Loss Of Coolant Accident Scenario Research Articles

Determination of critical material damage parameter for Indian clad tube burst behavior under severe accident scenario

During loss of coolant accident scenario in PHWRs, heat removal from clad tubes containing the uranium oxides fuel may get severely impaired. This results in increase in temperature of the fuel clad and subsequent increase in the clad tube internal pressure. This causes the clad tube to deform and balloon leading to the increase in diameter of the fuel clad. After a critical value of deformation is reached, the clad tube gets burst. The burst process is due to initiation of crack in the clad due to creep deformation process in the presence of internal pressure and high temperature environment. In this work, experimental data of fuel clad bursts from literature have been used to evaluate a critical damage parameter that can be used to predict the onset of burst phenomenon of clad tubes. Finite element modelling of the fuel clad has been carried out to simulate the ballooning deformation of the fuel clad tubes due to creep. Norton’s creep model has been used in the simulation. The critical damage parameter evaluated, using analysis and experimental data, has been used subsequently to predict burst time of Zircaloy-4 fuel clad when subjected to different types of loading environments. An exponential decay type correlation has been obtained from the finite element analysis results which can be used by safety analysts to predict burst pressure of Zircaloy-4 fuel clad of Indian PHWR as a function of burst temperature or vice-versa.

Stress rupture of SiC/SiC composite tubes under high‐temperature steam

AbstractSiC‐fiber–reinforced SiC matrix composite cladding for light water reactor fuel elements must withstand high‐temperature steam oxidation in a loss‐of‐coolant accident scenario (LOCA). Current composite designs include an outer monolithic SiC layer, in part, to increase steam oxidation resistance. However, it is not clear how such a structure would behave under high‐temperature steam in the case when the monolithic layer cracks and carbon interphases and SiC fibers are exposed to the environment. To fill this knowledge gap, stress‐rupture tests of prototypic SiC composite cladding at 1000°C under steam and inert environments were conducted. The applied stress was ∼120 MPa, which was beyond the initial cracking stress. The failure lifetime under steam was 400–1300 s, while 75% of the composite specimens did not fail after 3 h of total exposure under inert gases. Microstructural observations suggest that steam oxidation activated slow crack growth in the fibers, which led to failure of the composite. The results from this study suggest that stress rupture in steam environments could be a limiting factor of the cladding under reactor LOCA conditions.

Numerical analysis on in-core ignition and subsequent flame propagation to containment in OPR1000 under loss of coolant accident

Since Fukushima nuclear power plant (NPP) accident in 2011, the importance of research on various severe accident phenomena has been emphasized. Particularly, detailed analysis of combustion risk is necessary following the containment damage caused by combustion in the Fukushima accident. Many studies have been conducted to evaluate the risk of local hydrogen concentration increases and flame propagation using computational code. In particular, the potential for combustion by local hydrogen concentration in specific areas within the containment has been emphasized. In this study, the process of flame propagation generated inside a reactor core to containment during a loss of coolant accident (LOCA) was analyzed using MELCOR 2.1 code. Later in the LOCA scenario, it was expected that hydrogen combustion occurred inside the reactor core owing to oxygen inflow through the cold leg break area. The main driving force of the oxygen intrusion is the elevated containment pressure due to the molten corium–concrete interaction. The thermal and mechanical loads caused by the flame threaten the integrity of the containment. Additionally, the containment spray system effectiveness in this situation was evaluated because changes in pressure gradient and concentrations of flammable gases greatly affect the overall behavior of ignition and subsequent containment integrity.

Hydrogen Isotopes Permeation in Clean or Unoxidized FeCrAl Alloys: A Review

The US Department of Energy is working with fuel vendors to develop accident tolerant fuels (ATF) for the current fleet of light water reactors (LWRs). The ATF should be more resilient to loss of coolant accident scenarios and help extending the life of the operating LWRs. One of the proposed ATF concepts is to use iron-chromium-aluminum (FeCrAl) alloys for the cladding of the fuel. A concern in using ferritic FeCrAl is that this type of cladding may result in an increase in the concentration of tritium in the coolant. The objective of the current critical review is to collect and assess information from the literature regarding diffusion or permeation of hydrogen (H) and its isotopes deuterium (D) and Tritium (T) across industrial alloys (including FeCrAl) used or intended for the nuclear industry. Over a hundred years of data reviewed shows that the solubility of hydrogen in ferritic alloys is lower than in austenitic alloys but hydrogen permeates faster through a ferritic material than through austenitic materials. The tritium permeation rates in FeCrAl alloys are between those in austenitic stainless steels and in ferritic FeCr steels. The activation energy for hydrogen permeation is approximately 30 pct higher in the austenitic alloys compared with the ferritic (typically ∼ 50 kJ/mol in ferritic vs. ∼ 65 kJ/mol in the austenitic). None of the major elements in FeCrAl alloys react with hydrogen to form detrimental hydride phases. The effect of surface oxides on FeCrAl delaying hydrogen entrance into FeCrAl alloy is not part of this review.

Sensitivity study of thermal-hydraulic nodalization for MELCOR simulations of severe accidents in a pressurized water reactor

The CVH package in the MELCOR code is responsible for modelling the thermal–hydraulic behavior during severe accident. This work presents a sensitivity study of thermal–hydraulic nodalization in the core region for MELCOR simulations of postulated severe accidents in a pressurized water reactor (PWR). Three nodal schemes are developed with a good agreement of steady-state parameters. Two accident scenarios: loss of coolant accident (LOCA) and station blackout (SBO) are simulated. The analysis is focused on the effect of the control volumes (CVs) in the CVH nodalization on the simulation results of in-vessel accident progression, including the core degradation, hydrogen generation and fission products release, etc. It is found that compared with the coarse CVH nodalization (1 CV), the radial refinement of CVH nodalization (7 CVs) leads to different impacts on accident progression in the two scenarios: faster core relocation and more hydrogen generation is predicted in the LOCA scenario, but it is opposite in the SBO scenario. The finest nodal scheme (49 CVs) with refinement in both radial and axial direction tends to predict an earlier occurrence of cladding rupture, RPV failure and faster core relocation, as well as more hydrogen generation. The CVH refinement makes little difference on the radioactive release.

Risk analysis virtual ENvironment for dynamic event tree-based analyses

Conventional Event-Tree (ET) based methodologies are extensively used as tools to perform reliability and safety assessment of complex and critical engineering systems. One of the disadvantages of these methods is that timing/sequencing of events and system dynamics is not explicitly accounted for in the analysis. In order to overcome these limitations several techniques, also known as Dynamic Probabilistic Risk Assessment (DPRA), have been developed. Monte-Carlo (MC) and Dynamic Event Tree (DET) are two of the most widely used DPRA methodologies to perform safety assessment of Nuclear Power Plants (NPP). Since 2012, the Idaho National Laboratory (INL) is developing its own tool to perform Dynamic PRA: RAVEN (Risk Analysis and Virtual ENvironment). RAVEN has been designed in a high modular and pluggable way to enable easy integration of different programming languages (i.e., Python, C++) and coupling with other application including, among the others, several thermal–hydraulic and severe accident codes (e.g., RELAP5-3D, MELCOR, MAAP5, TRACE, etc.). RAVEN is aimed to provide a framework/container of capabilities for engineers and scientists to analyze the response of systems, physics and multi-physics, employing advanced numerical techniques and algorithms. Moreover, RAVEN models stochastic events, such as components failures, and performs uncertainty quantification (UQ). Such stochastic modeling is employed by using sampling strategies among which both MC and DET algorithms, which are going to be employed in this paper. In addition, RAVEN processes the large amount of data generated by sampling the physical models using data-mining based algorithms and risk assessment techniques. This paper provides an overview of the DET methodologies that have been deployed within the RAVEN framework, showing the potential of such techniques for the analysis of complex systems. A brief background of classical methodologies and their limitation is also reported and represent the motivation for the deployment of such dynamic technique. In addition, results from a pressurized water reactor loss of coolant accident scenario, using RELAP5-3D as physical model, are reported.

Modification of fuel performance code to evaluate iron-based alloy behavior under LOCA scenario

Accident tolerant fuels (ATF) has been studied since the Fukushima Daiichi accident in the research efforts to develop new materials which under accident scenarios could maintain the fuel rod integrity for a longer period compared to the cladding and fuel system usually utilized in Pressurized Water Reactors (PWR). The efforts have been focused on new materials applied as cladding, then iron-base alloys appear as a possible candidate. The aim of this paper is to implement modifications in a fuel performance code to evaluate the behavior of iron-based alloys under Loss-of-Coolant Accident (LOCA) scenario. For this, initially the properties related to the thermal and mechanical behavior of iron-based alloys were obtained from the literature, appropriately adapted and introduced in the fuel performance code subroutines. The adopted approach was step by step modifications, where different versions of the code were created. The assessment of the implemented modification was carried out simulating an experiment available in the open literature (IFA-650.5) related to zirconium-based alloy fuel rods submitted to LOCA conditions. The obtained results for the iron-based alloy were compared to those obtained using the regular version of the fuel performance code for zircaloy-4. The obtained results have shown that the most important properties to be changed are those from the subroutines related to the mechanical properties of the cladding. The results obtained have shown that the burst is observed at a longer time for fuel rods with iron-based alloy, indicating the potentiality of this material to be used as cladding with ATF purposes.

Large scale experimental facility for performance assessment of the vacuum vessel pressure suppression system of ITER

The nuclear fusion reactor ITER (International Thermonuclear Experimental Reactor) foresees a Pressure Suppression System (PSS) in order to manage a Loss of Coolant Accident (LOCA) in the Vacuum Vessel (VV). A LOCA scenario is postulated to occur in the cooling systems of plasma facing components. The Vacuum Vessel Pressure Suppression System (VVPSS) has a key safety function because a large internal pressure in the VV could lead to a breach of the primary confinement barrier.Steam condensation would occur in ITER at sub-atmospheric pressure, differently than the applications in nuclear fission reactors. Steam direct contact condensation (DCC) in sub-atmospheric conditions have never been implemented before, therefore an experimental assessment of the VVPSS is necessary to simulate this phenomenon in ITER's thermal-hydraulic conditions.Under the financial support of ITER Organization, an experimental program was performed at the Department of Civil and Industrial Engineering (DICI) of the University of Pisa in the past 3 years, using a small-scale experimental rig (1/22 scale factor).In order to assess the scale laws elaborated by means of the experimental results obtained in this small scale facility, a Large Scale facility was built at University of Pisa. This facility is almost a full scale mock-up of ITER's VVPSS from the geometric point of view (1/1.09 scale factor) and it is 1/10 scale factor from the steam mass flow rate point of view.This paper illustrates the results of the first test campaign. A 1/10 scale sparger system SPS-A (having a diameter equal to DN125 and 100 holes) was used with a maximum steam mass flow rate of 0.5 kg/s. This experimental configuration represents real operating conditions on the basis of the results of elaborated similitude analysis. The preliminary analysis of the experimental results confirms the agreement in terms of condensation regimes between the two reduced scale (1/22 and 1/10 scale) as foreseen by the elaborated scale laws.

Studies on the criterion for choking process in two-phase flow

The critical flow (also named choked flow) is a phenomenon occurring at the break during a typical loss-of-coolant accident (LOCA) in nuclear power plant (NPP); it has high impact on the LOCA scenario and its magnitude has to be reproduced accurately. This work is focused on the choking process and especially on the choking criteria, as mathematical ways to establish the choking point. The process of critical flow is analyzed quantitatively. The convergence analysis of the critical flow calculation ─ stiff ordinary differential equations (ODEs) ─ was also focused on. Two kinds of choking criteria (determinant and pressure gradient criteria) are discussed and their strict derivations for single-phase flow are recalled in this paper. The criteria may be also extended for two-phase flow. In order to confirm that, the two criteria were numerically investigated for long pipe, short pipe and orifice discharge tests. The results obtained by using the two different criteria are consistent as long as the chosen pressure gradient value remains large enough. Anyhow, according to our results, this value is larger for the case of orifice and short pipe discharges (compared with the long pipe discharge). In addition, for the determinant criterion (regardless of single- or two-phase flow), a compatibility condition should be considered for the system of ODEs at the critical point. An introspection of the compatibility condition is offered in this paper from algebraic considerations.

Development and demonstration of a methodology to evaluate high burnup fuel susceptibility to pulverization under a loss of coolant transient

For economic reasons, the US nuclear industry is renewing efforts to build a technical basis to extend rod average burnup limits above the current regulatory burnup limit of 62 GWd/MTU. The primary driver is to increase pressurized water reactor cycle lengths to 24 months, reducing the number of fresh fuel assemblies and core design constraints, thereby making core energy utilization more efficient. However, fuel pellet fragmentation and pulverization, termed high burnup fuel fragmentation (HBFF), has been observed in the high burnup (>90 GWd/MTU) Halden loss-of-coolant-accident (LOCA) integral test series. The issue gained attention when fuel fragmentation and pulverization were also observed closer to the current US regulatory limit during the US Nuclear Regulatory Commission (NRC) sponsored out-of-core integral test at Studsvik Nuclear in early 2011. This led to NRC concerns with potential changes to fuel and core designs relative to fuel pellet pulverization. In a letter to the NRC Commissioners, the staff specifically identified a need to “…define the boundary of safe operation for key fuel design and operating parameters,” stating that “the staff is challenged to evaluate the acceptability of future fuel design advancements and fuel utilization changes.” As such, it can be concluded that HBFF and potential dispersal into the reactor coolant system introduces additional complications in light-water reactor (LWR) fuel safety evaluations. However, it is not clear how much fuel will be susceptible to HBFF; nor has there been a methodology developed to evaluate fuel susceptibility to HBFF. To that end, this paper proposes an analysis methodology to assess fuel susceptibility to HBFF during LOCA scenarios. The work presented here uses the BISON fuel performance code to evaluate a representative pressurized water reactor fuel rod exposed to a rod average burnup of 75 GWd/MTU. Sensitivity studies investigated the impact of the peak cladding temperature, transient fission gas released, and pre-transient fission gas release on cladding ballooning and burst timing. Subsequently, a methodology to assess fuel susceptibility to HBFF will be developed based on experimental data published in the open literature. The methodology will then be demonstrated by calculating the mass of fuel susceptibility to HBFF. The BISON results conclude that increasing peak cladding temperature drastically decreased time to failure, and decreased balloon size both of which have been confirmed experimentally. Additionally, the effect of pre-transient and transient fission gas release affected cladding balloon size and burst timing. Lastly, fuel susceptibility to HBFF significantly decreased as a function of peak cladding temperature.

POSCA: A computer code for fission product plateout and circulating coolant activities within the primary circuit of a high temperature gas-cooled reactor

Numerical prediction of fission product plateout and circulating coolant activities under normal operating conditions is crucial in the design of a high temperature gas-cooled reactor (HTGR). The results are used for the maintenance and repair of the components as well as the safety analysis regarding early source terms under loss of coolant accident scenarios. In this work, a new computer code named POSCA (Plate-Out Surface and Circulating Activities) was developed based on a one-dimensional model to evaluate fission product plateout and circulating coolant activities within the primary circuit of a HTGR. The verification and validation of study for the POSCA code was done using available analytical results and two in-pile experiments (i.e., OGL-1 and VAMPYR-1). The results of the POSCA calculations show that POSCA is able to simulate plateout and circulating coolant activities in a HTGR with fast computation and reasonable accuracy.

The Limitations of an Air-Oxidation Breakaway Model to Predict a Zirconium Fire in a Spent Nuclear Fuel Pool Accident

The OECD/NEA Spent Fuel Pool (SFP) project was conducted to investigate consequences of spent nuclear fuel pool accident scenarios. From the project, it was observed that cladding temperature could abruptly increase at a certain point and the cladding was completely oxidized. This phenomenon was called a “zirconium fire”. This zirconium fire is one of the crucial concerns for spent fuel pool safety under a postulated loss of coolant accident scenario, since it would lead to an uncontrolled mass release of fission products into the environment. To capture this critical phenomenon, an air-oxidation breakaway model has been implemented in the MELCOR code. This study examines this air-oxidation breakaway model by comparing the SFP project test data with a series of MELCOR code sensitivity calculation results. The air-oxidation model parameters are slightly altered to investigate their sensitivities on the occurrence of the zirconium fire. Through such sensitivity analysis, limitations of the air-oxidation breakaway model are identified, and needs for model improvement is recommended.

Modelling of a Nordic BWR containment and suppression pool behavior during a LOCA with GOTHIC 8.1

Boiling water reactors use the Pressure Suppression Pool (PSP) to relieve the containment pressure in case of an accident. During the event of a Loss of Coolant Accident (LOCA), drywell air and steam are injected into the PSP through blowdown pipes. This may lead to thermal stratification, which is a relevant safety issue as it leads to higher water surface temperatures than in mixed conditions and thus, to higher containment pressures. The Effective Heat (EHS) and Momentum (EMS) Source models were previously introduced to predict the effect of small-scale direct contact condensation phenomena on the large-scale pool water circulation. In this paper, the EHS/EMS models are extended by adding the effect of non-condensable gases on the chugging regime. The EHS/EMS models are implemented in the GOTHIC code to model a full-scale Nordic BWR containment under different LOCA scenarios. The results show that thermal stratification can be developed in the PSP.

Loss of coolant accident analysis under restriction of reverse flow

Abstract This paper analyzes a new method for reducing boiling water reactor fuel temperature during a Loss of Coolant Accident (LOCA). The method uses a device called Reverse Flow Restriction Device (RFRD) at the inlet of fuel bundles in the core to prevent coolant loss from the bundle inlet due to the reverse flow after a large break in the recirculation loop. The device allows for flow in the forward direction which occurs during normal operation, while after the break, the RFRD device changes its status to prevent reverse flow. In this paper, a detailed simulation of LOCA has been carried out using the U.S. NRC's TRACE code to investigate the effect of RFRD on the flow rate as well as peak clad temperature of BWR fuel bundles during three different LOCA scenarios: small break LOCA (25% LOCA), large break LOCA (100% LOCA), and double-ended guillotine break (200% LOCA). The results demonstrated that the device could substantially block flow reversal in fuel bundles during LOCA, allowing for coolant to remain in the core during the coolant blowdown phase. The device can retain additional cooling water after activating the emergency systems, which maintains the peak clad temperature at lower levels. Moreover, the RFRD achieved the reflood phase (when the saturation temperature of the clad is restored) earlier than without the RFRD.

Fretting wear comparison of cladding materials for reactor fuel cladding application

Relative motion between the fuel rods and fuel assembly spacer grids can lead to excessive fuel rod wear and, in some cases, to fuel rod failure. Based on industry data, such grid-to-rod-fretting is a significant cause of fuel failures in U.S. pressurized water reactor power plants. Kanthal advanced powder metallurgy technology or APMT, an FeCrAl steel alloy, and a braided SiC fiber, Chemical Vapor Infiltration SiC matrix (SiC/SiC) cladding by General Atomics are possible alternatives to conventional fuel cladding in a nuclear reactor due to their favorable performance under accident conditions. Tests were performed to examine the reliability of the cladding candidates and a conventional cladding, Zircaloy-4, under dry fretting conditions at elevated temperature. The contact was simulated with a rectangular and a cylindrical specimen over a line contact area. Confocal scanning laser microscopy was used to obtain a 3D map of the surface, which was in turn used for wear and work rate calculations on the samples. The wear rate coefficient was used as a measure of the performance and wear under fretting. Additionally, Energy Dispersive Spectroscopy was performed to qualitatively describe the microchemical changes the material undergoes during fretting. While APMT steel and SiC/SiC can perform favorably in loss of coolant accident scenarios, they also need to perform well when compared to Zircaloy-4 with respect to fretting wear. Wear coefficient measurements showed that APMT steel performs favorably in comparison to Zircaloy-4 with respect to fretting wear.

Efficacy analysis of hydrogen mitigation measures of CANDU containment under LOCA scenario

Mitigation of hydrogen risk has been a critical objective to assure ultimate safety of the nuclear power plant (NPP). Since the CANDU containment showed potential vulnerability over hydrogen explosion during severe accident, efficacy of hydrogen mitigation measures equipped in the current CANDU NPP need to be investigated thoroughly. Herein we report the effect of mitigation measures such as spray, igniter, passive autocatalytic recombiner (PAR), local air cooler (LAC), and filtered containment venting system (FCVS) on the hydrogen behavior using MELCOR 2.1 code. Wolsong Unit 1 was selected for target NPP and Loss of Coolant Accident (LOCA) was assumed as a plausible severe accident scenario. The analysis showed that the hydrogen concentration in the lower region of the containment was higher in the unmitigated base case. The application of FCVS with other safety measures successfully prevented the containment failure. Expected safety function of the FCVS was to filter out the fission products but substantial amount of oxygen was also vented out. This leads to the oxygen starvation in the containment and prevention of the containment failure in the long run.

Study of in-containment source term behavior for VVER-1000 under LOCA conditions using the IRBURN code system

The purpose of this paper is to present a kinetic model for simulation and calculation of in-containment activity in LWR power reactors after severe accidents. In-containment accident source term depends on the reactor containment specification as well as its safety performance under different accident conditions. A module is developed for quantification of the in-containment activity and loss of coolant accident scenario in a VVER-1000 reactor was modeled asan illustrative example in this study. This module is implemented in the new version of the IRBURN code system. The effects of engineering safety features on the fission products activities during accident time have been investigated.According to the results, the impact of re-suspension rate from free surfaces on the source term production is dominant during initiation of the accident, while the influence of volumetric time dependent source increases during the evolution of the accident. Besides, the deposition on the containment surfaces has almost the largest contribution in removal of activity during the accident.

A forward-backward coupled source term estimation for nuclear power plant accident: A case study of loss of coolant accident scenario

Source term information of radioactive release in a nuclear accident is important for emergency response. Two major categories of source term estimation techniques are forward method based on the status data of nuclear power plant and backward method based on environmental monitoring data. Both data contain considerable uncertainties which can propagate through estimation and result in a biased estimate. To solve the problem, a coupled estimation method is proposed in this study, which combines forward method (Response Technical Manual, RTM-96) and backward method (ensemble Kalman filter). The coupled method builds an evolution model based on a forward estimate and uses it to constrain the temporal correlation of estimates in the backward part, so that the propagation of uncertainties is reduced. Numerical experiments and sensitivity analysis are performed to verify the proposed method, based on a hypocritical loss-of-coolant (LOCA) accident process and the records of a tracer field experiment for Sanmen nuclear power plant. The results demonstrate that the coupled method provides the most accurate estimate in all tests and is more robust to uncertainties in various parameters than both RTM-96 and ensemble Kalman filter.

The modelling of condensation in horizontal tubes and the comparison with experimental data

The condensation in horizontal tubes plays an important role in determining the operation mode of passive safety systems of modern nuclear power plants. In this paper, two different approaches for modelling of this phenomenon are compared and verified against experimental data. The first approach is based on the flow regime map developed by Tandon. Depending on the regime, the heat transfer coefficient is calculated according to corresponding semi-empirical correlation. The second approach uses a general, fully empirical correlation proposed by Shah. Both models are developed with utilization of the object-oriented, equation-based Modelica language and the open-source Open-Modelica environment. The results are compared with data obtained during a large scale integral test, simulating a Loss of Coolant Accident scenario performed at the dedicated Integral Test Facility Karlstein (INKA) which was built at the Components Testing Department of AREVA in Karlstein, Germany. The INKA facility was designed to test the performance of the passive safety systems of KERENA, the new AREVA boiling water reactor design. INKA represents the KERENA containment with a volume scaling of 1:24. Components heights and levels over the ground are in the full scale. The comparison of simulations results shows a good agreement.

An exploratory work on the use of a Lagrangian-Eulerian model for simulating heat transfer of subchannels under reflooding conditions

In case of a Loss-of-Coolant Accident (LOCA) in a Light Water Reactor (LWR), the reactor core becomes partially uncovered and the quick degradation of heat removal—from the nuclear fuel elements—constitutes a critical factor in the integrity of the reactor core. As time elapses from the beginning of the accident, the temperature of the fuel clad continues to rise until the increased cooling produced by the injection of water, reduces the clad temperature and propitiates the rewet of the rods. If the surface temperature of the fuel elements exceeds the activation temperature of the Zr-O2 reaction, the core is damaged (partial or full meltdown). Hence, the need of having analytical tools to accurately predict the cladding and steam temperature under a LOCA scenario, is paramount to design palliative measures to prevent a possible core damage. The multi-scale nature of the physics behind reflooding makes the modelling of this type of problem a challenging task. In a first attempt to study a reflooding problem, within a Lagrangian-Eulerian framework, we focus our attention on the cooling effect of small droplets entrained in the superheated steam. We assess the effect of drag, lift and thrust forces—acting on individual droplets—on the steam temperature and liquid fraction distribution. Different correlations for the lift force are evaluated and uncertainties associated to the drag and lift coefficients of evaporating droplets are discussed.