Loss Of Coolant Accident Research Articles

A Risk Approach to the Management for the Pre-defueled Phase in the Decommissioning Transition Stage

We discuss the specific risk significance in the extended pre-defueled (PD) phase of the decommissioning process, particularly if spent fuels are still in the core due to the low-power and shutdown refueling plant operating state (POS). The issue of full-core discharge capability after permanent shutdown during the PD phase motivated this study on the evolution of system risks using a reference plant design of the two-unit/BWR-4/Mark-I. The effects of the reactor core and the spent fuel pool (SFP) on the incorporative risks are explored. The probabilistic risk assessment methodology, including the technical elements, is systematically developed by defining two primary configurations from the internal event analysis under the models 30, 60, 180, 365, and 942 days after permanent shutdown, respectively. The movable refueling gate between the reactor core and the SFP, as well as the residual heat removal (RHR) system, have been subjected to two sensitivity studies on system configurations in order to examine the induced impacts by the refueling gate and cooling systems. MELCOR, a realistic thermal-hydraulic code, is utilized to determine the decay heat levels and the success criteria after shutdown. The two operator tasks are assumed to be independent in the situation of decreasing decay heat after shutdown and a long time available for human actions. In addition, the WinNUPRA software package is used for the fuel uncovery sequence quantification. Plant-centered loss-of-offsite power (LOOP), flow diversion loss-of-coolant accidents (LOCAs) to the suppression pool via the RHR system, switchyard-centered LOOPs, and LOCAs in the connected systems via the RHR, have proven to be the most significant initiating events for the configurations. When compared to the low-power and shutdown refueling POS, the realistic quantification results in terms of fuel uncovery frequencies and the evolution of the risk profile for the basic and sensitivity configurations meet the expectations under the PD-phase condition of low-decay heat levels.

Effectiveness of Ni-based and Fe-based cladding alloys in delaying hydrogen generation for small modular reactors with increased accident tolerance

This study investigates the high temperature oxidation behaviour of a Ni–20Cr-1.2Si (wt.%) alloy in steam from 1200 °C to 1350 °C by Thermogravimetric Analysis (TGA), Scanning Electron Microscopy (SEM), Energy Dispersive X-ray Spectroscopy (EDS) and X-ray Diffraction (XRD). The results demonstrate that exposed Ni-based alloy developed a thin oxide scale, consisted mainly of Cr2O3. The oxidation kinetics obtained from the experimental results was applied to evaluate the hydrogen generation considering a simplified reactor core model with different cladding alloys following an unmitigated Loss-Of-Coolant Accident (LOCA) scenario in a hypothetical Small Modular Reactor (SMR). Overall, experimental data and simulations results show that both Fe-based and Ni-based alloys may enhance cladding survivability, delaying its melting, as well as reducing hydrogen generation under accident conditions compared to Zr-based alloys. However, a substantial neutron absorption occurs when Ni-based alloys are used as cladding for current uranium-dioxide fuel systems, even when compared to Fe-based alloys.

Effects of azimuthal temperature distribution and rod internal gas energy on ballooning deformation and rupture opening formation of a 17 × 17 type PWR fuel cladding tube under LOCA-simulated burst conditions

ABSTRACT In order to contribute to better modeling and evaluation of fuel fragmentation, relocation, and dispersal expected under loss of coolant accident (LOCA) conditions, LOCA-simulated cladding burst experiments were performed on as-received nonirradiated 17 × 17 type Zircaloy-4 cladding specimens that were internally pressurized. The experiments were designed to terminate at burst occurrence to focus on ballooning and rupture opening formation and to investigate the effects of various factors. The postburst cladding hoop strain decreased with the increase in azimuthal temperature difference (ATD) of the cladding, as found previously. The rupture opening size increased with the increase in ATD and the increase in energy of the pressurized gas stored inside the pressure boundary of the test sample system. Comparison with the existing database, which included tests on irradiated rods containing fuel pellets, suggested that formation of the rupture opening was influenced by the characteristic behavior of high burnup fuels, such as limited gas migration in the cladding tube due to fuel–cladding bonding and interaction of the ejected fuel fragments with the cladding tube.

Effects of fuel relocation on fuel performance and evaluation of safety margin to 10CFR50.46c ECCS acceptance criteria in APR1400 plant

Characteristics of fuel relocation and its effect on safety margin in a loss-of-coolant accident (LOCA) safety analysis have been analyzed. Packing fraction model developed by Quantum Technology was validated and slight modification was conducted based on the formation of high burnup structure model. FAMILY computer code that is an integrated code with FRAPTRAN and MARS-KS is used for the evaluation. The peak cladding temperature (PCT) and peak local oxidation (PLO) of the APR1400 nuclear power plant with the PLUS7 fuel during LOCA were evaluated by a best-estimate plus uncertainty analysis methodology up to 60 MWd/kgU. Fuel relocation resulted in an increase in 95 %/95 % probability/confidence level of the reflood PCT (95/95 reflood PCT) and the PLO (95/95 PLO) until the fuel burnup of ∼ 50 MWd/kgU. Maximum increase of 95/95 reflood PCT was about 130 K at 30 MWd/kgU burnup. This results in the minimum safety margin of the reflood PCT at that burnup with respect to the U.S. NRC proposed 10CFR50.46c PCT limit. Maximum increase of 95/95 PLO was 3.9 percent point at 30 MWd/kgU. However, minimum safety margin was observed at different burnups such as 40 MWd/kgU and 60 MWD/kgU. Regarding the cladding burst characteristics, fuel relocation may lead to an increase in the burst temperature and burst strain of the cladding, and it also induces an earlier failure of cladding.

Effect analysis of break size on source term release and radioactive consequences of marine nuclear reactor during loss of coolant accident

Effect analysis of break size on source term release and radioactive consequences of marine nuclear reactor during loss of coolant accident

Highly underexpanded helium jet and its effect on dust resuspension under in-vessel loss of coolant accident of fusion reactor

During an in-vessel loss of coolant accident (LOCA) of a helium-cooled blanket in a fusion reactor, the high-pressure helium gas ingress into the vacuum vessel (VV) will produce highly underexpanded jet, resulting in the resuspension and migration of the radioactive dust deposited in the VV. In this study, the flow field in the VV and the influence of friction velocity were simulated and analyzed by using ANSYS Fluent code. In addition, the differences between in-vessel LOCA and loss of vacuum accident (LOVA) were compared. The results showed that the pressure of the VV reaches the pressure relief limit of the VV pressure suppression system rapidly after a LOCA. With the development of a LOCA, the Mach disk continued to move towards the equatorial break of the VV, its shape continued to shrink and the number of structures gradually increased. Compared with the LOVA, the LOCA had a higher friction velocity, resulting in a greater probability of dust resuspension and radioactive risk.

Research on modeling in operator mental workload based on VACP method

AbstractIn the context of the applications of higher automation and complex system configuration in nuclear power plants (NPPs), ensuring nuclear safety by maintaining the reliability of the operators in main control room is become more crucial than ever before. As an essential dimension of physiological indicator, human mental workload is widely used in evaluating human performance and reliability. This paper presents an approach to evaluate the mental workload of operators using multiple resources in Visual, Auditory, Cognitive, and Psychomotor (VACP) model, and takes time pressure (TP) as a critical index. We have performed the approach by a real case study and analyzed the relationship between the operator mental workload and TP, task type and the error probability during the Loss of Coolant Accident (LOCA) under full power operation. There are two operating phases in the workload has been analyzed. Moreover, the relationship between human‐error probability (HEP) and TP had been studied via volatility gain analysis for distinguishing which step of a task is of most important. The results shows that the fault remind (“informing LOCA”) is the first important fluctuation in the phase ignoring TP and HEP. On the other hand, the information validity (“ensuring control rods has been inserted to the bottom”) in the case considering TP is the most important. Because the high task load is positively correlated with human error, the research provides an important basis for preventing human errors in the phase of NPP designs, which will be especially useful in safety‐critical tasks.

Towards coupling conventional with high-fidelity fuel behavior analysis tools

Multi-dimensional fuel performance codes have the potential to enhance the understanding of complex phenomena affecting fuel integrity, but their application is often limited by the high computational cost. Recognizing that many multi-dimensional aspects such as the ballooning of the cladding during a loss of coolant accident or the pellet-cladding mechanical interaction (PCMI) during power ramps are of interest mainly for operational and accidental transients, we have developed a coupling methodology between conventional and high-fidelity fuel behavior analysis. This consists in using a well-validated 1.5-D code to rapidly simulate the base irradiation and set the initial and boundary conditions for the 2-D or 3-D transient model, leading to an optimized use of computational resources. In this paper we present the coupling methodology, and we apply it to link the TRANSURANUS code and the multi-dimensional code OFFBEAT. To demonstrate the capabilities of the coupling tool, we study the PCMI during the end-of-life power ramp of the AN3 segment from the Risø-3 experiment and we compare the predicted cladding deformation and bamboo structure against the experimental profilometry provided by the Post Irradiation Examination (PIE). In the conclusions, we also outline further potential applications of the coupling methodology.

Simulation of flow redistribution in 7 × 7 ballooned fuel bundle experiments using DRACCAR code

IRSN has developed DRACCAR code in order to model loss of coolant accidents (LOCA) in a light water nuclear reactor. The physics during a hypothetical loss of coolant accident involves several phenomena such as heat and mass transfers, fluid dynamics, mechanics, and chemistry. Even if a simulation requires modeling these conjugated processes, it is important to validate them separately first—for example, fluid dynamics calculations of the flow redistribution after the clad ballooning that can occur during a LOCA. In this paper, we present DRACCAR simulations of the fluid dynamics in two 7 × 7 ballooned bundles with different blockage, two ballooned lengths and with two different flow rates, one laminar and the other turbulent. DRACCAR results for the axial velocities are compared to experimental data obtained by MRI techniques on the MASCARA facility. The present results show that, in most of the sub-channels studied, DRACCAR is capable of predicting correctly the velocity profiles for both bundles and flow rates. In general, DRACCAR provided fairly accurate estimates of the axial velocity, especially for the intact and less-blocked sub-channels. For the most blocked subchannel, DRACCAR is able to estimate the majority of the axial velocity in a margin of ±15% with respect to the MASCARA values, except for the most blocked bundle at laminar flow where only 20% of the calculated velocities were found within this margin.

BEPU Analyses for the Postulated Loss-of-cooling Accident Scenario in the Spent Fuel Pool at the Fukushima Daiichi Unit 4

The aim of the paper is to present the results obtained from the study devoted on the Best Estimate plus Uncertainty (BEPU) Methods application. To this end, the simulation of the Loss of cooling accident (LOCA) scenario in the Fukushima Daiichi spent fuel pool was carried out. Special attention was paid on the uncertainty analysis with the domain fission products release. The work presented herein was carried out by using the ASTECv2.2b/SUNSET in the coupling mode. A brief description of the ASTECv2.2b/SUNSET coupling is provided and a short discussion concerning the BEPU methodology is also presented.

Development of a cladding burst model under the loss-of-coolant accident conditions in a spent fuel pool

ABSTRACT The behavior of spent fuel rods under loss-of-coolant accident (LOCA) conditions in a spent fuel pool (SFP) is not fully understood. Prediction of the cladding burst is important because it results in the release of fission gas into the environment. The objective of this study is to develop a simulation model to predict cladding bursts under SFP-LOCA conditions. In the developed model, assuming that the profiles of temperature and internal rod pressure are known, the cladding hoop stress is evaluated sequentially until the hoop stress exceeds the threshold value. To improve the prediction accuracy of the creep strain, different correlations of the reported creep strain rates for Zircaloy-4 (Zry-4) and Zircaloy-2 (Zry-2) were used, depending on the range of temperature and hoop stress. The hoop stress criterion for a burst was correlated as a function of temperature by interpolating the burst stress measured in an out-of-pile degradation facility, DEGREE. The newly developed model was applied to simulate experimental bursting tests during heating and isothermal holding, and the bursting timing was successfully predicted. Developed model was also used to simulate the cladding burst under DBA-LOCA conditions and its applicability was confirmed.

Revisiting weight-gain correlations for as-fabricated Zircaloy-4 cladding tubes during steam oxidation

During a loss of coolant accident, elevated temperatures and steam environment induce fuel cladding oxidation. During cladding high-temperature oxidation, the weight increase is considered as a material property when divided by the surface exposed to the environment, this is the so-called weight-gain. Many laboratories tested Zircaloy-4 specimens and used several approaches to determine weight-gain correlations. Two frequently used methods are weight measurements and metallographic measurements of scale thickness combined with diffusion simulations. Usually, assuming that weight-gain is controlled by diffusion processes, the correlations have a parabolic dependency to steam exposure duration. In the present study, a metallographic method relying on scale thickness measurements is proposed with few assumptions on the oxygen transport process in the oxide and α(O) layers. These two layers contain the main oxygen inventory. The local weight-gain is obtained by the integration of a simplified oxygen profile across the specimen thickness. The proposed method is compared to specimen weighing approach and leads to accurate predictions. The method was applied to five experimental campaigns performed in different laboratories combining more than 375 tests. A single non-parabolic weight-gain correlation covering a large temperature range (900–1500 °C) is finally provided.

Thermal‐hydraulic investigation of a proposed concept for a passive containment cooling system and evaluation of heat exchanger performance

Reduction in ambient pressure within the containment of a light water reactor in the event of a hypothetical Loss of Coolant Accident (LOCA) is crucial for containment integrity. This article introduces the distinct concept of a Passive Containment Cooling System (PCCS). In the proposed PCCS concept, the containment is divided into two compartments with appropriate volumes. An external heat exchanger is connected between the two compartments of the containment for heat transfer and pressure reduction in the case of LOCA. This concept has unique characteristics and provides many advantages over previous designs. A RELAP5 model was developed to perform thermal-hydraulic analysis of the proposed PCCS concept. The performance of the PCCS after its deployment on a small reactor (998.6 MWth) has been assessed under various conditions following LOCA. The results indicate that partitioning of the containment creates a differential pressure that acts as a driving force for flow through the PCCS heat exchanger. Consequently, a decrease in long-term containment pressure is observed. In the case of equal volume fractions in the two containment compartments, the PCCS reduces long-term pressure in the containment by 16.5%. Sensitivity analysis was conducted in which the flow area of the PCCS inlet line, the heat transfer area, and the size of the heat exchanger tank were varied. The PCCS heat transfer capacity increases as these parameters are increased. However, the former is less sensitive than the latter two. The performance of the PCCS heat exchanger under steady-state conditions was validated using RELAP5 simulations, empirical correlations, and analytical models. The evaluation of heat exchanger parameters by various methods (presented in Appendix 1 ) was found in close agreement.

Effectiveness of SFP Spray Cooling during Loss of Coolant Accidents

For a large Spent Fuel Pool (SFP) loss of coolant accidents, properly sized SFP spray can slowdown or possibly preclude fast heat-up of spent fuel. The MELCOR 2.1 model of NPP Krsko pool was developed and tested for cases of loss of cooling accidents. The simple spray system with spray nozzles distributed at specified location at the top of the pool was added to the model. Different loss of coolant rates where studied for different fuel heat loadings, and different openings and flow rates of the spray nozzles. Traditionally, spray nozzles able to produce larger diameter droplets are used close to the fuel locations with higher heat loadings. According to preliminary results, spray nozzles that will be installed are able to limit or delay long-term heat-up of the spent fuel, but in the case of late actuation it is possible to have temporary high oxidation rates and corresponding production of hydrogen.

Mitigation of jet cross-flow induced vibrations using an innovative biomimetic nozzle design inspired by shark gill geometry

Shark gill slits enable sharks to eject the water after the oxygen has been removed in ram ventilation. The reduced effect of jet flow from the gill slits gives sharks smooth maneuverability. A biomimetic “shark nozzle” is proposed to improve mixing between the jet flow and surrounding fluid. Jet flow systems are an essential component of many industrial applications. The key characteristic of jet flow is the mixing process that occurs between two fluid streams to allow heat and/or mass transfer between them. Many industrial and propulsion devices that use jet flows need rapid mixing for effective and environmentally friendly operation. Fuel-injection systems, chemical reactors, and heating and air conditioning systems are examples of devices where a mixing process takes place. Recently, jet flow plays an important role in design and operation of specialized nuclear pressurized water reactors (PWRs). The loss-of-coolant accident (LOCA) holes and slots machined in the core periphery baffle plates are designed to mitigate the effects of a severe LOCA event. However, in normal operation, the holes are a source of a jet flow that can induced vibrations in the fuel assemblies near the baffle before being mixed with the surrounding fluid. This may cause wear and fretting in fuel rods with their supports. The ultimate solution to prevent the fuel assembly vibrations from LOCA hole jetting in a reactor is enhancing mixing of a jet flow with ambient flow in order to rapidly reduce jet momentum. This work proposes a new shark-inspired nozzle design that exploits the observed high efficiency capacity of shark gill slits. Tests are conducted to evaluate the performance of the new design. The obtained results show that the shark-inspired biomimetic nozzle has a greater effect on the rod bundle vibration, and the critical velocity at which the unstable vibration occurs in the rod bundle is delayed by 20% using the biomimetic nozzle. In addition to delaying instability, a vibration amplitude reduction of 85% was obtained by using the proposed shark-inspired nozzle instead of the circular nozzle.

CFD analysis of water vapor condensation in large containments: Numerical model, verification and validation

In the event of a loss of coolant accident (LOCA) in a Pressurized Water Reactor (PWR), large quantities of water vapor and hydrogen might be released into the containment building. Increasing pressure and the risk of hydrogen deflagration can threaten the containment integrity. Water vapor condensation on cold containment structures modifies the pressure load on the containment structures. On the one hand, the total pressure reduces due to the decrease of the water vapor content of the containment atmosphere. On the other hand, the concentration of hydrogen increases, which can reach locally the threshold of deflagration or detonation. A Computational Fluid Dynamics (CFD) model is developed and tested in order to predict transients of the pressure and the concentrations of gaseous species in reactor containments after a LOCA. Since the free volume of the containment of typical French PWRs is between 50,000 m3 and 80,000 m3, the CFD model requires significant simplifications.The physical model and its simplifications as well the implementation in the CEA OpenSource CFD code TrioCFD are described. Model tests are presented in two steps. In a first step, verification tests are discussed to show that the condensation model is implemented correctly in the CFD code. In a second step, the model is validated against experimental data of the International Standard Problem ISP47. In this benchmark, steady state conditions between vapor injection and condensation were reached experimentally in the MISTRA test containment of the CEA. In phase A of the benchmark, an equilibrium was achieved between water vapor injection and water vapor condensation on temperature controlled cold walls. One incondensable gas, namely air, was present in the test vessel. In phase B, a second incondensable gas was added to the containment atmosphere, namely helium. Both steady state situations were analyzed with the CFD model. The calculations represent well the experiment when the predominant condensation paths are modelled. It is shown that the pressure in the containment vessel as well as the mean mass fractions of water vapor and air, as well as of helium in phase B, are calculated in accordance to the experiment. The temperature in the containment is overestimated. Measured vertical profiles of the species concentrations are reproduced correctly.

Application of Best Estimate Plus Uncertainty (BEPU) Methodology in a Final Safety Analysis Report (FSAR) of a Generic Plant

The licensing process of a nuclear power plant is motivated by the need to protect humans and the environment from ionizing radiation and, at the same time, sets out the basis for the design and determining the acceptability of the plant. An important part of the licensing process is the realization of accident analysis related to the design basis, which should be documented in the Final Safety Analysis Report (FSAR). There are different options on accidents calculation area by combining the use of computer codes and data entry for licensing purposes. One is the Best Estimate Plus Uncertainty (BEPU), which considers realistic input data and associated uncertainties. Applications of BEPU approaches in licensing procedures were initiated in the 2000s, first to analysis of Loss of Coolant Accident (LOCA), and then to the accident analysis as a whole, documented in Chapter 15 of the FSAR. This work has as main objective the implementation of BEPU methodology in all analyses contained in FSAR, through the homogenization of the analytical techniques and identification of key disciplines and key topics in the licensing process.

Accident Simulation Study for Nuclear Power Plants Impacting Louisiana: Differences in 2017 vs. 1992 Protective Action Guidelines.

Louisiana has aligned its radiological emergency program with the 2017 US Environmental Protection Agency Protective Action Guides Manual but has added a child thyroid dose evacuation threshold in lieu of distributing potassium iodide to the public. The nuclear power plants will continue to align with the 1992 manual for the foreseeable future, which could lead to possible accident scenarios in which state recommendations would differ from those of the utility. The objective of this study is to predict what accident and weather conditions will lead to a differing set of recommendations. This study performs a representative set of simulations of potential nuclear power plant accidents using a combination of the RASCAL software package, provided by the Nuclear Regulatory Commission, and a Software system used by Entergy combining an older RASCAL dose modeling methodology with plant-specific input. Four preliminary results of this study are presented: a spent fuel fire where differences in whole body dose lead to very different evacuations, a loss of coolant accident in which the child thyroid dose is the determining factor, a core melt accident using stack monitors to locate the evacuation threshold point, and a spiked coolant accident that could lead to an evacuation order before the plant declares a General Emergency. Weather plays as great a role as accident conditions in determining whether the evacuation recommendations differ. The completed results of this study can provide guidance to states as they evaluate the transition to the 2017 guidelines.

Numerical investigation of the depressurization performance in blowdown pipe using method of characteristics

The coolant leaks quickly after the break of the high-pressure pipeline system. The rapid pressure decrease and fluctuation will further affect the integrity of the pressure boundary and the effective heat removal of the reactor core in the early phase of blowdown. Generally, numerical simulation plays a vital role in the safety analysis of nuclear power plants (NPPs) under severe accidents like loss of coolant accident (LOCA) conditions. In this paper, one-dimensional two-phase hydrodynamic equations and essential closure models are established to describe the internal pipe flow firstly. Then, method of characteristics (MOC) realized by a self-programming code is introduced, which is different from the numerical scheme of system analysis code RELAP5 in the simplification and discretization. Finally, combined with the Edwards blowdown experiments, the two numerical solution methods are compared on modeling, sensitivity analysis, and calculation results under various working conditions. The influence of back pressure variation on the solution of thermal–hydraulic parameters has been considered during the blowdown process. The solution result of RELAP5 is highly related to the set discharge coefficients, which are usually given according to experience. However, the solution result of MOC can be enveloped by that of RELAP5 with different discharge coefficients. In addition, the results solved by MOC are closer and more conservative to the experimental values during the early phase of blowdown with low void fraction fluid. It lays a foundation for the further calculation and safety analysis of the complex loop system under LCOA conditions.

Research on Countercurrent Flow Limitation in Reactor Hot Leg at the Loss of Coolant Accident-Thermohydraulic Calculation With System Code RELAP5

Since the counter current flow limitation (CCFL) of the two-phase flow has an important influence on the safety analysis of the reactor, it is also of great significance to study some characteristics and flow pattern transitions of the two-phase flow. In order to explore the characteristics and influencing factors of two-phase counter current flow in the hot leg under the condition of the Loss of Coolant Accident, this research, using RELAP5 simulation calculation, analyzed the effect of the lengths of the flow path of the hot leg, the shapes of the upper outlet of the inclined pipe, and the shapes of the elbow connecting the horizontal pipe and the inclined pipe on the characteristics of the CCFL. This research also evaluated the limitation of the model. The simulation founded that under the same liquid flow rate, the longer the horizontal pipe, the lower the gas flow rate at the beginning of the counter current flow; the effect of the elbow on the CCFL was not great; compared with the sharp outlet shape, the Jf of the gentle circle was slightly larger. The method established in this study is suitable for the prediction of CCFL in the hot leg during Loss of Coolant Accident.