Core thermal–hydraulic design

The core thermal–hydraulic design for the HTTR is carried out to evaluate the maximum fuel temperature at normal operation and anticipated operation occurrences. To evaluate coolant flow distribution and maximum fuel temperature, we use the experimental results such as heat transfer coefficient, pressure loss coefficient obtained by mock-up test facilities. Furthermore, we evaluated hot spot factors of fuel temperatures conservatively.As the results of the core thermal–hydraulic design, an effective coolant flow through the core of 88% of the total flow, is achieved at minimum. The maximum fuel temperature appears during the high-temperature test operation, and reaches 1492 °C for the maximum through the burn-up cycle, which satisfies the design limit of 1495 °C at normal operation. It is also confirmed that the maximum fuel temperature at any anticipated operation occurrences does not exceed the fuel design limit of 1600 °C in the safety analysis.On the other hand, result of re-evaluation of analysis condition and hot spot factors based on operation data of the HTTR, the maximum fuel temperature for 160 effective full power operation days is estimated to be 1463 °C. It is confirmed that the core thermal–hydraulic design gives conservative results.     

Local thermal–hydraulic behaviour in tight 7-rod bundles

Advanced water-cooled reactor concepts with tight lattices have been proposed worldwide to improve the fuel utilization and the economic competitiveness. In the present work, experimental investigations were performed on thermal–hydraulic behaviour in tight hexagonal 7-rod bundles under both single-phase and two-phase conditions. Freon-12 was used as working fluid due to its convenient operating parameters. Tests were carried out under both single-phase and two-phase flow conditions. Rod surface temperatures are measured at a fixed axial elevation and in various circumferential positions. Test data with different radial power distributions are analyzed. Measured surface temperatures of unheated rods are used for the assessment of and comparison with numerical codes.In addition, numerical simulation using sub-channel analysis code MATRA and the computational fluid dynamics (CFD) code ANSYS-10 is carried out to understand the experimental data and to assess the validity of these codes in the prediction of flow and heat transfer behaviour in tight rod bundle geometries. Numerical results are compared with experimental data. A good agreement between the measured temperatures on the unheated rod surface and the CFD calculation is obtained. Both sub-channel analysis and CFD calculation indicates that the turbulent mixing in the tight rod bundle is significantly stronger than that computed with a well established correlation.     

Transient thermal–hydraulic simulations of direct cycle gas cooled reactors

This work concerns the design and safety analysis of gas cooled reactors. The CATHARE code is used to test the design and safety of two different concepts, a High Temperature Gas Reactor concept (HTGR) and a Gas Fast Reactor concept (GFR). Relative to the HTGR concept, three transient simulations are performed and described in this paper: loss of electrical load without turbo-machine trip, 10 in. cold duct break, 10 in. break in cold duct combined with a tube rupture of a cooling exchanger. A second step consists in modelling a GFR concept. A nominal steady state situation at a power of 600 MW is obtained and first transient simulations are carried out to study decay heat removal situations after primary loop depressurisation. The turbo-machine contribution is discussed and can offer a help or an alternative to “active” heat extraction systems.     

CFD analysis of thermal–hydraulic behavior of heavy liquid metals in sub-channels

CFD analysis was carried out for thermal–hydraulic behavior of heavy liquid metal flows, especially lead–bismuth eutectic, in sub-channels of both triangular and square lattices. Effect of various parameters, e.g. turbulence models and pitch-to-diameter ratio, on the thermal–hydraulic behavior was investigated. Among the turbulence models selected, only the second order closure turbulence models reproduce the secondary flow. For the entire parameter range studied in this paper, the amplitude of the secondary flow is less than 1% of the mean flow. A strong anisotropic behavior of turbulence is observed. The turbulence behavior is similar in both triangular and square lattices. The average amplitude of the turbulent velocity fluctuation across the gap is about half of the shear velocity. It is only weakly dependent on Reynolds number and pitch-to-diameter ratio. A strong circumferential non-uniformity of heat transfer is observed in tight rod bundles, especially in square lattices. Related to the overall average Nusselt number, CFD codes give similar results for both triangular and square rod bundles. Comparison of the CFD results with bundle test data in mercury indicates that the turbulent Prandtl number for HLM flows in rod bundles is close to 1.0 at high Peclet number conditions, and increases by decreasing Peclet number. Based on the present results, the SSG Reynolds stress model with semi-fine mesh structures is recommended for the application of HLM flows in rod bundle geometries.     

Treatment of the thermal–hydraulic uncertainties in the pressurized thermal shock analysis

The pressurized thermal shock (PTS) analysis is a quantitative analysis to calculate the vessel failure probability of the embrittled reactor pressure vessel. The PTS analysis consists of three major parts, such as the probabilistic safety analysis (PSA), the thermal–hydraulic analysis (T/H), and the probabilistic fracture mechanics (PFM) analysis. Because each analysis involves many parameters and assumptions associated with the uncertainties, it is important to identify and incorporate them into the analysis. Though the PSA and PFM analysis can be easily treated statistically, the thermal–hydraulic analysis results are very difficult to be treated statistically. Instead, sensitivity analyses of the thermal–hydraulic inputs were performed to understand the significance of the variation in the thermal–hydraulic inputs to the PFM analysis. In this study, the existing PFM code was modified to incorporate the uncertainties in the thermal–hydraulic inputs for the PFM analysis. The effects of the uncertainties in the thermal–hydraulic inputs for the vessel failure probabilities were evaluated using the modified code. The results showed the effects of uncertainties in the thermal–hydraulic inputs on the vessel failure probabilities are not significant for the ranges of the transient types. Even for the larger uncertainties, the effects on the vessel failure probabilities are small. Also, the effects of the thermal–hydraulic uncertainties vary depending on the transient characteristics such that the effects are greatest for the pressure dominant transient. Within the transient, the relative increases in the failure probabilities are greatest for the circumferentially oriented semi-elliptical flaws. It was found that the results of the sensitivity analysis using one standard deviation are conservative enough to bound the analysis results considering the uncertainties in the thermal–hydraulic inputs.     

Quantitative assessment of thermal–hydraulic codes used for heavy water reactor calculations

The RD-14M large LOCA test, characterized by a reliable set of experimental data, was selected for an international standard problem exercise (SPE) entitled “Intercomparison and validation of computer codes for thermal–hydraulics safety analyses”. The activity was performed within the frame of International Atomic Energy Agency's (IAEAs) Technical Working Group on Advanced Technologies for Heavy Water Reactors (TWG-HWR). In this study, the recently improved fast Fourier transform based method (FFTBM) was used for accuracy quantification of RD-14M large LOCA test B9401 calculations of six participants using four different thermal–hydraulic codes. In addition, developing the capability to calculate the accuracy as a function of time-continues-valued accuracy, did further improvement of FFTBM. Namely, in the past only single valued accuracy parameters for selected time windows and time intervals were calculated. The objective of the study was to demonstrate that the new FFTBM is a powerful tool for quantitative assessment of thermal–hydraulic codes. For demonstration, the test from the facility simulating heavy water reactor was used. The blind accuracy analysis was completed based on solely experimental and calculated data. However, short discussions were held with the representative from Italy (co-author, here) regarding phenomenological windows, variables and void fraction weights selection. In general, the open accuracy analysis confirmed the results obtained in blind accuracy analysis. The main conclusions from accuracy analysis agree with the conclusions from the SPE intercomparison report, which was written independently. Finally, the results suggest that the accuracy of the best calculations of the RD-14M test is comparable with the best calculations of light water reactor experiments.     

Development of a thermal–hydraulic system code for simulators based on RELAP5 code

A thermal–hydraulic system code for simulators, RELAPSIM, was developed at NSSE based on RELAP5. The development procedure consists of three major parts. Firstly, time control function was added into the code to meet real-time calculation needs. Secondly, controlled dynamic data communication was improved, so that thermal–hydraulic parameters can be easily modified for further applications. Finally, functions controlling the computation procedure were embedded to achieve a full capability to simulate multiple operations, such as start-up, shutting down or freeze. This paper describes the main features of the new code. The results of code assessment and code application are presented and discussed.     

Transient analysis of Ghana Research Reactor-1 using PARET/ANL thermal–hydraulic code

PARET/ANL (Version 7.3 of 2007) thermal–hydraulic code was used to perform transient analysis of the Ghana Research Reactor-1. The reactivities inserted were 2.1 mk, 4 mk and 6.71 mk. The results obtained are similar to experiment and theoretical studies performed to demonstrate that the reactor is safe to operate. The PARET/ANL (Version 7.3 of 2007) could not simulate the reactivities above 5 mk insertions which were successfully performed in earlier theoretical and experimental studies. This may be attributed to different fluid flow and heat transfer regimes within the flow channels of the reactor that were considered by the codes.     

Analysis of potential waterhammer at the Ignalina NPP using thermal–hydraulic and structural analysis codes

The RBMK (Russian acronym for ‘channeled large power reactor’)-1500 reactors at the Ignalina nuclear power plant (NPP) have a series of check valves in the main circulation circuit (MCC) that serve the coolant distribution in the fuel channels. In the case of a hypothetical guillotine break of pipelines upstream of the group distribution headers (GDH), the check valves and adjusted piping integrity is a key issue for the reactor safety during the rapid closure of check valve. An analysis of the waterhammer effect (i.e. the pressure pulse generated by the valves slamming closed) is needed. The thermal–hydraulic and structural analysis of waterhammer effects following the guillotine break of pipelines at the Ignalina NPP with RBMK-1500 reactors was conducted by employing the RELAP5 and PipePlus codes. Results of the analysis demonstrated that the maximum values of the pressure pulses generated by the check valve closure following the hypothetical accidents remain far below the value of pressure of the hydraulic tests, which are performed at the NPP and the risk of failure of the check valves or associated pipelines is low. Sensitivity analysis of pressure pulse dependencies on calculation time step and check valve closure time was performed. Results of RELAP5 calculations are benchmarked against waterhammer transient data obtained by employing structural mechanics code BOS fluids.     

Uncertainty and sensitivity analyses of the Kozloduy pump trip test using coupled thermal–hydraulic 3D kinetics code

The modeling of complex transients in nuclear power plants (NPP) remains a challenging topic for best estimate three-dimensional coupled code computational tools. This technique is, nowadays, extensively used since it allows decreasing conservatism in the calculation models and performs more realistic simulation and more precise consideration of multidimensional effects under complex transients in NPPs. Therefore, large international activities are in progress aiming to assess the capabilities of coupled codes and the new frontiers for the nuclear technology that could be opened by this technique. In the current paper, a contribution to the assessment and validation of coupled code technique through the Kozloduy VVER100 pump trip test is performed. For this purpose, the coupled RELAP5/3.3-PARCS/2.6 code is used. The code results were assessed against experimental data. Deviations between code predictions and measurements are mainly due to the used models for evaluating and modeling of the Doppler feedback effect. Further investigations through the use of two “antagonist” uncertainty GRS and the CIAU methods, were considered in order to evaluate and quantify the origin of the observed discrepancies. It was revealed on one hand that relative error quantification discrepancies exist between the two approaches, and further enhancements for both methods are needed.     

CAST3M/ARCTURUS: A coupled heat transfer CFD code for thermal–hydraulic analyzes of gas cooled reactors

The safety of gas cooled reactors (High Temperature Reactors (HTR), Very High Temperature Reactors (VHTR) or Gas Cooled Fast Reactors (GFR)) must be ensured by systems (active or passive) which maintain loads on component (fuel) and structures (vessel, containment) within acceptable limits under accidental conditions. To achieve this objective, thermal–hydraulics computer codes are necessary tools to design, enhance the performance and ensure a high safety level of the different reactors. Some key safety questions are related to the evaluation of decay heat removal and containment pressure and thermal loads. This requires accurate simulations of conduction, convection, thermal radiation transfers and energy storage. Coupling with neutronics is also an important modeling aspect for the determination of representative parameters such as neutronics coefficient (Doppler coefficient, Moderator coeffcient, …), critical position of control rods, reactivity insertion aspects, …. For GFR, the high power density of the core and its necessary reduced dimension cannot rely only on passive systems for decay heat removal. Therefore, forced convection using active safety systems (gas blowers, heat exchangers, …) are highly recommended. Nevertheless, in case of station black-out, the safety demonstration of the concept should be guaranteed by natural circulation heat removal. This could be performed by keeping a relatively high back-up pressure for pure helium convection and also by heavy gas injection. So, it is also necessary to model mixing of different gases, the on-set of natural convection and the pressure and thermal loads onto the proximate or guard containment. In this paper, we report on the developments of the CAST3M/ARCTURUS thermal–hydraulics (Lumped Parameter and CFD) code developed at CEA, including its coupling to the neutronics code CRONOS2 and the system code CATHARE. Elementary validation cases are detailed, as well as application of the code to benchmark problems such as the HTR-10 thermal–hydraulic exercise. Examples of containment thermal–hydraulics calculations for fast reactor design (GFR) are also detailed.     

International standard problem on containment thermal–hydraulics ISP47: Step 1—Results from the MISTRA exercise

The understanding of hydrogen distribution during severe accidents in a nuclear reactor containment is still an open issue. Several containment thermal–hydraulics international standard problems (ISP) have been conducted to address this topic. However, the predictions made by the available lumped parameter or CFD computer codes were generally not satisfactory. Therefore, a new exercise was launched in 1999 using new state-of-the-art experimental facilities TOSQAN, MISTRA and ThAI that included sophisticated 3D instrumentation and well-controlled boundary conditions. Predictive capabilities of important and still uncertain phenomena such as wall condensation, natural circulation and gas stratification are assessed. In addition, comparison between lumped parameter (LP) and CFD codes and assessment of the capability of CFD codes to deal with scaling effects are performed. This article reports on the part of the exercise which concerns the MISTRA facility including experimental results and blind benchmark exercises.     

Thermal–hydraulic analysis for a sodium-heated steam generator using a multi-shell method

A multi-shell analysis method has been applied to predict the Thermal–hydraulics in a steam generator of a liquid metal reactor. The method is intended to improve the calculation accuracy for temperature profiles at a wide range of heat exchange rates in a large sized steam generator for the future plant. The calculation accuracy has been examined using experimental 50 MWth steam generator test data that were measured during 1970's–1980's. The calculation results of temperature profiles were compared with experimental data at 20, 30, 50, 75, 90, and 100% of nominal operating condition. Responses for the stepwise flow rate variations were also evaluated. In conclusion, the calculation accuracy for the temperature profile in the steam generator was improved by using the multi-shell analysis method for a wide range of heat exchange rates.     

Thermal hydraulic study on a high-temperature gas–gas heat exchanger with helically coiled tube bundles

An experimental study was carried out to investigate flow-induced vibration, heat transfer and pressure drop of helically coiled tubes of an intermediate heat exchanger (IHX) for the HTTR, using a full-size partial model and air as the fluid. The test model has 54 helically coiled tubes separated into three layer bundles, surrounding the center pipe. The vibration of the tube bundles was mainly at the center pipe, and the individual vibrations of the tube bundles were not significant under the operation conditions of the IHX. The heat transfer of the tube outside, due to forced convection, was obtained as a function of Re^0.51Pr^0.3, and the friction factor, depending on the tube arrangement, was proportional to Re^−0.14.     

The inherent safety characteristics of an unconventional LMR design—The Trench Reactor

This paper presents the transient behavior during off-normal operation of an unconventional liquid metal reactor design, called the Trench Reactor. Under the postulated accident conditions, this reactor design responds in an inherently safe manner to loss of heat sink accidents, loss of flow accidents, overcooling accidents and transient overpower accidents with 25 cents of reactivity insertion. The characteristics that cause such inherently save behavior are the properties of the materials and the configuration of the reactor primary system, even without any activated safety devices.     

Effects of transverse power distribution on thermal hydraulic analysis

To investigate the effects of transverse power distribution on fuel temperature, a two-dimensional thermal analysis model was developed in this study. An equilibrium reactor core with 22 fuel assemblies facilitated with plate-type fuel was modeled using Monte Carlo N-Particle (MCNP) code, and the fuel assembly that released the largest amount of power was obtained. The fuel plates were divided into 4 or 12 vertical stripes within the fuel width in order to obtain the transverse power distributions. With 4 stripes in the fuel, the highest power peaking was 2.36, whereas the highest power peaking was 2.70 with 12 stripes in the fuel. A 6th order polynomial was generated to predict the local power peaking at the edge of the fuel. Using this 6th order polynomial, the maximum power peaking at the edge of the fuel was 3.06. As per transverse power distributions, the temperature at the edge of the fuel should have been higher with a higher power peaking. However, the maximum temperature in the fuel decreased with a power peaking higher than 2.65. This was because the high power locally released from the edge of the fuel was immediately dissipated to the cladding by lateral heat conduction. As with the maximum temperature, the heat flux also overshot and converged at a certain value. This showed that the fuel did not need to be divided into more than 18 vertical stripes within the fuel width in order to obtain the local power peaking from nuclear physics calculations.     

Thermal–hydraulic evaluation of the RBMK-1500 accident confinement system using CONTAIN 11AF

The response of the RBMK Accident Confinement System to a large break LOCA, medium break LOCA and small break LOCA is analyzed using the CONTAIN 11AF code. The effect of Condenser Tray Cooling System failure is investigated for the large break LOCA case. The analysis employs a best estimate mass/energy source and considers both short and long-term responses of the Accident Confinement System. Parametric studies are performed to evaluate the effects of water deposition on the short-term pressure peak and of by-pass leakage on long-term pressure increases.     

Numerical study on thermal hydraulic performance of a Printed Circuit Heat Exchanger

Very High Temperature Reactors (VHTRs) require a high-temperature and high integrity Intermediate Heat Exchanger (IHX) with high effectiveness to efficiently transfer the core thermal output to a secondary fluid for electricity generation, hydrogen production, and/or other industrial process heat applications. A class of compact plate-type heat exchanger, namely, Printed Circuit Heat Exchanger (PCHE), is one of the leading candidate IHX configuration being considered for VHTR applications. In the current study, simplified computational models of PCHE are investigated using Fluent™ software. The geometry of the models considered in the study replicate the PCHEs that were fabricated using Alloy 617 plates for use in a High-Temperature Helium Facility (HTHF) at The Ohio State University. The computational cases investigated are based on the design conditions of the HTHF, i.e., a maximum operating pressure of 3 MPa, hot and cold side inlet temperatures of 1173 K and 813 K, respectively, and mass flow rates varying from 10 to 80 kg/h. This range of mass flow rates correspond to laminar and laminar-to-turbulent transition flows in the PCHE flow channel passages. The laminar-to-turbulent transition behavior has been numerically investigated for the semicircular and circular channel geometries. The numerical study showed that the transition is observed at Reynolds numbers of 2300 and 3100 for the circular and semicircular channels, respectively. Heat transfer and pressure drop characteristics are evaluated to provide preliminary performance data for the PCHEs fabricated at operating temperatures similar to those of the VHTRs. Local convective heat transfer coefficients are calculated for the hot and cold sides and compared with the available correlations for the circular and semicircular ducts. Overall performance characteristics of the PCHE computational model are computed and described in terms of the thermal effectiveness, number of transfer units, and overall heat transfer coefficient.     

Neutronic performance of new coolants in a fusion–fission (hybrid) reactor

Selection of coolant used in the fuel zone of a fusion–fission (hybrid) reactor affects the neutronic performance of the blanket much. Recently, two coolants namely, Flinabe and Li₂₀Sn₈₀ have been investigated to use in fusion reactors as tritium breeder and energy carrier due to their advantages of low melting point, low vapor pressure. In this study, neutronic performance of these coolants in a hybrid reactor using Canada Deuterium Uranium Reactor (CANDU) spent fuel was investigated for an operation period of 48 months. And also that of natural lithium and Flibe was also examined for comparison. Neutron transport calculations were conducted on a simple experimental hybrid blanket in a cylindrical geometry with the help of the SCALE4.3 system by solving the Boltzmann transport equation with the XSDRNPM code in 238 neutron groups and a S₈–P₃ approximation.