Core Exit Thermocouples Research Articles

ROSA/LSTF test and RELAP5 code analyses on PWR 1% vessel upper head small-break LOCA with accident management measure based on core exit temperature

Abstract An experiment was performed using the large-scale test facility (LSTF), which simulated a 1% vessel upper head small-break loss-of-coolant accident with an accident management (AM) measure under an assumption of total-failure of high-pressure injection (HPI) system in a pressurized water reactor (PWR). In the LSTF test, liquid level in the upper head affected break flow rate. Coolant was manually injected from the HPI system into cold legs as the AM measure when the maximum core exit temperature reached 623 K. The cladding surface temperature largely increased due to late and slow response of the core exit thermocouples. The AM measure was confirmed to be effective for the core cooling. The RELAP5/MOD3.3 code indicated insufficient prediction of primary coolant distribution. The author conducted uncertainty analysis for the LSTF test employing created phenomena identification and ranking table for each component. The author clarified that peak cladding temperature was largely dependent on the combination of multiple uncertain parameters within the defined uncertain ranges.

Verification of SAMGs in SBO sequences with Seal LOCA. Multiple damage domains

The Integrated Safety Assessment (ISA) methodology, developed by the Spanish Nuclear Safety Council, has been applied to PWR Station Blackout (SBO) sequences in the context of the IDPSA (Integrated Deterministic-Probabilistic Safety Assessment) network. The ISA methodology allows obtaining the Damage Domain (DD), the region of the uncertain parameters space where the damage limit is exceeded, for each sequence of interest as a function of the operator actuation times. Several damage limits have been taken into account within this analysis: cladding embrittlement criteria (Peak cladding temperature >1477K); Inadequate core cooling conditions (Core Exit Thermocouples temperature >922K); local fuel melting (fuel temperature >2499K); fuel relocation in lower plenum and vessel failure. Other continuous damages, such as percentage of relocated fuel are also studied. Every one of these damage variables provides a specific DD. The application to the severe accident management (SAM) actions shows the capability of a methodology such as ISA in order to analyze the impact of different SAM strategies and to obtain the available times for different operator actions.

Experimental approach to investigate the dynamics of mixing coolant flow in complex geometry using PIV and PLIF techniques

The aim of this work is to investigate experimentally the increase of mixing phenomenon in a coolant flow in order to improve the heat transfer, the economical operation and the structural integrity of Light Water Reactors-Pressurized Water Reactors (LWRs-PWRs). Thus the parameters related to the heat transfer process in the system will be investigated. Data from a set of experiments, obtained by using high precision measurement techniques, Particle Image Velocimetry and Planar Laser-Induced Fluorescence (PIV and PLIF, respectively) are to improve the basic understanding of turbulent mixing phenomenon and to provide data for CFD code validation. The coolant mixing phenomenon in the head part of a fuel assembly which includes spacer grids has been investigated (the fuel simulator has half-length of a VVER 440 reactor fuel). The two-dimensional velocity vector and temperature fields in the area of interest are obtained by PIV and PLIF technique, respectively. The measurements of the turbulent flow in the regular tube channel around the thermocouple proved that there is rotation and asymmetry in the coolant flow caused by the mixing grid and the geometrical asymmetry of the fuel bundle. Both PIV and PLIF results showed that at the level of the core exit thermocouple the coolant is homogeneous. The discrepancies that could exist between the outlet average temperature of the coolant and the temperature at in-core thermocouple were clarified. Results of the applied techniques showed that both of them can be used as good provider for data base and to validate CFD results.

An analysis on the severe accident progression with operator recovery actions

Analyses on the severe accident progressions for the station blackout (SBO) accident and small break LOCA (SBLOCA) initiated severe accident were performed for APR1400 by using MELCOR computer code. Operator recovery actions for decay heat removal and inventory make up using a depressurization system and safety injection pump were simulated in parallel with a simulation of the severe accident progression. Sensitivity studies on the operator actions were performed to investigate the changes in the timing of the reactor vessel failure and to determine the time allowed for the operator to prevent reactor vessel failure. Sensitivity analyses on the effect of major modeling parameters were performed additionally to quantify the uncertainties in timing. It is found that the operator has about 2h for the recovery actions after the indication of core damage by the signal of core exit thermocouple (CET) for the SBLOCA initiated severe accident, while the operator has to take immediate actions after the indication of core damage by CET for the SBO accident.

Reliability of Core Exit Thermocouple for Accident Management Action during SBLOCA and Abnormal Transient Tests at ROSA/LSTF

Presented are experimental results on the general performances of core exit thermocouples (CETs) to detect core overheat for accident management (AM) action by using the Large-Scale Test Facility (LSTF) of the ROSA Program of the Japan Atomic Energy Agency. The LSTF is a full-height, full-pressure, and 1/48-volumetric-scaled model of a 4-loop pressurized water reactor (PWR). This study was motivated by a significant delay in the time and temperature rise of the CETs from core heat-up during a vessel top head small break loss-of-coolant accident (SBLOCA) test. A certain delay in time and temperature rise of the CETs was also observed in various SBLOCA and abnormal transient tests. Such CET performances are derived from thirteen LSTF tests as follows: (1) general CET performances are obtained in the form of equations including cases under limited influences of water fall-back from hot legs, (2) the major reason for the delay is the interaction of three-dimensional steam flows with low-temperature structures in and around the core and core exit, (3) break location was insignificant except for the PV top and bottom break cases, and (4) CET superheat is suitable for AM instead of the temperature value for significantly high or low pressure transients. The applicability of the results to PWR is discussed further.

Reliability of Core Exit Thermocouple for Accident Management Action during SBLOCA and Abnormal Transient Tests at ROSA/LSTF

Presented are experimental results on the general performances of core exit thermocouples (CETs) to detect core overheat for accident management (AM) action by using the Large-Scale Test Facility (LSTF) of the ROSA Program of the Japan Atomic Energy Agency. The LSTF is a full-height, full-pressure, and 1/48-volumetric-scaled model of a 4-loop pressurized water reactor (PWR). This study was motivated by a significant delay in the time and temperature rise of the CETs from core heat-up during a vessel top head small break loss-of-coolant accident (SBLOCA) test. A certain delay in time and temperature rise of the CETs was also observed in various SBLOCA and abnormal transient tests. Such CET performances are derived from thirteen LSTF tests as follows: (1) general CET performances are obtained in the form of equations including cases under limited influences of water fall-back from hot legs, (2) the major reason for the delay is the interaction of three-dimensional steam flows with low-temperature structures in and around the core and core exit, (3) break location was insignificant except for the PV top and bottom break cases, and (4) CET superheat is suitable for AM instead of the temperature value for significantly high or low pressure transients. The applicability of the results to PWR is discussed further.

Nuclear plant temperature instrumentation

Most critical process temperatures in nuclear power plants are measured using resistance temperature detectors (RTDs) and thermocouples. In addition to excellent reliability and accident survivability, nuclear safety-related RTDs are expected to have good calibration and fast dynamic response time, as these characteristics are important to plant safety and economy. In plants where RTDs are installed in thermowells in the primary coolant pipes, response-time requirements have a range of 4.0–8.0 s versus the direct-immersion RTDs installed in bypass loops which have a required response range of 1.0–3.0 s. The variety of problems that can affect the accuracy and response time of RTDs is extensive: dynamic response problems, failure of extension leads, low-insulation resistance, premature failure, wrong calibration tables, loose or bad connections, large EMF effects, open elements, thinning of the platinum wire, lead-wire imbalance, seeping of chemicals from the connection head into the thermowell, cracking of the thermowell, and erroneous indication. The causes of core-exit thermocouples failure can take the form of large calibration shifts, erratic and noisy output, saturated output, accidental reverse connections, and response-time degradation. Several effective methods for detecting RTD and thermocouple performance failure while the plant is operating are available. To detect accuracy problems, the cross-calibration technique is effective for both RTDs and core-exit thermocouples. It involves recording the readings of redundant online RTDs, averaging these readings, and calculating the deviation of each RTD from the average, less any outliers. To detect response time degradation online, the loop current step response (LCSR) test is the most accurate method. However, the noise analysis technique remains the most popular for detecting response time degradation in core-exit thermocouples.

Performance of Core Exit Thermocouple for PWR Accident Management Action in Vessel Top Break LOCA Simulation Experiment at OECD/NEA ROSA Project

Presented are experiment results of the Large Scale Test Facility (LSTF) conducted at the Japan Atomic Energy Agency (JAEA) with a focus on core exit thermocouple (CET) performance to detect core overheat during a vessel top break loss-of-coolant accident (LOCA) simulation experiment. The CET temperatures are used to start accident management (AM) action to quickly depressurize steam generator (SG) secondary side in case of core temperature excursion. Test 6-1 is the first test of the OECD/NEA ROSA Project started in 2005, simulating withdraw of a control rod drive mechanism penetration nozzle at the vessel top head. The break size is equivalent to 1.9% cold leg break. The AM action was initiated when CET temperature rose up to 623K. There was no reflux water fallback onto the CETs during the core heat-up period. The core overheat, however, was detected with a time delay of about 230s. In addition, a large temperature discrepancy was observed between the CETs and the hottest core region. This paper clarifies the reasons of time delay and temperature discrepancy between the CETs and heated core during boil-off including three-dimensional steam flows in the core and core exit. The paper discusses applicability of the LSTF CET performance to pressurized water reactor (PWR) conditions and a possibility of alternative indicators for earlier AM action than in Test 6-1 is studied by using symptom-based plant parameters such as a reactor vessel water level detection.

Characterization of Unsteady Core Flow Patterns through Noise Analysis Techniques

This work describes the application of noise analysis techniques to study the unsteady flow patterns in the core of nuclear reactor plants. The study is focused on the determination of the position of the flow perturbations and their causes. The analysis was performed using sampled data collected from a four loop plant instrumentation that included the loop flow rates, the ex-core neutron detectors and the core exit thermocouples. As part of the analysis, the thermocouple signals were cross-correlated to the flow data to determine approximate flow perturbation mappings in the core. The results of the analysis illustrate the value of noise analysis as a tool for the characterization of the flow conditions in the core.

3533 PWRアクシデントマネジメントと原子炉計装の役割に関するROSA-V実験研究(S49-4 流動,原子炉用機器(3),S49 原子炉システムおよびその要素技術)

Shown below are experimental results on characteristics of reactor instrumentations including a coolant mass tracking method and core exit thermocouples (CETs) which are necessary to precise operator actions for accident management (AM) during a loss-of-coolant accident (LOCA) at a pressurized water reactor (PWR). The experiments at the ROSA-V/LSTF facility of the Japan Atomic Energy Research Institute simulated small break LOCAs at the PWR vessel bottom and clarified effects of secondary depressurization as one of the AM measures in case of high pressure injection system failure and non-condensable gas inflow from the accumulator injection system. It was shown that the coolant mass tracking method based on three types of water level instruments could detect most of the primary coolant mass change between the initial state and core-heatup starting condition. The CET characteristics to detect the core heatup conditions were significantly degraded by the condensed water fall-back during the secondary depressurization action.

Development of a Method for Measuring the Moderator Temperature Coefficient by Noise Analysis and Its Experimental Verification in Ringhals-2

This paper deals with the estimation of the moderator temperature coefficient of reactivity (MTC) by noise analysis. The current noise analysis-based MTC measurement, relying on the cross correlation between the neutron noise measured by a single in-core neutron detector and the local temperature noise given by a single core-exit thermocouple located at the top of the same fuel assembly, or of a neighboring fuel assembly, is not accurate. The MTC is systematically underestimated by a factor of 2 to 5 compared to its design-predicted value. A theoretical study shows that, in case of nonhomogeneous moderator temperature noise, the core-averaged moderator temperature noise should be used for the MTC estimation. The new estimation method can reach up to 3% accuracy as compared with the results of core calculations for the Swedish Ringhals-2 pressurized water reactor (PWR). We show via noise measurements performed at the Ringhals-2 PWR that the moderator temperature noise is actually radially strongly heterogeneous and loosely coupled. The new MTC noise estimator is demonstrated to provide an accurate MTC evaluation, with the core-averaged moderator temperature noise estimated via the use of many radial in-core gamma-thermometers. More important, different forms of weighting functions are suggested to calculate the core-averaged moderator temperature noise. This new MTC noise estimator, which is nonintrusive and free of calibration, can therefore be applied to monitor the MTC throughout the cycle.

Analysis of an MTC noise measurement performed in ringhals-2 using gamma-thermometers and in-core neutron detectors

A noise measurement in the Swedish Ringhals-2 PWR was performed in January 2002 by using twelve gamma-thermometers and two in-core neutron detectors, all located on the same axial level in the reactor. The gamma-thermometers are very versatile tools since they allow estimating the core-averaged moderator temperature noise throughout the core. This core-averaged temperature noise was then used to estimate the MTC by noise analysis, via a new MTC noise estimator. It was shown that whatever the location of the neutron detector might be, the MTC is always correctly estimated by this new MTC noise estimator, without any calibration to a known value of the MTC prior to the noise measurement. For the purpose of comparisons, the MTC was also estimated by using a single gamma-thermomemeter and a single core-exit thermocouple, together with an in-core neutron detector. In such cases, the MTC was systematically underestimated, with a stronger bias for the core-exit thermocouple than for the gamma-thermometer. This shows that the main reason of the MTC underestimation by noise analysis in all the experimental work until now was due to the radially non-homogeneous temperature noise throughout the core. The resulting deviation from point-kinetics of the reactor response has a negligible effect.

Monitoring of the temperature reactivity coefficient at the PWR nuclear power plant

For monitoring temperature coefficient of reactivity of pressurized water reactors a method based on the correlation of fluctuations in signals of in-core neutron detectors and core-exit thermocouples and neural network paradigm is used. It is shown that the moderator temperature coefficient of reactivity can be predicted with the aid of the backpropagation neural network technique by measuring the frequency response function between the in-core neutron flux and the core-exit coolant temperature.

Noise Analysis Method for Monitoring the Moderator Temperature Coefficient of Pressurized Water Reactors: Neural Network Calibration

A neural network was trained with data for the frequency response function between in-core neutron noise and core-exit thermocouple noise in a pressurized water reactor, with the moderator temperature coefficient (MTC) as target. The trained network was subsequently used to predict the MTC at other points in the same fuel cycle. Results support use of the method for operating pressurized water reactors provided noise data can be accumulated for several fuel cycles to provide a training base.

Noise Analysis Method for Monitoring the Moderator Temperature Coefficient of Pressurized Water Reactors: I. Theory

In this paper a method is described for monitoring the moderator temperature coefficient of reactivity of pressurized water reactors based on the correlation of fluctuations in in-core neutron detector and core-exit thermocouple signals. The theoretical basis of the method is explained.

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.

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.

On-line monitoring of a PWR for plant surveillance by noise analysis

For the real-time surveillance of the nuclear power plant Borssele (PWR) a large set of dynamic reactor signals has been linked since early 1982 through a special telephone line of 200 km to a mini-computer in the laboratory of the Netherlands Energy Research Foundation (ECN) at Petten, in order to follow these signals and to identify changes in statistical characteristics of the signals (mean values and noise components). During the operation of two fule cycles of the power plant such information has been collected at normal operation, but also several planned occurrences have been followed which is intersting for the diagnostic information. Several experiences have been obtained during two years of continuous operation of the instrumentation system of the power plant and it was also possible to deduce physical parameters of the reactor from on-line analysis. This system has also been used for in-situ tests of reactor safety channel instrumentation both at the start-up and during continuous operation of the reactor. In this paper hardware and software of the on-line reactor monitoring system are described and the computing capabilities of the system will be given. In the coarse of the development of the system, the following operating experiences were obtained: • - Practical experience with the operation of the system and its reliability. • - On-line monitoring of noise power of several plant sensors both for short time intervals and for long time trend information. • - In-situ test of the reactor safety channels, by measuring spectra, response time of the sensors, etc. • - Continuous monitoring of primary system vibrations. • - Continuous monitoring of core physics (and related) parameters through noise analysis such as the core coolant velocity, reactivity effects, local vibration of the in-core instrumentation the core coolant velocity, reactivity effects, local vibration of the in-core instrumentation guide tube and core support barrel (CSB) motions using in-/ex-core neutron detector and core exit thermocouple signals. • - Some experiences obtained from the return information provided from the far distance on-line monitoring system. The system allowed easy diagnostics of a nuclear power plant from a far distance. It may also serve as an emergency data retrieval center in case of need of measurements under special conditions or during an unforeseen event.

Sensor Fault Analysis Using Decision Theory and Data-Driven Modeling of Pressurized Water Reactor Subsystems

Instrument fault detection and estimation is important for process surveillance, control, and safety functions of a power plant. The method incorporates the dual-hypotheses decision procedure and system characterization using data-driven time-domain models of signals representing the system. The multivariate models can be developed on-line and can be adapted to changing system conditions. For the method to be effective, specific subsystems of pressurized water reactors were considered, and signal selection was made such that a strong causal relationship exists among the measured variables. The technique is applied to the reactor core subsystem of the loss-of-fluid test reactor using in-core neutron detector and core-exit thermocouple signals. Thermocouple anomalies such as bias error, noise error, and slow drift in the sensor are detected and estimated using appropriate measurement models.