Loss-of-Fluid Test Research Articles

Effect of critical flow model in MARS-KS code on uncertainty quantification of large break Loss of coolant accident (LBLOCA)

Abstract The critical flow phenomenon has been studied because of its significant effect for design basis accidents in nuclear power plants. Transition points from thermal non-equilibrium to equilibrium are different according to the geometric effect on the critical flow. This study evaluates the uncertainty parameters of the critical flow model for analysis of DBA (Design Basis Accident) with the MARS-KS (Multi-dimensional Analysis for Reactor Safety-KINS Standard) code used as an independent regulatory assessment. The uncertainty of the critical flow model is represented by three parameters including the thermal non-equilibrium factor, discharge coefficient, and length to diameter (L/D) ratio, and their ranges are determined using large-scale Marviken test data. The uncertainty range of the thermal non-equilibrium factor is updated by the MCDA (Model Calibration through Data Assimilation) method. The updated uncertainty range is confirmed using an LBLOCA (Large Break Loss of Coolant Accident) experiment in the LOFT (Loss of Fluid Test) facility. The uncertainty ranges are also used to calculate an LBLOCA of the APR (Advanced Power Reactor) 1400 NPP (Nuclear Power Plants), focusing on the effect of the PCT (Peak Cladding Temperature). The results reveal that break flow is strongly dependent on the degree of the thermal non-equilibrium state in a ruptured pipe with a small L/D ratio. Moreover, this study provides the method to handle the thermal non-equilibrium factor, discharge coefficient, and length to diameter (L/D) ratio in the system code.

Node configuration uncertainty in nuclear safety analyses

In nuclear safety analyses, a 1-D thermal hydraulic system analysis code with the best estimate plus uncertainty (BEPU) approach is generally used. There are sources of uncertainty such as nodalization, physical models, and numerical scheme. Among them, the authors specifically focused on the node configuration uncertainty related to spatial discretization that is regarded as less dominant sources of uncertainty in the code previously. Furthermore, it was not easy to quantify the uncertainty from the node configuration. The authors quantified and analyzed the node configuration uncertainty to understand how much it affects the solution. In this study, the effect of spatial discretization are studied by discussing simulation results for various spatially discretized node systems. The discussion on the node configuration uncertainty will be presented by examples of SUbcooled BOiling (SUBO) test simulation and Loss Of Fluid Test (LOFT) simulation using Multi-dimensional Analysis of Reactor Safety (MARS) code developed in South Korea. From SUBO test simulation, it was found that when the number of nodes is small than the node configuration uncertainty can overwhelm the uncertainty from the physical model. From LOFT simulation, it was revealed that the node configuration can cause bifurcation in the complex nuclear system analyses. Thus, it is concluded that the node configuration uncertainty can be quite significant and has to be considered in the nuclear system analyses.

Large-break LOCA analysis with modified boiling heat-transfer model in TRACE code

Numerical analyses were conducted to replicate several tests for simulating a double-ended cold-leg large break loss-of-coolant accident (LBLOCA) in the Loss-of-Fluid Test (LOFT) using the TRACE (version 5/patch level 4) code. Analytical results by the original TRACE code were so conservative that especially a first peak of cladding temperature was estimated higher than the experimental data at the blowdown phase and subsequent temperature drop corresponding to the temporal quench was not seen. We were interested in minimum film boiling temperature (Tmin) as a heat transfer model factor estimating the quench at the moment, investigated correlation equations for Tmin in previous studies and especially focused on ones given as a function of coolant mass flow because the complicated flow transient and decompression in the core region at the blowdown phase was interpreted as having an influence on the cladding temperature behavior. There are several correlations meeting the above condition but it was revealed that they are insufficient to apply for high pressure especially. Therefore, a new term including an effect of mass flow flux and time derivative of pressure was defined and added with a proportional coefficient hypothetically to the current correlation in the TRACE code for modification. The LOFT analyses were conducted again using the modified TRACE code, and it was shown by applying roughly the same proportional coefficient to all the cases of LOFT analyses that estimation of the cladding temperature behavior was improved more precisely at the blowdown phase. Also, the transition during the phase was explained phenomenologically with the wall heat transfer mode and boiling curve.

Improvement and validation of the wall heat transfer package of RELAP5/MOD3.3

The process of energy transfer from heat structure to control volume is determined by the wall-to-fluid heat transfer package, which is crucial for nuclear reactor safety analysis codes. The current logic for selection of heat transfer modes of RELAP5/MOD3.3 code is too complex and may result in incorrect heat transfer mode judgment. Also, the narrow application scope of film boiling heat transfer correlations may result in large errors in film boiling region which is of paramount importance for the predicted peak clad temperatures during hypothetical LB-LOCAs in PWRs. In this study, a new heat transfer package has been developed and incorporated into the RELAP5/MOD3.3 code. Differing from the original package, the modified one consists of twelve heat transfer modes and proposes a new logic for selection of heat transfer modes. For each mode, the models in the existing safety analysis codes and the leading models in literature have been reviewed in order to determine the best model which can easily be applicable to the RELAP5/MOD3.3 code. Particularly (1) a new package of heat transfer correlations are produced; (2) a new logic for selection of film boiling and transition boiling heat transfer modes is proposed which use minimum film boiling temperature and critical heat flux temperature as distinguished points. The modified code has been validated by comparing the analysis results with available experimental data from tube post dryout experiments and loss-of-fluid test (LOFT) facility. The calculation results showed that the improved package could better predict the experimental phenomena with higher prediction accuracy.

Model Uncertainty in Severe Accident Calculations: A Structural Methodology with Application to LOFT LP-FP-2 Experiment

The influence of model uncertainty is most pronounced in areas of limited knowledge and large uncertainties like severe accident (SA) calculations. Lack of a systematic methodology for this purpose makes this assessment difficult. This paper describes the treatment of model uncertainty in SA analysis for nuclear power plants, which is an area that has had limited past research. This paper aims at a systematic subject assessment. By review of available approaches, a methodology is structured to deal with alternative modeling options in SA code structure. The proposed methodology comprises three phases: the probability of each model is estimated (phase 1), the input uncertainty is quantified (phase 2), and the Bayesian model averaging technique is utilized to integrate the calculations of alternative models into the SA code (phase 3). Through this process, the degree of belief is quantified for the performance of alternative code models. The methodology evaluates available information and data from experiments and code predictions. The application of the proposed methodology is demonstrated on fission product release models for the LP-FP-2 SA experiment of the LOFT (Loss-of-Fluid Test) facility.

A Bayesian ensemble of sensitivity measures for severe accident modeling

In this work, a sensitivity analysis framework is presented to identify the relevant input variables of a severe accident code, based on an incremental Bayesian ensemble updating method. The proposed methodology entails: (i) the propagation of the uncertainty in the input variables through the severe accident code; (ii) the collection of bootstrap replicates of the input and output of limited number of simulations for building a set of finite mixture models (FMMs) for approximating the probability density function (pdf) of the severe accident code output of the replicates; (iii) for each FMM, the calculation of an ensemble of sensitivity measures (i.e., input saliency, Hellinger distance and Kullback–Leibler divergence) and the updating when a new piece of evidence arrives, by a Bayesian scheme, based on the Bradley–Terry model for ranking the most relevant input model variables. An application is given with respect to a limited number of simulations of a MELCOR severe accident model describing the fission products release in the LP-FP-2 experiment of the loss of fluid test (LOFT) facility, which is a scaled-down facility of a pressurized water reactor (PWR).

Kriging as an alternative for a more precise analysis of output parameters in nuclear safety—Large break LOCA calculation

AbstractThe methods using the best estimate codes are now applied in safety demonstration for nuclear power plants (NPP) to evaluate uncertainties of the relevant output parameters. Towards this objective, it is useful to further analyse the outputs, for example, to learn more about sensitivity to input parameters. In addition, this first analysis can be used to assess uncertainty. Such an analysis is difficult to obtain using the code itself because it is quite time‐consuming. One approach, called response surface methodology, consists in replacing the code by a simpler model, estimated with few runs. Linear regression is often used. In this paper, we propose kriging as introduced by Sacks et al. (Technometrics 1989; 31:41–47; Stat. Sci. 1989; 4(4):409–435) as an alternative.Kriging was applied to the Loss‐of‐Fluid Test (LOFT) loss of coolant experiment L2‐5, which was the subject of the former ISP 13 and the ongoing BEMUSE (ISP: International Standard Problem. BEMUSE: Best‐Estimate Methods–Uncertainty and Sensitivity Evaluation) international problem. The output is the second maximum peak cladding temperature (PCT) of the fuel. The best estimate code used is CATHARE2 V1.3L. We observe that kriging is more flexible and can handle irregularities. As a result, it gives more accurate predictions. In addition, sensitivity analysis is provided. This method offers complementary information and constructs a response surface more accurately, with a more realistic evaluation of risk. Copyright © 2009 John Wiley & Sons, Ltd.

Integrated Methodology for Thermal-Hydraulic Code Uncertainty Analysis with Application

This paper discusses an integrated thermal-hydraulic (TH) uncertainty analysis methodology with an application to the Loss-of-Fluid Test (LOFT) test facility large-break loss-of-coolant accident (LBLOCA) transient. The methodology is intended for applications to best-estimate analyses of complex TH codes. The goal is to develop an integrated method to make such codes capable of comprehensively supporting the uncertainty assessment with the ability to handle important accident transients. The proposed methodology considers the TH code structural uncertainties (generally known as model uncertainty) explicitly by treating internal submodel uncertainties and by propagating such model uncertainties in the code calculations, including uncertainties about input parameters. The methodology is probabilistic, using the Bayesian approach for incorporating available evidence in quantifying uncertainties in the TH code predictions. The types of information considered include experimental data, expert opinion, and limited field data, in treating both model and input parameter uncertainties. The code output is further updated through additional Bayesian updating with available experimental data from the integrated test facilities. The methodology uses an efficient Monte Carlo sampling technique for the propagation of uncertainty, in which a modified Wilks’ sampling criteria of tolerance limits is used to significantly reduce the number of simulations. This paper describes the key elements of the uncertainty analysis methodology and summarizes its application to the LOFT test facility LBLOCA.

Analysis of the Chemical Behavior of Iodine in the Suppression Tank of the LOFT Facility During Experiment LP-FP-2 with IODE and IMPAIR-2/M

The significance of iodine for source term quantification has been studied by investigating its chemical behavior under the prototypical conditions of a hypothetical severe accident within the containment. As a result, some computer codes were developed and their validation is currently under way. The loss-of-fluid test (LOFT) program was one of the most relevant research projects in the area of nuclear safety. Its last experiment, LP-FP-2, simulated a V-sequence. A great deal of information was recorded on the fission product release, transport, and deposition. A theoretical approach to the chemical behavior of iodine in the blowdown suppression tank (BST) of the LOFT facility was attempted with the IODE and IMPAIR-2/M codes. The comparison of the predictions with the existing experimental data led to the conclusion that the BST system behaved as a low-volatility system, with most of the iodine in the form of the soluble nonvolatile species iodide. Only a partial conversion to volatile molecular iodine was observed due to the presence of radiation. However, the intensity of the γ field was so weak that this transformation was not quantitatively meaningful.

Hydrogen Generation Behavior in the LOFT FP-2 and Other Experiments: Comparative Assessment for Mitigated Severe Accident Conditions

In this paper Zircaloy oxidation and hydrogen generation data for the loss-of-fluid test (LOFT) FP-2 test are presented and compared with findings from other severe fuel damage experiments. In the LOFT FP-2 test, the majority of hydrogen generation occurred as a consequence of bundle reflooding, where significant hydrogen production was also noted in other reflood experiments and in the Three Mile Island Unit 2 accident. Common findings also indicate that during fuel uncovery, bundle oxidation is largely controlled by steam supply conditions, that high rates of hydrogen production continue after melt formation and relocation, and that partial flow-area blockages do not drastically reduce the rates of hydrogen production. Tests results thus indicate no apparent limitations to Zircaloy oxidation other than that due to steam supply conditions and known reaction kinetics, and that the potential for significant hydrogen generation exists during reflooding of cores containing molten metallic debris.

LOFT experiment LP-02-6 analysis by RELAP5/MOD2 code with improved minimum film boiling temperature.

Groeneveld-Stewart's minimum film boiling temperature correlation was incorporated into the RELAP5/MOD2 code in order to explicitly define the minimum film boiling temperature. The transition boiling curve in the code was also modified. The Loss-of-Fluid Test (LOFT) experiment, Experiment LP-02-06 which was a cold-leg double-ended break LOCA experiment with minimum emergency core coolant injection, was analyzed with the modified RELAP5/MOD2 code. The modified RELAP5/MOD2 code well calculated system transients including the rod surface temperature transient. The temporary rewetting of rods in the early phase of blowdown, which had not been predicted by the original RELAP5/MOD2 and other codes, was predicted by the modified RELAP5/MOD2 code.

Results and interpretation of multivariate autoregressive analysis applied to the loft reactor process noise data

Multivariate noise analysis of power reactor signals is useful for characterizing baseline operation and plant diagnostics, for isolating process anomaly from sensor maloperation, and for automated plant monitoring. We have developed a systematic and reliable procedure of accomplishing these tasks by improving upon the previously established techniques of empirical modeling of fluctuation signals in power reactors. The application of the algorithm to data from the Loss-of-Fluid Test (LOFT) reactor showed that earlier results (based on physical modeling) regarding the perturbation sources in a pressurized water reactor (PWR) affecting coolant temperature and neutron power fluctuations can be explained using multivariate autoregressive (MAR) analysis. This methodology has important implications regarding plant diagnostics, and system or sensor anomaly detection.

Evaluation of PWR-LOCA under LOFT L2-3 condition.

Analysis of large break LOCA (loss-of-coolant accident) of a commercial PWR (pressurized water reactor) is made with the THYDE-P2 code under the LOFT (Loss-of-Fluid Test) L2-3 condition. Subcooled reflooding is calculated to occur, whereas in LOFT L2-3 saturated reflooding seems to have been observed. Except for this fact, despite the thermal and geometrical differences, the calculated results show that the LOFT system well simulates PWR-LOCA under the LOFT L2-3 condition. The existing safety criteria are satisfied by the commercial PWR's with a big margin in case of LOCA under the LOFT L2-3 condition.

Application of Noise Analysis Technique for Monitoring the Moderator Temperature Coefficient of Reactivity in Pressurized Water Reactors

A new technique, based on the noise analysis of neutron detector and core-exit coolant temperature signals, is developed for monitoring the moderator temperature coefficient of reactivity in pressurized water reactors (PWRs). A detailed multinodal model is developed and evaluated for the reactor core subsystem of the loss-of-fluid test (LOFT) reactor. This model is used to study the effect of changing the sign of the moderator temperature coefficient of reactivity on the low-frequency phase angle relationship between the neutron detector and the core-exit temperature noise signals. Results show that the phase angle near zero frequency approaches - 180 deg for negative coefficients and 0 deg for positive coefficients when the perturbation source for the noise signals is core coolant flow, inlet coolant temperature, or random heat transfer.

Stochastic Estimation Approach for the Evaluation of Thermal-Hydraulic Parameters in Pressurized Water Reactors

A method based on the extended Kalman filter is developed for the estimation of the core coolant mass flow rate in pressurized water reactors. The need for flow calibration can be avoided by a direct estimation of this parameter. A reduced-order neutronic and thermal-hydraulic model is developed for the Loss-of-Fluid Test (LOFT) reactor. The neutron detector and core-exit coolant temperature signals from the LOFT reactor are used as measurements in the parameter estimation algorithm. The estimation sensitivity to model uncertainties was evaluated using the ambiguity function analysis. This also provides a lower bound on the measurement sample size necessary to achieve a certain estimation accuracy. A sequential technique was developed to minimize the computational effort needed to discretize the continuous time equations, and thus achieve faster convergence to the true parameter value. The performance of the stochastic approximation method was first evaluated using simulated random data, and then applied to the estimation of coolant flow rate using the operational data from the LOFT reactor at 100 and 65% flow rate conditions.

In-core coolant flow monitoring of pressurized water reactors using temperature and neutron noise

Noise measurements were performed at the Loss-of-Fluid-Test (LOFT) and Sequoyah-1 pressurized water reactors (PWRs) in order to investigate the possibility of inferring in-core coolant velocities from cross-power spectral density (CPSD) phases of core-exit thermocouple and in-core neutron detector signals. These noise measurements were used to investigate the effects of inlet coolant temperature, core flow, reactor power, and random heat transfer fluctuations on the noise-inferred coolant velocities. The effect on the inferred velocities of varying in-core neutron detector and core-exit thermocouple locations was also investigated. Theoretical models of temperature noise were developed, and the results were used to interpret the experimental measurements. Results of these studies indicate that the neutron detector/thermocouple phase is useful for monitoring core flow in PWRs. Our results show that the interpretation of the phase between these signals depends on the source of temperature noise, the response times and locations of the sensors, and the neutron dynamics of the reactor. At Sequoyah-1 we found that the in-core neutron detector/core-exit thermocouple phase can be used to infer in-core coolant velocities, provided that the measurements are corrected for the thermocouple response time.

LOFT FUEL ROD TRANSIENT DNB PROBABILITY DENSITY FUNCTION STUDIES

Significantly improved calculated DNB safety margins were defined by the development and use of probability density functions (PDF) for transient MDNBR nuclear fuel rods in the Loss of Fluid Test (LOFT) reactor. Calculations for limiting transients and response surface methods were used thereby including transient interactions and trip uncertainties in the MDNBR PDF. Applicability and sensitivity studies determined that the PDF and resulting nominal MDNBR limits are stable, applicable over a wide range of potential input parameters, and applicable to most transients.

RETRAN-02 Analysis of Loss-of-Fluid Test Experiments L9-3 and L9-4

Two simulations of pressurized water reactor (PWR) anticipated transient without scram (ATWS) sequences were performed in the loss-of-fluid test (LOFT) facility. These were designated as tests L9-3 and L9-4. Test L9-3 is a loss-of-feedwater transient, while L9-4 simulates an ATWS accompanied by loss of off-site power. In the latter case, the main steam valve and primary circulation pumps are tripped at the beginning of the experiment, along with the steam generator feedwater flow. The behavior of these experiments was analyzed with the RETRAN-02 computer code in order to evaluate the capability of the code to predict ATWS behavior against the experimental evidence exhibited by LOFT. The complex sequence of events, which occurs in L9-3, creates a difficult system modeling problem. The primary influence on the entire system response is the steam generator heat transfer rate (given core power as a boundary condition). Results of the analysis demonstrate how the course of the computer calculation is influenced by the steam generator model and its initial conditions. Test L9-4 does not contain the thermal-hydraulic complexity of test L9-3 due to the immediate isolation of the steam generator. Nonetheless, it reinforces the general conclusion that RETRAN-02 will produce an adequate simulation of PWR more » ATWS behavior if the initial and boundary conditions are completely defined. « less

In-core coolant flow monitoring of pressurized water reactors using temperature and neutron noise

Noise measurements were performed at the Loss-of-Fluid-Test (LOFT) and Sequoyah-1 pressurized water reactors (PWRs) in order to investigate the possibility of inferring in-core coolant velocities from cross-power spectral density (CPSD) phases of core-exit thermocouple and in-core neutron detector signals. These noise measurements were used to investigate the effects of inlet coolant temperature, core flow, reactor power, and random heat transfer fluctuations on the noise-inferred coolant velocities. The effect on the inferred velocities of varying in-core neutron detector and core-exit thermocouple locations was also investigated. Theoretical models of temperature noise were developed, and the results were used to interpret the experimental measurements. Results of these studies indicate that the neutron detector/thermocouple phase is useful for monitoring core flow in PWRs. Our results show that the interpretation of the phase between these signals depends on the source of temperature noise, the response times and locations of the sensors, and the neutron dynamics of the reactor. At Sequoyah-1 we found that the in-core neutron detector/core-exit thermocouple phase can be used to infer in-core coolant velocities, provided that the measurements are corrected for the thermocouple response time.

Summary of the session

Noise measurements were performed at the Loss-of-Fluid-Test (LOFT) and Sequoyah-1 pressurized water reactors (PWRs) in order to investigate the possibility of inferring in-core coolant velocities from cross-power spectral density (CPSD) phases of core-exit thermocouple and in-core neutron detector signals. These noise measurements were used to investigate the effects of inlet coolant temperature, core flow, reactor power, and random heat transfer fluctuations on the noise-inferred coolant velocities. The effect on the inferred velocities of varying in-core neutron detector and core-exit thermocouple locations was also investigated. Theoretical models of temperature noise were developed, and the results were used to interpret the experimental measurements.Results of these studies indicate that the neutron detector/thermocouple phase is useful for monitoring core flow in PWRs. Our results show that the interpretation of the phase between these signals depends on the source of temperature noise, the response times and locations of the sensors, and the neutron dynamics of the reactor. At Sequoyah-1 we found that the in-core neutron detector/core-exit thermocouple phase can be used to infer in-core coolant velocities, provided that the measurements are corrected for the thermocouple response time.