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 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.     

Thermal–hydraulic studies related to the design of LBE neutron spallation target for accelerator driven systems

Recently, studies have been taken up in world's leading nuclear research institutes to develop accelerator driven systems (ADS). Our department has earlier proposed a one-way coupled fast-thermal reactor of 750 MW (thermal). This reactor requires current in the range of 1–2 mA for proton beam of 1 GeV. A suitable liquid metal lead bismuth-eutectic (LBE) target based on buoyancy as well as gas driven method has been designed for this reactor earlier. In this paper, detailed thermal analysis in the spallation and window region has been carried out to study the operability of the target from thermo-mechanical point of view. FLUKA and computational fluid dynamics (CFD) codes have been used for this analysis. The results indicate that, the temperatures, thermo-mechanical stresses are within the required values. The detailed analysis of this work is presented in this paper.     

Thermal hydraulic feasibility analysis of the IBED PHTS for ITER

One of the main challenges of the ITER fusion reactor is to effectively remove large amount of heat deposited to the surface of the plasma facing components. The tokamak cooling water system (TCWS) will accomplish the objective of removing about 1 GW of peak heat load from in-vessel components while maintaining pressures and temperatures of the coolant within acceptable and safe limits during different operational scenarios. A study of feasibility has been launched for the IBED PHTS (Integrated Blanket, Edge localized mode coils (ELMs) and Divertor Primary Heat Transfer System; it consists of five independent cooling trains (four operational and one in stand-by), one steam pressurizer, supply and return headers, ring manifolds and connections to the all in-vessel components (i.e. First Wall Blanket, Divertor, ELM, Diagnostics and other Ports clients).The dynamic behaviour of the IBED PHTS has been investigated by means of RELAP5^® code to simulate the response of the system during plasma pulse and baking operations. Due to the plasma heat deposition on the surfaces of the in-vessel components and subsequent increase in hot leg temperature, a large amount of water volume is transferred from the hot legs of the circuit to the surge-line of the pressurizer during each burn cycle. This causes rapid increase of pressure and temperature of the system and the following actions are proposed to counteract these variations: spray injection in the upper dome of the pressurizer from the Chemical and Volume Control System (CVCS) to reduce the pressure and active control of flow rates through heat exchangers and their bypass loops to regulate the heat transfer from the primary system to the environment via secondary and tertiary loops.This paper focuses on the prediction of the thermal hydraulic behaviour of the IBED PHTS during plasma pulses and baking scenarios, describing the various activity of the analysis, the geometrical assessment of the circuit and the modelling with RELAP5^® code. The results have been compared with design and operational requirement. Possible strategies to enhance the system performances have been formulated.     

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.     

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.     

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.     

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.     

Effects of wire-spacer shape in LMR on thermal–hydraulic performance

In this work, analyses of three-dimensional flow and convective heat transfer in wire-wrapped fuel assemblies with different shaped wire-spacers have been carried out using Reynolds-averaged Navier–Stokes equations with shear stress transport turbulence model. Three cross-sectional shapes of wire-spacer, circle, hexagon and rhombus have been tested. All the assemblies have been analyzed for single pitch of wire-spacer with periodic boundary conditions applied at inlet and outlet of the calculation domain. It is found that the assemblies exhibit the directional periodicity in radial gradients between the adjacent subchannels due to presence of wire-spacer. The overall pressure drop is highest in case of rhombus shaped wire-spacer assembly followed by hexagonal shaped. Although circular shaped wire-spacer gives lowest peak temperature as well as lowest overall temperature difference in the assembly, the rhombus shaped wire-spacer assembly gives highest Nusselt number on fuel rod surface.     

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.     

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.     

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.     

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.     

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 analysis code for the Pebble Bed Water-cooled Reactor

The Pebble Bed Water-cooled Reactor (PBWR) is a water-moderated water-cooled pebble bed reactor in which millions of tristructural-isotropic (TRISO) coated micro-fuel elements (MFE) pile in each assembly. Light water is used as coolant that flows from bottom to top in the assembly while the moderator water flows in the reverse direction out of the assembly.Steady-state thermal–hydraullic analysis code for the PBWR will provide a set of thermal hydraulic parameters of the primary loop so that heat transported out of the core can match with the heat generated by the core for a safe operation of the reactor. The key parameters of the core including the void fraction, pressure drop, heat transfer coefficients, the temperature distribution and the Departure from Nucleate Boiling Ratio (DNBR) is calculated for the core in normal operation. The code can calculate for liquid region, water-steam two phase region and superheated steam region. The results show that the maximum fuel temperature is much lower than the design limitation and the flow distribution can meet the cooling requirement in the reactor core. As a new type of nuclear reactor, the main design features with a sufficient safety margin were also put forward in this paper.     

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.     

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.