Thermal-hydraulic Computer Code Research Articles

Analytical model development for investigation on hydrogen management studies in the containment of a nuclear reactor

Containment is the ultimate physical barrier to prevent the spread of active air-borne fission products as well as radiation shielding in normal operation as well as accident condition and is designed to enclose whole reactor systems. The design of containment system for Fleet mode of 700 MWe Indian Pressurized Heavy Water Reactors (IPHWRs) incorporates a Primary Containment (PC) of pre-stressed concrete enveloped by a Secondary Containment (SC) of reinforced concrete. For the specification of the design pressure of the containment system, analysis is performed for various postulated accidents such as Loss of Coolant Accident (LOCA) and Main Steam Line Break Accident (MLSB). Further Studies on hydrogen generation and distribution for various accident scenarios in reactor containment of Indian Pressurized Heavy Water Reactors (IPHWR) indicate that during severe accident with core damage, the possibility of global deflagration in absence of mitigating features is high and may challenge the integrity of containment. Therefore, it is essential to incorporate mitigating provisions inside the containment to prevent the hydrogen reaching local detonation and global deflagration during such a scenario.In order to arrive at the design basis of containment system and simulation of large numbers of accident sequence a computer code is required, where mitigating features such as Passive Auto-Catalytic Recombiners, Containment Spray System & Containment Filtered Venting System can be modeled. A system thermal hydraulic computer code has been developed for containment response calculation during normal operation as well as accident conditions including, severe accident to include the new models for mitigating features. This article presents the details of various models of computer code PACSR-SI-2.0, its validation with data observed for hydrogen distribution & removal in Recombiner Test Facility,Battelle Spray Test for containment system & Spray test at IIT Mumbai.

Optimization for fuel loading and reactivity initiated accident analysis for TRR-1/M1

TRR-1/M1 reactor core consists of two types of standard TRIGA fuel, 8.5% and 20% weight uranium. Periodically, the reactor core configuration is rearranged in order to optimize its performance. The objective of this work is to show that it is safe to operate TRR-1/M1 at 1 MW steady state condition for the new core configurations of which the 8.5% weight fuels in the B ring are replaced by high burn up 20% weight fuel elements. To ensure sufficient safety margins of the reactor operation, thermal–hydraulic computer codes, i.e., COOLOD-N2 and EUREKA-2/RR, were coupled with neutronic calculation based on Monte Carlo N-Particle Transport concept in order to analyse some crucial operating parameters, e.g., coolant and fuel temperature in the reactor core as well as Departure from Nucleate Boiling Ratio (DNBR) along the fuel axial direction of the hottest channel. From this study, it can be concluded that no significant effect on the reactor operation was found when replacing the 8.5% weight fuel elements in the B ring with the high burnup 20% weight fuels. Transient phenomena of Reactivity Initiated Accident (RIA) were also performed for the rapid removal of a sample from the reactor core with reactivity insertion rate of $1.00/s. It is found that the negative temperature coefficient of TRIGA fuel provided safety for the reactor even during the prescribed reactor transients. The calculation showed slightly increase in the fuel temperature during the short period of reactivity insertion which caused insignificant energy release to heat up the fuel during the accident. The maximum fuel temperature found during the accident in this study was well below TRR-1/M1 safety.

Extension of the CIRCE methodology to improve the Inverse Uncertainty Quantification of several combined thermal-hydraulic models

The Inverse Uncertainty Quantification (IUQ) of physical correlations used in thermal-hydraulic system codes is a crucial issue in the BEPU (Best Estimate Plus Uncertainty) framework for the licensing and safety analyses of nuclear reactors. The CIRCE methodology is one of the most widely applied IUQ methods. It estimates the (log-)normal probability distribution of a multiplicative factor applied to the reference closure model used in thermal-hydraulic computer codes. CIRCE can jointly estimate the uncertainty of several physical models. However, in some particular cases, a problem of identifiability may occur resulting in less accuracy and precision of the estimated uncertainties.For this reason, in this work, the methodology is improved under the name of CIRCE 2-Steps. This new approach aims at improving the quantification of the model uncertainties when experiments with more than one physical phenomena interacting simultaneously on the output of interest has to be used. In a first step, the uncertainties of the least influential models are evaluated with classic CIRCE methodology using experiments for which only one physical phenomenon is dominant. Then, these uncertainties are properly taken into account when quantifying the remaining uncertainties on the experimental database. This approach is proven to provide more accurate and precise results on several analytical test cases where the uncertainties of two models has to be jointly estimated. The study of the impact of the relative importance of the two models shows that the biggest gain in precision is obtained when the hierarchy of models is such that one model is significantly more influential than the other.Finally, the newly developed methodology is applied to estimate the uncertainties of two condensation heat transfer correlations in the cold leg of a pressurised nuclear reactor.

Development of IVR-ERVC evaluation method and its application to the SMART

An IVR (In-Vessel corium Retention) through ERVC (External Reactor Vessel Cooling) is known to be an effective means for maintaining the reactor vessel integrity during severe accidents in nuclear power plants. The evaluation method of the IVR-ERVC for small reactors was developed and applied to a small integral reactor of the SMART (System-integrated Modular Advanced ReacTor). As a first step, the thermal load analysis from the corium to the external reactor vessel wall is necessary. As a second step, the natural circulation mass flow rate analysis, which is formed in the annular gap between the external reactor vessel wall and insulator, is necessary using the thermal hydraulic computer code. The CHF (Critical Heat Flux), which is the maximum heat removal rate from the external reactor vessel wall depending on the estimated natural circulation mass flow rate, is determined using experimental results. The success criteria of the IVR-ERVC to prevent reactor vessel failure during severe accidents is determined by comparing of the thermal load with the CHF. Finally, a structure analysis on the reactor vessel wall performs to evaluate the structure integrity of the reactor vessel wall in ERVC condition using the structure analysis computer code. Consequently, the IVR-ERVC strategy turned out to be an effective means for maintaining the SMART reactor vessel integrity during severe accidents from the thermal load analysis, CHF analysis, and structure integrity analysis.

Measurements for verification and validation of thermal-hydraulic computer code used for thermal-hydraulic analysis of VVER440 typical fuel assembly

The paper contains experimental data and analysis of the velocity distribution and pressure drop of turbulent flow through a fuel bundle of 19 rods as a partial model of the VVER-440 reactor core. Also, results from the code calculations at the outlet of the test assembly at different flow rates and comparison with experimental data are presented. A pitot tube was used to measure the velocity and pressure drop. Detailed measurements of the velocity distribution in a fuel bundle are done in a fully developed turbulent flow region. Moreover, detailed measurements of the pressure drop of wire and spacer grids typical for PWRs are discussed together with the total pressure drops performance. The experimental result was compared with the predictions by our code and good agreement was found. The calibration protocol and uncertainties of the measurements are presented.

CFD analyses of the European scaled pool experiment E-SCAPE

The European Scaled Pool Experiment (E-SCAPE) is a 1/6th-scale thermal hydraulic model of the MYRRHA reactor and has been developed within the European framework programs THINS and MYRTE. MYRRHA is a multi-purpose hybrid pool-type research reactor cooled with lead-bismuth eutectic (LBE) that is under design at the Belgian nuclear research center SCK•CEN.The experiments from the E-SCAPE facility are essential for the design and licensing of MYRRHA. On the one hand, the experimental data from E-SCAPE directly provide information to the designers on the flow patterns and heat transport in the LBE in the upper and lower plena of the reactor. On the other hand, the experimental data from E-SCAPE enable qualification of thermal hydraulic computer codes used for analyzing the flow and heat transport in the MYRRHA reactor. The accuracy of codes rely on the underlying modelling approaches, whose validity can only be assessed by comparing them with relevant experimental data. Given the rare conditions present in heavy liquid cooled nuclear reactors, the possibility to compare complete pool simulations with experimental data is a unique opportunity.This paper presents the modelling approaches and first results of the pre-test analyses performed for E-SCAPE by NRG, SCK•CEN and VKI using Computational Fluid Dynamics (CFD). The CFD analyses of NRG are performed with the STAR-CCM + code, at SCK•CEN with the ANSYS CFX code and at VKI OpenFOAM is used. A first comparison of the results obtained with three different CFD codes for flow and heat transport in the E-SCAPE pool is presented and, where possible, validated against experimental results.

Integrated framework for model assessment and advanced uncertainty quantification of nuclear computer codes under Bayesian statistics

A framework for model evaluation and uncertainty quantification (UQ) is presented with applications oriented to nuclear engineering simulation codes. Our framework is inspired by the previous research on Bayesian statistics and model averaging. The methodology is demonstrated by performing UQ of three thermal-hydraulic computer codes used for two-phase flow simulation inside nuclear reactors, and conclusions regarding their performance are drawn. The complexity of the framework implementation depends upon the information to be drawn about the models as well as the nature of the models and the data. Uncertainties inherent in the input parameters and experimental data, along with predictive and model-form uncertainty can be quantified in this methodology. A composite (average) model based on the competent models can be created for improved response prediction. Two benchmarks featuring steady-state void fraction data in full-scale light water reactor (LWR) channels are used to demonstrate the methodology. The results show that the three models/codes demonstrate variable competitiveness in reproducing the data (i.e. goodness of fit with the data). There is no consistent trend at which each code excels. The predictive uncertainty (representing individual model deficiency or discrepancy) dominates the model-form uncertainty for many cases in this study due to two reasons: (1) presence of a single competent model for a specific response and (2) poor agreement with experimental data for certain responses at which nuclear codes struggle, such as low pressure and subcooled boiling conditions. In general, improvements in composite predictions (based on posterior model weights) are observed for BFBT data, while slight improvement is found for PSBT. For PSBT, the predictive uncertainty of RELAP5 and TRACE dominates the response uncertainty causing weak improvement. Additional efforts are needed to improve the closure models of these codes in future to reduce the model discrepancy contribution. This framework can be utilized for this purpose at which various closure models for the same code can be assessed in terms of their effect on the final response uncertainty. The proposed framework is flexible and extendable to other types of physics, models, and data. Developing the underlying methodology of calculating the model weights is the main focus in the subsequent studies.

Problems of Reliability Indicators Increase of Critical Heat Flux Calculations in the Water-Cooled Nuclear Reactors Based on the Computer Thermal-Hydraulic Codes

The analysis of the current state of research and developments in the field of creation of thermal-hydraulic computer codes has been performed. The experience of creation of foreign versions of best-estimate codes was analyzed. Considerable attention is paid to the issue of critical heat flux calculation of nuclear reactors channels. It is demonstrated that now the efficiency of application of modern computer codes for estimating of the heat transfer crisis in the water-cooled nuclear reactors requires further improvement. Calculation methods for accuracy increase of predicting this thermal-hydraulic phenomenon in reactor channels are considered. The well-known methods of critical thermal flux in nuclear reactors channels have been analyzed. Peculiarities of determination of the heat transfer crisis in the forced of the vapor-water steam motion have been reviewed. Adequacy of software computer codes designed to calculate the main safety parameters of water-cooled nuclear reactors was analyzed. The idea of the physical mechanism of the heat transfer crisis under forced motion of a two-phase flow in heated channels was considered. Particular attention has been paid to analysis of experimental and calculated data on conditions of initiation of a heat transfer crisis in fuel assemblies rods.

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.

Modelling of the spent fuel heat-up in the spent fuel pools using one-dimensional system codes and CFD codes

Abstract The main functions of spent fuel pools are to remove the residual heat from spent fuel assemblies and to perform the function of biological shielding. In the case of loss of heat removal from spent fuel pool, the fuel rods and pool water temperatures would increase continuously. After the saturated temperature is reached, due to evaporation of water the pool water level would drop, eventually causing the uncover of spent fuel assemblies, fuel overheating and fuel rods failure. This paper presents an analysis of loss of heat removal accident in spent fuel pool of BWR 4 and a comparison of two different modelling approaches. The one-dimensional system thermal-hydraulic computer code RELAP5 and CFD tool ANSYS Fluent were used for the analysis. The results are similar, but the local effects cannot be simulated using a one-dimensional code. The ANSYS Fluent calculation demonstrated that this three-dimensional treatment allows to avoid the need for many one-dimensional modelling assumptions in the pool modelling and enables to reduce the uncertainties associated with natural circulation flow calculation.

Preliminary validation of RELAP5/Mod4.0 code for LBE cooled NACIE facility

The one-dimensional thermal hydraulic computer code RELAP5 was developed for thermal hydraulic study of light water reactor as well as for nuclear research reactors. The purpose of this work is to evaluate the code RELAP5/Mod4.0 for analysis of research reactors. This paper consists of three major sections. The first section presents detailed discussions on thermo-physical properties of Lead Bismuth Eutectic (LBE) incorporated in RELAP5/Mod4.0 code. In the second section, benchmarking of RELAP5/Mod4.0 has been done with the Natural Circulation Experimental (NACIE) facility in comparison with Barone’s simulations using RELAP5/Mod3.3. Three different power levels (10.8kW, 21.7kW and 32.5kW) under natural circulation conditions are considered. Results obtained for LBE temperatures, temperature difference across heat section, pin surface temperatures, mass flow rates and heat transfer coefficients in heat section heat exchanger are in agreement with Barone’s simulation results within 7% of average relative error. Third section presents validation of RELAP5/Mod4.0 against the experimental data of NACIE facility performed by Tarantino et al. test number 21 at power of 22.5kW comparing the profiles of temperatures, mass flow rate and velocity of LBE. Simulation and experimental results agree within 7% of average relative error.

MARS-KS1.3을 이용한 피동원자로건물냉각계통 열수력 성능 예비분석

Abstract - A passive containment cooling system has been design ed to remove the heat inside a containment during accidents without external power supply. In this work, t he PCCS was introduced in the APR1400 plant to replace the containment spray system and, then, the thermal-hydraulic performance of the PCCS was analyzed using the system thermal-hydraulic computer code, MARS. A double-ended cold-leg break accident, which is known to induce the maximum pressure in the containment, is simulated, where the thermal hydraulics of the PCCS, the reactor coolant system, and the containment are simultaneously simulated. The results of the calculations showed that the PCCS can replace the existing spray system and that the containment building and its internal structure al so play a very important role for the heat removal during the accident. Some sensitivity calculations were carried out to evaluate the model uncertainty and the effects of design parameters. The limitations of the PC CS are also discussed.

Assessment of uncertainties of the models used in thermal–hydraulic computer codes

The article deals with matters concerned with the problem of determining the statistical characteristics of variable parameters (the variation range and distribution law) in analyzing the uncertainty and sensitivity of calculation results to uncertainty in input data. A comparative analysis of modern approaches to uncertainty in input data is presented. The need to develop an alternative method for estimating the uncertainty of model parameters used in thermal–hydraulic computer codes, in particular, in the closing correlations of the loop thermal hydraulics block, is shown. Such a method shall feature the minimal degree of subjectivism and must be based on objective quantitative assessment criteria. The method includes three sequential stages: selecting experimental data satisfying the specified criteria, identifying the key closing correlation using a sensitivity analysis, and carrying out case calculations followed by statistical processing of the results. By using the method, one can estimate the uncertainty range of a variable parameter and establish its distribution law in the above-mentioned range provided that the experimental information is sufficiently representative. Practical application of the method is demonstrated taking as an example the problem of estimating the uncertainty of a parameter appearing in the model describing transition to post-burnout heat transfer that is used in the thermal–hydraulic computer code KORSAR. The performed study revealed the need to narrow the previously established uncertainty range of this parameter and to replace the uniform distribution law in the above-mentioned range by the Gaussian distribution law. The proposed method can be applied to different thermal–hydraulic computer codes. In some cases, application of the method can make it possible to achieve a smaller degree of conservatism in the expert estimates of uncertainties pertinent to the model parameters used in computer codes.

Thermal Analysis for an Ultra High Temperature Gas-Cooled Reactor with Pebble Type Fuels

This study presents a predictive thermal-hydraulic analysis with packed spheres in a nuclear gas-cooled reactor core. The predictive analysis considering the effects of high power density and the some porosity value were applied as a design condition for an Ultra High Temperature Reactor (UHTR). The thermal-hydraulic computer code was developed and identified as PEBTEMP. The highest outlet coolant temperature of 1316 oC was achieved in the case of an UHTREX at LASL, which was a small scale UHTR using hollow-rod as a fuel element. In the present study, the fuel was changed to a pebble type, a porous media. Several calculation based on HTGR-GT300 through GT600 were 4.8 w/cm3 through 9.6 w/cm3, respectively. As a result, the relation between the fuel temperature and the power density was obtained under the different system pressure and coolant outlet temperature. Finally, available design conditions are selected.

노심보충수탱크의 직접접촉응축에 대한 MARS의 계산능력평가

This study aimed at assessing the analysis capability of thermal-hydraulic computer code, MARS for the behaviors of the core make-up tank (CMT). The sensitivity study on the nodalization to simulate the CMT was conducted, and the MARS calculations were compared with KAIST experimental data and RELAP5/MOD3.3 calculations. The 12-node model was fixed through a nodalization study to investigate the effect of the number of nodes in the CMT (2-, 4-, 8-, 12-, 16-node). The sensitivity studies on various parameters, such as water subcooling of the CMT, steam pressure, and natural circulation flow were done. MARS calculations were reasonable in the injection time and the effects of several parameters on the CMT behaviors even though the mesh-dependency should be properly treated for reactor applications.

The Concept and RELAP5 Model of Thermal-Hydraulic System, Employing a Rapid Condensation for Coolant Circulation

The rapid condensation event is mostly considered a dangerous and undesirable side effect in thermal-hydraulic systems. This work demonstrates a different viewpoint, where condensation implosion is employed to perform mechanical work. Previous experimental study of the condensation implosion event, briefly presented in this article, showed that condensation implosion can be induced intentionally. These results were used as the basis for further investigations. In this work, a concept of the thermal-hydraulic system has been developed, where condensation-implosions-generated pressure difference could be used as a driving force. Numerical study has been performed to investigate the operation of the developed conceptual thermal-hydraulic system. A thermal-hydraulic computer code RELAP5 was selected for modeling the system operation. The RELAP5 code was found not able to predict the condensation implosion; therefore, a modified heat transfer model was implemented into the code. This modification allowed simulating the condensation implosion artificially in the thermal-hydraulic system and modeling the system response to the event. Final results show that a proposed circulation principle is possible and such a thermal-hydraulic system can operate.

Thermal–hydraulic simulation of a radiant steam boiler using Relap5 computer code

Numerical simulations are nowadays increasingly used to assess the thermal–hydraulic behavior of a steam boiler instead of carrying out expensive tests in experimental facilities. In the present work, the Relap5/Mod3.2 thermal–hydraulic computer code, which is largely used in the nuclear engineering framework, is used to carry out a tube rupture transient of a high power industrial steam boiler. For this purpose a Relap5 model was developed and qualified at the steady-state conditions using the plant available data. The control and regulation systems are also modeled. The main outcome of this work is a successful application of a thermal–hydraulic system code in predicting the main key parameters of an industrial radiant steam boiler under steady and transient operating conditions.

Feasibility analysis of the modified ATHLET code for supercritical water cooled systems

Since the existing thermal-hydraulic computer codes for light water reactors are not applicable to supercritical water cooled reactors (SCWRs) owing to the limitation of physical models and numerical treatments, the development of a reliable thermal-hydraulic computer code is very important to design analysis and safety assessment of SCWRs. Based on earlier modification of ATHLET for SCWR, a general interface is implemented to the code, which serves as the platform for information exchange between ATHLET and the external independent physical modules. A heat transfer package containing five correlations for supercritical water is connected to the ATHLET code through the interface. The correlations are assessed with experimental data. To verify the modified ATHLET code, the Edwards–O’Brian blow-down test is simulated. As first validation at supercritical pressures, a simplified supercritical water cooled loop is modeled and its stability behavior is analyzed. Results are compared with that of the theoretical model and SASC code in the reference and show good agreement. To evaluate its feasibility, the modified ATHLET code is applied to a supercritical water cooled in-pile fuel qualification test loop. Loss of coolant accidents (LOCAs) due to break of coolant supply lines are calculated for the loop. Sensitivity analysis of some safety system parameters is performed to get further knowledge about their influence on the function of the safety system. All the results achieved indicate that the modified ATHLET code has good feasibility in simulation of supercritical water cooled systems.

C204 球状燃料を有する高温ガス炉の燃料温度解析に及ぼす熱出力の影響(OS8 軽水炉・新型炉・原子力安全)

This paper indicates the predictive simulation of thermal-hydraulic analysis with packed spheres in high temperature reactor core. The pressure drop experiments were carried out in the air through cylindrical test section. The porosity simulation is calculated in a computer code of the particle method. The thermal-hydraulic computer code was developed, and it was identified as PEBTEMP. The predictive analysis was applied to the design condition of thermal power profile in the axial direction.

Projection-based curve clustering

This paper focuses on unsupervised curve classification in the context of nuclear industry. At the Commissariat à l'Energie Atomique (CEA), Cadarache (France), the thermal-hydraulic computer code CATHARE is used to study the reliability of reactor vessels. The code inputs are physical parameters and the outputs are time evolution curves of a few other physical quantities. As the CATHARE code is quite complex and CPU time-consuming, it has to be approximated by a regression model. This regression process involves a clustering step. In the present paper, the CATHARE output curves are clustered using a k-means scheme, with a projection onto a lower dimensional space. We study the properties of the empirically optimal cluster centres found by the clustering method based on projections, compared with the ‘true’ ones. The choice of the projection basis is discussed, and an algorithm is implemented to select the best projection basis among a library of orthonormal bases. The approach is illustrated on a simulated example and then applied to the industrial problem.