Measured Plant Data Research Articles

Estimating Non-Linear Heat Flux in Continuous Billet Casting Process With Limited Sensors

In a continuous billet caster, the heat transfer from the pool of molten steel to the cooling water via the copper mold depends non-linearly upon several casting parameters. A reliable and accurate methodology based on the principle of inverse heat transfer technique has been proposed in this work to obtain the nonlinear heat flux of a continuous billet casting mold. Differential evolution optimization technique has been used, in conjunction with the 2-D and 3-D steady state conduction heat transfer equations to develop the complete algorithm. It has been observed that both the 2-D and 3-D models are able to accurately predict the heat flux for different numbers and locations of temperature sensors. While the 2-D model estimates the 1-D heat flux profile varying along casting direction within a very short time period, the 3-D model offers the advantage of estimating a more accurate 2-D heat flux profile varying along both the mold length and width. The 2-D model has also been used to estimate the heat flux for an industrial continuous casting mold using measured plant data. It is observed that the present methodology accurately estimates the boundary heat flux in the continuous billet caster mold.

A simulation-based prediction model for coal-fired power plant condenser maintenance

Many coal-fired power plants (CFPPs) face special challenges such as ageing infrastructure, neglected equipment maintenance, and reduced funding. To assist with cost-effective decision-making, an integrated, semi-empirical thermohydraulic simulation model for a CFPP was developed. The model was calibrated with actual measurements to ensure accurate predictive maintenance schedules. This model was then integrated with economics and maintenance models. Such a fully comprehensive investigation could not be found in literature. As condensers have an important impact on maximum generation output, the simulation model was used to provide new insight into the cost implications associated with delayed or premature maintenance of a condenser. The results of the thermohydraulic simulation were verified against measured plant data and was within 5% error. The model was then utilised to determine the most efficient predictive maintenance schedule for the condensers of a 60 MW and 350 MW unit. Results showed that, when applying the new predictive maintenance schedule, considerable savings can be realised. When extrapolated to the entire South African power utility fleet, the potential savings are over R900 million11USD 1 = R14.26 (6 January 2020) (USD 63million) per annum. The most important contribution of this paper is that the approach described here can be used to investigate the effect on plant performance and economic viability of any change to any component in any CFPP in a truly industrial scenario.

Integrated Simulation of Coal-Fired Power Stations

South African coal-fired power plants (CFPPs) are set in a unique environment and are faced with unique challenges including ageing infrastructure, increased maintenance requirements, and reduced funds. A unique solution is required whereby plant performance can be maintained, or even enhanced, through the use of efficient techniques that utilise the available funds in a cost-effective manner. A simulation-backed model could help improve the effectiveness of operation and minimise downtime. Simulation software has become an intricate part of the operation of modern companies, aiding in decision making, process predictions, scenario investigations, and optimising operations. In order to compare the available CFPP simulation software, six functional evaluation criteria were investigated, namely; training, digital twinning, scenario investigation, optimisation, real-time monitoring, and dynamic analysis. With these criteria in mind, a simulation model was constructed of a South African-based CFPP. The model is based on a semiempirical thermohydraulic model. The results of the simulation were verified against measured plant data and was within 5% accuracy. Correct application of the model contributed towards maintenance scheduling, with an average daily profit per unit of R1.13 million. When extrapolated to all Eskom power stations, this value increases to over R1.6 billion per annum.

Verification and validation of STREAM/RAST-K for PWR analysis

Abstract This paper presents the verification and validation (V&V) of the STREAM/RAST-K 2.0 code system for a pressurized water reactor (PWR) analysis. A lattice physics code STREAM and a nodal diffusion code RAST-K 2.0 have been developed by a computational reactor physics and experiment laboratory (CORE) of Ulsan National Institute of Science and Technology (UNIST) for an accurate two-step PWR analysis. The calculation modules of each code were already verified against various benchmark problems, whereas this paper focuses on the V&V of linked code system. Three PWR type reactor cores, OPR-1000, three-loop Westinghouse reactor core, and APR-1400, are selected as V&V target plants. This code system, for verification, is compared against the conventional code systems used for the calculations in nuclear design reports (NDRs) and validated against measured plant data. Compared parameters are as follows: critical boron concentration (CBC), axial shape index (ASI), assembly-wise power distribution, burnup distribution and peaking factors. STREAM/RAST-K 2.0 shows the RMS error of critical boron concentration within 20 ppm, and the RMS error of assembly power within 1.34% for all the cycles of all reactors.

RELAP5/MOD3.3 Analysis of the Loss of External Power Event with Safety Injection Actuation

The code assessment typically comprises basic tests cases, separate effects test, and integral effects tests. On the other hand, the thermal hydraulic system codes like RELAP5/MOD3.3 are primarily intended for simulation of transients and accidents in light water reactors. The plant measured data come mostly from startup tests and operational events. Also, for operational events the measured plant data may not be sufficient to explain all details of the event. The purpose of this study was therefore besides code assessment to demonstrate that simulations can be very beneficial for deep understanding of the plant response and further corrective measures. The abnormal event with reactor trip and safety injection signal actuation was simulated with the latest RELAP5/MOD3.3 Patch 05 best-estimate thermal hydraulic computer code. The measured and simulated data agree well considering the major plant system responses and operator actions. This suggests that the RELAP5 code simulation is good representative of the plant response and can complement not available information from plant measured data. In such a way, an event can be better understood.

Estimation of surface heat flux in continuous casting mould with limited measurement of temperature

Continuous casting mould is exposed to high heat flux due to its contact with the hot molten metal. The state of heat transfer in a mould and steel solidification depends on the magnitude of the mould boundary heat flux. In the present study, a mathematical model is developed to determine the heat flux across the hot surface of continuous casting mould from limited temperature measurement. The model is based on two-dimensional inverse heat transfer technique that was solved using the conjugate gradient method. Direct problem which involves two-dimensional conduction in the mould is first solved and validated. The inverse problem was tested by using the simulated temperature data obtained from the solution of a direct problem, where good agreement between the actual and estimated boundary heat flux is found. The Gaussian noises are added to the simulated temperatures to mimic the temperature measurement errors for testing anti-noise ability of the inverse problem model. Further, boundary heat flux is also estimated for a continuous casting mould using actually measured plant data. The model is applied to three test cases with temperature data obtained under different operating conditions and the results are analyzed. It has been observed that the proposed methodology results in accurate boundary heat flux estimation. Higher boundary heat flux are obtained for cases with higher casting speed as compared to cases with lower casting speed.

Application of the ATHLET/COBRA-TF thermal-hydraulics coupled code to the analysis of BWR ATWS

Validation and qualification of thermal-hydraulics system codes based on measured plant data are essential for their reliability when applied to nuclear power plant analyses. To this purpose, the Institute for Neutron Physics and Reactor Technology (INR) at the Karlsruhe Institute of Technology (KIT) is involved in various validation and qualification activities of different CFD, subchannel and system codes. This paper is focused on the application of multi-scale thermal-hydraulics approach (combination of subchannel and system codes) for the analysis of BWR transients such as the Anticipated Transients Without Scram (ATWS).A description of the thermal-hydraulics coupling methodology between ATHLET/COBRA-TF is given. The multi-scale thermal-hydraulics coupling was implemented inside the SALOME platform using the MEDCoupling interface and some of the NURESIM platform interpolators. Finally, the coupled ATHLET/COBRA-TF model for the Oskarshamn-2 NPP is presented and the main results of the ATWS transient analysis are discussed. The work was carried out in the frame of the European NURESAFE (NUclear REactor SAFEty Simulation Platform) project of the 7th EURATOM Framework Program, aimed at addressing the engineering aspects of nuclear safety of light water reactors.

BWR loss of coolant accident simulation by means of RELAP5

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

Kinetic modeling and optimization of Claus reaction furnace

Optimization of Claus sulfur recovery plants requires a reliable kinetic model to describe the reaction furnace and waste heat boiler. In order to develop a suitable kinetic model for prediction of reaction furnace and waste heat boiler effluents, several reaction schemes were tested and the best reaction set was selected. The kinetic parameters of the best reaction scheme were optimized and different reactor networks were analyzed using the optimum reaction set. The modified model results were in good agreement with the experimental measured plant data both before and after waste heat boiler (The mean absolute errors were 7.62% and 7.88% respectively). In addition, the effects of increasing ammonia and oxygen content of feed, and air to acid gas ratio on reaction furnace effluents were investigated through various case studies. An optimization study was performed to control the operating variables of an operating sulfur recovery unit in order to obtain optimum conditions (including maximum reaction furnace temperature), maintain H2S to SO2 ratio at constant 2 and lower COS emissions from waste heat boiler.

Hot zero power reactor calculations using the Insilico code

In this paper we describe the reactor physics simulation capabilities of the Insilico code. A description of the various capabilities of the code is provided, including detailed discussion of the geometry, meshing, cross section processing, and neutron transport options. Numerical results demonstrate that Insilico using an SP N solver with pin-homogenized cross section generation is capable of delivering highly accurate full-core simulation of various pressurized water reactor problems. Comparison to both Monte Carlo calculations and measured plant data is provided.

Medium Density Control for Coal Washing Dense Medium Cyclone Circuits

The dense medium cyclone (DMC) process used in coal beneficiation plants is studied from a control system perspective. Employing the dynamic model of the DMC process derived from mass flow balance, a model-based control strategy is proposed. The controller adjusts the density of medium used to enhance separation in the DMC process according to measurements on percentages of different components in raw coal. The first objective of the control is to maintain the carbon content in the clean coal to a set level. The second purpose is to minimize energy consumption of the DMC process in view of the fast increasing electricity price. The controller solves an optimization problem formulated to determine the density of medium whenever new measurements are available. Both coal quality and DMC operational constraints are accounted for. Simulations, based on measured plant data, are carried out to verify the feasibility and effectiveness of the control strategy designed. The results show that the designed controller is able to fulfill its purpose satisfactorily when the characteristics of the raw coal varies and when measurement uncertainties are in presence.

Analysis of the OECD/NRC Oskarshamn-2 BWR stability benchmark

On February 25, 1999, the Oskarshamn-2 NPP experienced a stability event which culminated in diverging power oscillations. The event was successfully modeled by the TRACE/PARCS coupled system code, and further uncertainty analysis of the event is described in this paper. The results show very good agreement with the plant data, capturing the entire behavior of the transient including the onset of instability, growth of the oscillations, and oscillation frequency. This provides confidence in the prediction of other parameters which are not available from the plant records. The event provides coupled code validation for a challenging BWR stability event, which involves the accurate simulation of neutron kinetics (NK), thermal–hydraulics (TH), and TH/NK coupling. The success of this work has demonstrated the ability of the 3-D coupled systems code TRACE/PARCS to capture the complex behavior of BWR stability events. The problem was released as an international OECD/NEA benchmark, and it is the first benchmark based on measured plant data for a stability event with a DR greater than one.

Comparison of power plant steam condenser heat transfer models for on-line condition monitoring

In this paper heat transfer models for large power plant condenser were examined. The goal was to develop a model capable of predicting not only the condenser pressure but the overall heat transfer coefficient. Such a model can be used for condenser condition monitoring. The results of a two-dimensional (2-D) condenser heat transfer model and single-point, zero-dimensional (0-D) model are presented together with the results from Heat Exchanger Institute (HEI) standards curves. Both 0-D and 2-D models can account for the effects of steam-side pressure drop and in a simplified manner also some effects of tube bundle geometry. For all models an experimental correction as a function of cooling water temperature was implemented to improve their accuracy. The results are presented in comparison with the measured plant data for three different tube bundle geometries, with and without the experimental correction factor. The 2-D model proved to be the most consistently accurate of the models both without the correction, and at varying steam and coolant flow with the correction applied. The results indicate significant local variation of pressure drop related effects, which the 0-D model failed to accurately predict particularly in cases of close temperature approach. In predicting the heat transfer coefficient the HEI model was the least accurate, significantly overestimating the impact of coolant flow rate change, and failing to match the measurements even with a correction applied.

Numerical investigation of thermal–hydraulic characteristics in a steam generator using a coupled primary and secondary side heat transfer model

A coupled primary and secondary side heat transfer and thermal phase change model is used to investigate the thermal–hydraulic characteristics of a steam generator (SG) at Daya Bay Nuclear Power Plant (DBNPP). The simulation results reasonably reveal the ununiform boiling behavior in the secondary hot and cold legs. Vapor velocity is slightly higher than that of water and the corresponding slip-ratio first increases rapidly before gradually decreasing in the secondary hot leg and cold leg regions, a result which agrees with predictions using the drift-flux model. Cross-flow energy, which accounts for flow-induced vibration (FIV) at the U-bend tubes, is determined with the aid of localized thermal–hydraulic distributions, and the resulting FIV damage is predicted to be most severe at 0.35m on the cold leg side and −0.2m on the hot leg side of the U-bend region, respectively. These FIV damage predictions agree with measured plant data for the prototypical SG, showing that this model can provide the use information to improve thermal–hydraulic characteristics and help alleviate FIV damage in a SG.

Validation of Simulation Model for Cogeneration Power and Water Desalination Plant

Cogeneration power and MSF water desalination plant has been modelled using the IPSEpro software package based on plant operational scenarios and validated against measured recorded data from the plant. The relative differences between the model results and measured plant data vary from 1.1% to 3.7% for the power plant and 1.0 % to 1.8 % for MSF desalination. The model uncertainties could be attributed to either modelling assumptions or to input data uncertainties, with measured plant performance uncertainties due to measurement device precision and effects of external factors.

Development of life assessment procedures for power plant headers operated under flexible loading scenarios

A finite element methodology for thermo-mechanical fatigue analysis of a subcritical power plant outlet header under realistic loading conditions is presented. The methodology consists of (i) a transient heat transfer model, (ii) a sequential anisothermal cyclic viscoplastic model and (iii) a multiaxial, critical-plane implementation of the Ostergren fatigue indicator parameter. The methodology permits identification of the local thermo-mechanical stress–strain response at critical locations and prediction of fatigue life and cracking orientation for complex transient, anisothermal, cyclic elastic–plastic-creep material behaviour. Measured plant data, in the form of steam and pipe temperature transients and steam pressure data, are employed to identify heat transfer constants and validate the predicted thermal response, with particular attention given to plant start-up and attemperation effects. The predictions indicate out-of-phase temperature-strain response at the header inside surface and in-phase response on the outside surface. Cooling transients are predicted to control damage and crack initiation at the inner bore, whereas heating transients are predicted to have a more damaging effect at weld locations. A representative test cycle is presented, which is shown to capture the salient thermo-mechanical cyclic damage of the realistic cycle. The predicted results correlate well with industrial experience in terms of crack (initiation) orientation, location and life.

Numerically investigating the influence of tube support plates on thermal-hydraulic characteristics in a steam generator

A three-dimensional “unit pipe” model coupled with broached tube support plates (TSPs) is developed to investigate the influence of TSPs on the thermal-hydraulic characteristics of the primary and secondary side of steam generator (SG). The model reasonably captures the unbalanced boiling phenomena in the SG secondary side, particularly in the regions surrounding the TSPs. At these points, there is an abrupt increase in secondary fluid velocity, generating recirculation distribution and local vortex. Nonuniform outer wall temperatures periodically fluctuate around the circumference of the tubes at the TSPs, with the corresponding temperature difference between the support location and flow hole gradually decaying in the primary flow direction. These thermal-hydraulic distributions are then used to determine the cross-flow energy, which is strongly related to flow-induced vibration (FIV) damage. According to the FIV distributions, FIV damage is predicted to be most severe at angles of θ = 65° on the cold-leg side and θ = 120° on the hot-leg side of the U-bend tube bundles. These simulated FIV damage results are in agreement with the measured plant data from a prototypical SG. Therefore, this developed “unit pipe” model can provide technical support to help alleviate FIV damage and minimize the probability of tube failure.

Thermal combustion of lean methane–air mixtures: Flow reversal research and demonstration reactor model and its validation

Problem of the coal mine shafts ventilation air methane (VAM) utilization is briefly presented. A research and demonstration plant experiments were carried out as the first step to a future industrial application. The plant contained a thermal flow reversal reactor (TFRR) as the main unit where lean methane–air mixture is combusted. The study consisted of the plant experiments and mathematical model simulations as well. The paper presents the mathematical model description and its experimental validation based on comparison of the model results with measured plant data. The aim of the paper is to assess whether dynamic, one dimensional (in space) model may be useful for a fairly large scale reactor simulations. The model was developed and tested in the 2 versions: without and with heat accumulation in the reactor wall.

Validation and Application of the Thermal Hydraulic System Code TRACE for Analysis of BWR Transients

The Karlsruhe Institute of Technology (KIT) is participating on (Code Applications and Maintenance Program) CAMP of the US Nuclear Regulatory Commission (NRC) to validate TRACE code for LWR transient analysis. The application of TRACE for the safety assessment of BWR requires a throughout verification and validation using experimental data from separate effect and integral tests but also using plant data. The validation process is normally focused on safety-relevant phenomena for example, pressure drop, void fraction, heat transfer, and critical power models. The purpose of this paper is to validate selected BWR-relevant TRACE-models using both data of bundle tests such as the (Boiling Water Reactor Full-Size Fine-Mesh Bundle Test) BFBT and plant data recorded during a turbine trip event (TUSA) occurred in a Type-72 German BWR plant. For the validation, TRACE models of the BFBT bundle and of the BWR plant were developed. The performed investigations have shown that the TRACE code is appropriate to describe main BWR-safety-relevant phenomena (pressure drop, void fraction, and critical power) with acceptable accuracy. The comparison of the predicted global BWR plant parameters for the TUSA event with the measured plant data indicates that the code predictions are following the main trends of the measured parameters such as dome pressure and reactor power.

BWR stability analyses with SIMULATE-3K benchmark against measured plant data

The best-estimate coupled neutronic/thermal-hydraulics code, SIMULATE-3K (S3K), is used by many utilities, research institutes, and regulatory authorities in Europe for performing BWR stability analysis. Analysis of many measured BWR stability tests (often performed in European BWRs) provides the basis for the validation for stability parameter calculations (decay ratio and natural frequency) with S3K. This paper summarizes part of the extensive validation database for the code, and discusses the influence of fuel pin model parameters on the stability results.