deemed to be accurate even for break sizes that are not included in the training data sets. The robustness metric proposed in our previous work is applied to compare the robustness of different diagnostic models. The data used for training these models consists of a number of time-series data sets, each for a different break size, with the transient behavior of different measurable variables in the coolant system of a nuclear reactor, following a simulated loss of coolant accident in a high-fidelity simulator. Given the simulation data for different break sizes, four different neural network architectures are investigated and their properties are compared and discussed. These models include a fully-connected multilayer perceptron with one hidden layer, a multilayer perceptron with one hidden layer that is pruned using the optimal brain surgeon algorithm, a fully-connected multilayer perceptron with two hidden layers, and a group method of data handling neural network. In this paper, an interpolation pre-processing method is investigated and shown to be effective to further improve the capability of neural networks for robustly predicting the break size of a loss of coolant accident. Both linear interpolation and cubic spline interpolation are studied as alternatives for the pre-processing approach. The performance of models developed with and without interpolation pre-processing are compared with the previously proposed robustness metric. Moreover, three blind cases are introduced to evaluate and compare the performance of the diagnostic models. Finally, a combined diagnostic model is proposed based on three different architectures to obtain high prediction accuracy and good robustness.

Reflooding investigation on light water reactor model fuel assemblies and debris under the loss of coolant accident simulation

The reflooding investigations of large-scale model fuel assemblies and sphere beds under Design Basis Accident and Beyond Design Accident of Light Water Reactors WWER and PWR are presented. The generalizing correlation on the quench front velocity for model fuel rod assemblies and sphere beds was developed on the base of heat balance and Leidenfrost temperature correlation. It is shown that reflooding duration calculation results of model fuel assemblies and sphere beds are in good agreement with the experimental data.

Study of clad ballooning and rupture behaviour of Indian PHWR fuel pins under transient heating condition in steam environment

This paper presents the results of the investigation on the deformation and rupture characteristics of Indian pressurized heavy water reactor (IPHWR) fuel pins under simulated loss of coolant accident (LOCA) condition in steam environment. Transient heating experiments were carried out on single fuel pin internally pressurized with argon gas in the range 3–70 bar. Effect of internal pressure on burst temperature, influence of burst temperature on the circumferential strain and rupture opening area were also studied. Two circumferential strain maxima at the burst temperatures of 740 & ∼979 °C and a minimum at the burst temperature of ∼868 °C were observed. It was found that oxidation had considerable effect on the burst behavior. Test data were used to derive a direct empirical correlation for burst stress exclusively as a function of temperature. The ballooning and rupture behaviours in steam and argon environments have been compared. Experimental data were examined against various correlations using Erbacher equation and author's previous correlation in argon.A second burst correlation has also been developed combining the equation in argon from the previous work of the authors and an exponential factor with oxygen content as a parameter assuming the burst stress to be a function of both temperature and oxygen concentration. The burst temperatures predicted by this empirical correlation are in good agreement with the test data.

BWR loss of coolant accident simulation by means of RELAP5

Application of best estimate codes and methodologies for LOCA simulation can provide the high level of details of the real plant response necessary to assess the adequacy of an emergency core cooling system.The current KKM RELAP5 model developed for the KKM plant (Mühleberg nuclear power station) has been demonstrated to be valid for simulating a wide range of transients including loss of coolant type accidents. Using this model, KKM LOCA analyses have been performed for the recirculation line break. The simulations have been done for the full break size spectrum. The sensitivity to the most important physics phenomena, e.g. CCFL, and model parameters, e.g. nodalization, has also been assessed.Results from the KKM RELAP5 LOCA analyses have been compared to the KKM LOCA results obtained with TRACG code by GEH. The RELAP5 and the TRACG models were both validated and showed very good agreement with the measured plant data. Regarding the LOCA calculations, the plant behavior and PCT results predicted by the RELAP5 model are consistent with those reported by GEH using the TRACG model. The differences observed can firstly be attributed to the diversity in the basic modeling of physical phenomena in the two codes. In addition, differences may originate from varying details in modeling the KKM plant and core parameters. However, considering that completely independent codes and models of the KKM plant predict similar PCT and plant behaviors during a LOCA, both KKM RELAP5 and KKM TRACG models can confidently be used for comparison against safety criteria.

An attempt for a unified description of mechanical testing on Zircaloy-4 cladding subjected to simulated LOCA transient

During a Loss Of Coolant Accident (LOCA), an important safety requirement is that the reflooding of the core by the emergency core cooling system should not lead to a complete rupture of the fuel rods. Several types of mechanical tests are usually performed in the industry to determine the degree of cladding embrittlement, such as ring compression tests or four-point bending of rodlets. Many other tests can be found in the open literature. However, there is presently no real intrinsic understanding of the failure conditions in these tests which would allow translation of the results from one kind of mechanical testing to another. The present study is an attempt to provide a unified description of the failure not directly depending on the tested geometry. This effort aims at providing a better understanding of the link between several existing safety criteria relying on very different mechanical testing. To achieve this objective, the failure mechanisms of pre-oxidized and pre-hydrided cladding samples are characterized by comparing the behavior of two different mechanical tests: Axial Tensile (AT) test and “C”-shaped Ring Compression Test (CCT). The failure of samples in both cases can be described by usual linear elastic fracture mechanics theory. Using interrupted mechanical tests, metallographic examinations have evidenced that a set of parallel cracks are nucleated at the inner and outer surface of the samples just before failure, crossing both the oxide layer and the oxygen rich alpha layer. The stress intensity factors for multiple crack geometry are determined for both AT and CCT samples using finite element calculations. After each mechanical test performed on high temperature steam oxidized samples, metallography is then used to individually determine the crack depth and crack spacing. Using these two important parameters and considering the applied load at fracture, the stress intensity factor at failure is derived for each tested sample. This procedure provides an assessment scheme to determine experimentally the fracture toughness of the prior-β region in the mid-wall of the oxidized samples. The obtained fracture toughness for CCT and AT samples are thus compared, confirming that the linear elastic fracture mechanics is a relevant tool to describe the strength of LOCA embrittled cladding alloys.

Study of oxide and α-Zr(O) growth kinetics from high temperature steam oxidation of Zircaloy-4 cladding

Oxidation kinetics of Zircaloy-4 cladding of fuel pins of Indian pressurized heavy water reactors (IPHWRs) under a simulated loss of coolant accident (LOCA) condition was investigated. The kinetic rate constants for the oxide and oxygen stabilized α-Zr phase growth were established from the isothermal metal-steam reaction at high temperatures (900–1200 °C) with soaking periods in the range of 60–900 s. Oxide and α-Zr(O) layer thickness were measured to derive the respective growth rates. The observed rates obeyed a parabolic law and Arrhenius expressions of rate constants were established. Percentage equivalent clad reacted (%ECR) was calculated using Baker-Just equation. Hydrogen estimation was carried out on the oxidized samples using inert gas fusion technique. The hydrogen pick up was found to be in the range 10–30 ppm. The measured values of oxide and α-Zr(O) layer thickness were compared with the results obtained using OXYCON, an indigenously developed model. The model predicts the oxide growth reasonably well but under predicts the α-Zr(O) growth significantly at thickness values higher than 80 μm.

A STUDY ON THE AGING DEGRADATION OF ETHYLENE-PROPYLENE-DIENE MONOMER (EPDM) UNDER LOCA CONDITION

The aging degradation and lifetime assessment of a domestic class 1E Ethylene-Propylene-Diene-Monomer (EPDM), which is a popular insulating elastomer for electrical cables in the nuclear power plants, were studied for equipment qualification verification under the Loss of Coolant Accident (LOCA) conditions. The specimens were acceleratively aged, underwent a LOCA environment, as well as tested mechanically, thermo-gravimetrically, and spectroscopically according to the American Society of the Testing of Materials (ASTM) procedures. The tensile test results revealed that the elongation at break gradually decreased with an increasing aging temperature. The lifetime of EPDM aged isothermally at 140℃ was 1,316 hours and reduced to 1,120 hours after experiencing the severe accident test. The activation energies of the elongation reduction were 1.10 ± 0.196 eV and 0.93 ± 0.191 eV before and after the LOCA condition, respectively. The TGA test results also showed that the activation energy of the aging decomposition decreased from 1.35 eV to 1.02 eV after undergoing the LOCA environment. Although the mechanical property changes were discernibly observed during the aging process, along with the LOCA simulation, the FT-IR analysis showed that the spectroscopic peaks and their intensities did not alter significantly. Therefore, it can be concluded that the degradation of the domestic class 1E EPDM due to aging can be tolerable, even in severe accident conditions such as LOCA, and thus it qualifies as a suitable insulating material for electrical cables in the nuclear power plants.

Embrittlement behavior of zircaloy-4 cladding during oxidation and water quench

Simulated loss of coolant accident (LOCA) tests on Zircaloy-4 cladding were carried out to evaluate the thermal shock property during the injection of emergency core coolant. A Zircaloy-4 specimen was oxidized in a steam environment between 1000 and 1250°C followed by a flooding of the cooling water. After the test, the ductility of the thermally embrittled specimen was measured by a ring-compression test and a microstructural analysis was carried out. The results showed that the threshold equivalent cladding reacted (ECR) value to cause a failure was higher than the conventional 17% criterion calculated by the Baker–Just equation. A residual metal thickness under 0.3mm as well as a ring-compression ductility below 0.2mm for a fracture is effective to assess the thermal shock embrittlement of Zircaloy-4 in an axially unrestrained, or even in a restrained condition.

Deformation of zircaloy-4 cladding during a LOCA transient up to 1200 C

The high temperature ballooning properties of Zircaloy-4 cladding were evaluated at the simulated loss of coolant accident (LOCA) conditions. Two types of tests were performed. The first one is an isothermal-ballooning test in which a Zircaloy-4 cladding with inner gas pressures between 10 and 60 MPa was ballooned in a high-temperature steam environment. The second one is a transient ballooning test in which an initially pressurized cladding was heated until rupture. Burst elongation and burst temperature were measured after the test. The results showed that the phase transformation between a and β has a great influence on the ballooning behavior in that the a phase acts in a transgranular slip mode and the β phase in intergranular sliding mode. In the case of isothermal ballooning, oxidation has a great effect on the cladding deformation by reducing the burst elongation at high temperature.

Nuclear qualification of PVC insulated cables

Cables insulated with eleven different PVC compounds representing fire-retardant, non-fire retardant, 60, 75 and 90/spl deg/C temperature ratings were evaluated. The specimens were subjected to accelerated thermal, radiation, and sequential radiation and thermal exposure simulating 15, 20, 30 and 40 yr in-service environments in Ontario Power Generation Nuclear Plants. The radiation level was limited to 12 to 30 Mrad. Two sets of samples were evaluated. The first contained 15 cm insulation specimens in tubular form with the conductors removed. The material performance of the insulations was assessed using the 15 cm specimens during the various stages of aging exposures by conventional elongation measurement. The second set was made from 4.3 m single or twisted pair of cable samples (unjacketed) wound on mandrels. The cable samples were sequentially irradiated and thermally aged and subjected to an additional 5 Mrad accident radiation exposure, followed by a simulated loss of coolant accident (LOCA). Similarly thermal only aged specimens with 2 Mrad of background radiation representing powerhouse environment were subjected to a main steam line break (MSLB) steam test. During the steam tests, samples were energized and their insulation resistance (IR) values were continuously monitored using an on-line setup and periodically with a megohm meter at 500 V/sub dc/. The results from the insulation samples were compared with the electrical performance of the cable samples wound on mandrels that had been subjected to the LOCA/MSLB steam test. The test results indicate that elongation values were not a good indicator of electrical performance of PVC insulated cables subjected to radiation and thermal environment. With remaining elongation values ranging from 50 to 150% absolute, four out of eleven compounds performed poorly electrically during the steam test. All thermal only aged specimens functioned satisfactorily. In addition, two additional compounds rated for 60/spl deg/C applications were investigated and functioned unsatisfactorily despite high elongation values. It was found that specific additives were responsible for poor electrical behavior. During radiation exposure ionic salts were formed as a result of break down of PVC and of some plasticizers. Upon thermal aging or later in the steam test these ionic compounds migrated to the conductor surface. Then, under high temperature and pressure steam conditions, the salts were dissolved and created an electrical leakage path. Similarly, in thermal only aged specimens, breakdown products of specific plasticizers caused poor electrical behavior.

Condition assessment of cable insulation systems in operating nuclear power plants

Four crosslinked polyethylene (XLPE) and ethylene propylene rubber (EPR) insulated cables were subjected to accelerated thermal, radiation, and sequential radiation and thermal exposure to simulate 20, 30, and 40 yr in-service environments in Ontario Hydro Nuclear Plants. Two sets of samples were evaluated. The first set contained 15 cm cable specimens in tubular form with the conductors removed, while the second set was made up of 4.3 m cable samples wound on mandrels. The cable samples were sequentially irradiated and thermally aged and subjected to an additional accident radiation exposure followed by a simulated loss of coolant accident (LOCA)/main steam line break (MSLB) steam test. During the steam test, samples were energized and their electrical characteristics continuously monitored. The material performance of the insulation was assessed using the 15 cm specimens during the various stages of aging exposure by conventional elongation measurement and by the use of micro-specimens with differential scanning calorimeter (DSC), oxidation induction time (OIT) and infrared carbonyl absorption techniques. The results from these samples were compared with the electrical performance of the cable samples wound on mandrels that had been subjected to the LOCA/MSLB steam test. The test results indicate that with the use of micro-specimens meaningful changes were obtained until the elongation values of the samples were reduced to 50% absolute. Electrical test data obtained for XLPE and EPR during the steam tests indicate that 50% absolute elongation values provide sufficient margin to withstand additional accident radiation, and to function electrically during design basis events.

Qualification of Cables to IEEE Standards 323-1974 and 383-1974

Wire and Cable manufacturers generally qualify products for class IE application by envelope type testing to user specifications and environmental conditions recommended by IEEE Standards 323-1974 and 383-1974. The General Electric Wire and Cable Business Department recently completed two such qualification programs. Cable constructions tested were 600V control cables and 600V, 2KV, and 15KV power cables insulated with flame resistant mineral filled crosslinked polyethylene. The 15KV samples included taped field splices. To determine the ability of samples to withstand a loss of coolant accident (LOCA) both early in and at the end of design life, both thermally aged and unaged cables were irradiated and subjected to a 100 day plus 10 day margin simulated LOCA test. Samples were heat aged to a time-temperature condition in excess of 40 years of service life. The aging criterion was determined from experimental data fitted to an Arrhenius curve by the Regression Method of IEEE Standard 101-72. All test specimens were irradiated using a Cobalt 60 source to a total integrated dose of 220 Mrad. The simulated LOCA conditions of one test program followed the combined requirements for Boiling Water Reactors and Pressurized Water Reactors. In the second test program, the steam pressure-temperature profile included a simulated main steam line break. Test specimens were wrapped on grounded mandrels and were electrically loaded throughout the simulated LOCA tests. After completion of environmental testing, samples were subjected to the IEEE 383 simulated post-LOCA test.

Simulated and Simultaneous Loss-of-Coolant Accident Testing of Protective Coatings for the Nuclear Industry

An apparatus was designed and constructed to test the performance of coatings under conditions of high-pressure steam and radiation that might exist under a loss-of-coolant accident (LOCA). Results from the ''simultaneous'' exposure of coatings to high-pressure steam and radiation are compared to results obtained from the conventional ''simulated'' test procedures. Coating systems were selected that had a history of performing well under simulated LOCA conditions and included zinc-based, epoxy, and phenolic primers with phenolic and modified phenolic topcoats. Coatings were exposed to $sup 60$Co radiation doses in the range of 10$sup 8$ to 10$sup 9$ rad. The study showed that the conventional simulated LOCA conditions were more severe on the coatings than those tested under the simultaneous exposure to high-pressure steam and $sup 60$Co radiation. It was concluded that coatings that satisfactorily passed the simulated LOCA tests will also pass the simultaneous LOCA tests.