Design and experiment of hot gas duct for the HTR-10

A horizontal coaxial double-tube hot gas duct is a key component connecting the reactor pressure vessel and the steam generator pressure vessel for the 10 MW High Temperature Gas-cooled Reactor—Test Module. Hot helium gas from the core outlet flows into the steam generator through the liner tube, while helium gas after being cooled returns to the core through a passage formed between the inner tube and the duct pressure vessel. Thermal insulation material is packed into the space between the liner tube and the inner tube to resist heat transfer from the hot helium to the cold helium. The thermal compensation structure is designed in order to avoid large thermal stress because of different thermal expansions of the duct parts under various conditions. According to the design principal of the hot gas duct, the detailed structure design and strength evaluation for it has been done. A full-scale duct test section was then made according to the design parameters, and its thermal performance experiment was carried out in a helium test loop. With helium gas at pressure of about 3.0 MPa and a temperature over 900 °C, the continuous operation time for the duct test section lasted 98 h. At a helium gas temperature over 700 °C, the cumulative operation time for the duct test section reached 350 h. The duct test section also experienced 20 pressure cycles in the pressure range of 0.1–3.4 MPa, 18 temperature cycles in the temperature range of 100–950 °C. Thermal test results show an effective thermal conductivity of the hot gas duct thermal insulation is 0.47 W m⁻¹ °C⁻¹ under normal operation condition. In addition, a hot gas duct depressurization test was carried out; the test result showed that the pressure variation occurred on the liner tube was not more than 0.2 MPa for an assumed maximum gas release rate.     

An evaluation of reactor cooling and coupled hydrogen production processes using the modular helium reactor

The high-temperature characteristics of the modular helium reactor (MHR) make it a strong candidate for producing hydrogen using either thermochemical or high-temperature electrolysis (HTE) processes. Using heat from the MHR to drive a sulfur-iodine (SI) thermochemical hydrogen production process has been the subject of a U.S. Department of Energy sponsored Nuclear Engineering Research Initiative (NERI) project led by General Atomics, with participation from the Idaho National Laboratory (INL) and Texas A&M University. While the focus of much of the initial work was on the SI thermochemical production of hydrogen, recent activities included development of a preconceptual design for an integral HTE hydrogen production plant driven by the process heat and electricity produced by a 600 MW MHR.This paper describes ATHENA analyses performed to evaluate alternative primary system cooling configurations for the MHR to minimize peak reactor vessel and core temperatures while achieving core helium outlet temperatures in the range of 900–1000 °C that are needed for the efficient production of hydrogen using either the SI or HTE process. The cooling schemes investigated are intended to ensure peak fuel temperatures do not exceed specified limits under normal or transient upset conditions, and that reactor vessel temperatures do not exceed American Society of Mechanical Engineers (ASME) code limits for steady-state or transient conditions using standard light water reactor vessel materials. Preconceptual designs for SI and HTE hydrogen production plants driven by one or more 600 MW MHRs at helium outlet temperatures in the range of 900–1000 °C are described and compared. An initial SAPHIRE model to evaluate the reliability, maintainability, and availability of the SI hydrogen production plant is also described. Finally, a preliminary flowsheet for a conceptual design of an HTE hydrogen production plant coupled to a 600 MW modular helium reactor is presented and discussed.     

Development of control technology for HTTR hydrogen production system with mock-up test facility: System controllability test for loss of chemical reaction

The Japan Atomic Energy Agency has been planning the demonstration test of hydrogen production with the High Temperature Engineering Test Reactor (HTTR). In a HTTR hydrogen production system (HTTR-H2), it is required to control a primary helium temperature within an allowable value at a reactor inlet to prevent a reactor scram. A cooling system for a secondary helium with a steam generator (SG) and a radiator is installed at the downstream of a chemical rector in a secondary helium loop in order to mitigate the thermal disturbance caused by the hydrogen production system. Prior to HTTR-H2, the simulation test with a mock-up test facility has been carried out to establish the controllability on the helium temperature using the cooling system against the loss of chemical reaction. It was confirmed that the fluctuations of the helium temperature at chemical reactor outlet, more than 200 K, at the loss of chemical reaction could be successfully mitigated within the target of ±10 K at SG outlet. A dynamic simulation code of the cooling system for HTTR-H2 was verified with the obtained test data.     

CFD investigations of helical wire-wrap fuel pin bundle and its comparison with straight wire bundle

Computational Fluid Dynamics (CFD) investigations of a fast reactor fuel pin bundle wrapped with helical and straight spacer wires have been carried out and the advantages of using helical spacer wire have been assessed. The flow and temperature distributions in the fuel pin bundle are obtained by solving the statistically averaged 3-Dimensional conservation equations of mass, momentum and energy along with high Reynolds number k-ε turbulence model using a customized CFD code CFDEXPERT. It is seen that due to the helical wire-wrap spacer, the coolant sodium not only flows in axial direction in the fuel pin bundle but also in a transverse direction. This transverse flow enhances mixing of coolant among the sub channels and due to this, the friction factor and heat transfer coefficient of the coolant increase. Estimation of friction factor, Nusselt number, sodium temperature uniformity at the outlet of the bundle and clad hot spot factor which are measures of the extent of coolant mixing and non-homogeneity in heat transfer coefficient around fuel pin are paid critical attention. It is seen that the friction factor and Nusselt number are higher (by 25% and 15% respectively) for the helical wire wrap pin bundle compared to straight wire bundle. It is seen that for 217 fuel pin bundle the maximum clad temperature is 750 K for straight wire case and the same for helical wire is 720 K due to the presence of transverse flow. The maximum temperature occurs at the location of the gap between pin and wire. The ΔT between the bulk sodium in the central sub-channel and peripheral sub-channel is 30 K for straight wire and the same for helical wire is 18 K due to the presence of secondary transverse flow which makes the outlet temperature more uniform. The hotspot factor and the hot channel factors predicted by CFD simulation are 10% lower than that used in conventional safety analysis indicating the conservatism in the safety analysis.     

Thermal–hydraulic feasibility analysis on uprating the HTR–PM

In order to meet energy demand in China, the high temperature gas-cooled reactor–pebble-bed module (HTR–PM) is being developed. It adopts a two-zone core, in which graphite balls are loaded in the central zone and the outer part is fuel ball zone, and couple with a steam cycle. Outer diameter of the reactor core is 4.0 m and height of the core is 9.43 m. The helium inlet and outlet temperature are 250 and 750 °C, respectively. The reactor thermal power is 380 MW. Preliminary studies show that the HTR–PM is feasible technologically and economically. In order to increase the reactor thermal power of the HTR–PM, some efforts have been made. These include increasing the height of reactor core, optimizing the thickness of fuel zone and better selection of the scheme of central graphite zone, etc. Basic design concepts and thermal–hydraulic parameters of the HTR–PM are given. Measures to increase the thermal power are introduced. Thermal–hydraulic analysis results are presented. The results show that, from the viewpoint of thermal–hydraulics, it is possible to increase the reactor power.     

Experimental investigation on heat transfer from a heated rod with a helically wrapped wire inside a square vertical channel to water at supercritical pressures

A supercritical water heat transfer test section has been built at Xi’an Jiaotong University to study the heat transfer from a 10 mm rod inside a square vertical channel with a wire-wrapped helically around it as a spacer. The test section is 1.5 m long and the wire pitch 200 mm. Experimental conditions included pressures of 23–25 MPa, mass fluxes of 500–1200 kg/m² s, heat fluxes of 200–800 kW/m², and inlet temperatures of 300–400 °C. Wall temperatures were measured with thermocouples at various positions near the rod surface. The experimental Nusselt numbers were compared with those calculated by empirical correlations for smooth tubes. The Jackson correlation showed better agreement with the test data compared with the Dittus-Boelter correlation but overpredicted the Nusselt numbers almost within the whole range of experimental conditions. Both correlations cannot predict the heat transfer accurately when deterioration occurred at low mass flux and relatively high heat flux in the pseudocritical region. Comparison of experimental data at two different supercritical pressures showed that the heat transfer was more enhanced at the lower supercritical pressure but the deterioration was more likely to occur at the higher pressure, meaning increased safety. Based on a comparison with an identical channel without the helical wrapped wire, it was found that the wire spacer does not enhance the heat transfer significantly under normal heat transfer conditions, but it contributes to the improvement of the heat transfer in the pseudocritical region and to a downstream shift of the onset of the deterioration. The Jackson buoyancy criterion is found to be valid and works well in predicting the onset of heat transfer deterioration occurring in the experiments without wire.     

Simulation tests for temperature mixing in a core bottom model of the HTR-Module

Interatom and Siemens are developing a helium-cooled Modular High Temperature Reactor. Under nominal operating conditions temperature differences of up to 120°C will occur in the 700°C hot helium flow leaving the core. In addition, cold gas leakages into the hot gas header can produce even higher temperature differences in the coolant flow. At the outlet of the reactor only a very low temperature difference of maximum ±15°C is allowed in order to avoid damages at the heat exchanging components due to alternating thermal loads. Since it is not possible to calculate the complex flow behaviour, experimental investigations of the temperature mixing in the core bottom had to be carried out in order to guarantee the necessary reduction of temperature differences in the helium. The presented air simulation tests in a 1:2.9 scaled plexiglass model of the core bottom showed an extremely high mixing rate of the hot gas header and the hot gas duct of the reactor. The temperature mixing of the simulated coolant flow as well as the leakage flows was larger than 95%. Transfered to reactor conditions this means a temperature difference of only ±3°C for the main flow at a quite reasonable pressure drop. For the cold gas leakages temperature differences in the hot gas up to 400°C proved to be permissible. The results of the simulation experiments in the Aerodynamic Test Facility of Interatom permitted to design a shorter bottom reflector of the core.     

Experimental and analytical studies on turbulent heat transfer performance of a fuel rod with spacer ribs for high temperature gas-cooled reactors

Turbulent heat transfer performance of a fuel rod with three-dimensional trapezoidal spacer ribs for high temperature gas-cooled reactors was studied for various Reynolds numbers using an annular channel at the same coolant condition as the reactor operation, maximum outlet temperature of 1000 °C and pressure of 4 MPa, and analytically by a numerical simulation using the k- turbulence model. The turbulent heat transfer coefficients of the fuel rod were 18–80% higher than those of a concentric smooth annulus at a region of Reynolds number exceeding 2000. On the other hand, the predicted average Nusselt number of the fuel rod agreed well with the empirical correlation obtained from the experimental data within a relative error of 10% with Reynolds number of more than 5000. It was verified that the numerical analysis results had sufficient accuracy. Furthermore, the numerical prediction could clarify quantitatively the effects of the heat transfer augmentation by the spacer ribs and the axial velocity increase due to a reduction in the annular channel cross-section.     

Effects of an intermediate heat treatment during a cold rolling on the tensile strength of a 9Cr–2W steel

The effects of an intermediate heat treatment during a cold rolling on the tensile strength of a 9Cr–2W steel were evaluated. Before a cold rolling, the steel was normalized at 1050 °C and tempered at 550 °C in order to avoid the formation of M₂₃C₆ and V-rich MX precipitates in the martensitic structure. A 75% cold rolling and a heat treatment at 750 °C for 30 min induced the formation of large M₂₃C₆ carbides in a fully recrystallized structure. However, three cold rollings with an intermediate heat treatment at 750 °C for 10 min after each cold rolling led to the formation of fine and uniform M₂₃C₆ carbides in a partially recrystallized structure, providing an enhanced tensile strength at 650 °C. It is thus concluded that an intermediate heat treatment during a cold rolling could be an effective procedure for fabricating a high strength 9Cr–2W steel at high temperatures.     

Local heat transfer and hot-spot factors in wire-wrap tube bundle

Heat transfer coefficients and hot-spot factors have been determined from measured local temperatures and calculated local mass flux in seven adjacent tubes and associated subchannels of a 61 wire-wrap tube bundle characteristic of the blanket of a GCFR (Gas Cooled Fast Reactor). The bundle consisted of 2.11 cm OD stainless steel tubes on a triangular array with a pitch/diameter ratio of P/D = 1.05. The helical wire of 0.1067 cm in diameter was coiled on the tube with a respective initial orientation of 0–120–240°C and 30.48 cm helical pitch. The experiment used water at atmospheric pressure and temperature as coolant. The resulting dimensionless correlation for heat transfer is applicable to gases and all non-metal fluids in one phase flow when the fluid properties at subchannel bulk temperature are used. This correlation is based on local subchannel mass flux and is applicable to all wire-wrap configurations. Local subchannel mass fluxes were determined with a computer program COBRA IV and used to correlate the average Nusselt number for each subchannel in terms of local Reynolds number and fluid Prandtl number. The differences of up to 19% between that correlation and the one presented in earlier work are discussed in the text. The hot-spot factors on the convective heat transfer coefficient for tubes and subchannels are given as a function of Reynolds number based on a bundle average mass flux and a local subchannel hydraulic diameter. These factors are specific to the bundle configuration and are also dependent on the wire-wrap configuration.     

Design and fabrication of a high-temperature helium regenerator

Refractory metallic foams can increase heat transfer efficiency in gas-to-gas and liquid metal-to-gas heat exchangers by providing an extended surface area for better convection, i.e. conduction into the foam ligaments providing a “fin-effect,” and by disruption of the thermal boundary layer near the hot wall and ligaments by turbulence promotion. In this article, we describe the design of a high-temperature refractory regenerator (closed-loop recuperator) using computational fluid dynamics (CFD) modeling of actual foam geometries obtained through computerized micro-tomography. The article outlines the design procedure from geometry import through meshing and thermo-mechanical analysis and discusses the challenges of fabrication using pure molybdenum and TZM. The foam core regenerator is more easily fabricated, less expensive and performs better than refractory flat plate-type heat exchangers. The regenerator can operate with a maximum hot leg inlet temperature of 900 °C and transfer 180 kW to the cold leg using 100 g/s helium at 4 MPa. Future high heat flux experiments on helium-cooled plasma facing components will utilize the high temperature and high pressure capabilities of this unique regenerator. Similar components will be required to adapt fusion power reactors to high-efficiency Brayton power conversion systems and enable operation of advanced divertor and blanket systems.     

Conceptual design of a helium heater for high temperature applications

The Karlsruhe Advanced Technologies Helium Loop (KATHELO) has been designed for testing divertor modules as well as qualifying materials for high heat flux, high temperature (up to 800 °C) and high pressure (10 MPa) applications. The test section inlet temperature level is controlled using a process electrical heater. To cope with the extreme operating conditions, a special design of this unit has been proposed. In this paper the conceptual design of the unit will be presented and the impact of the coupling between the cold and hot helium gas on the overall efficiency of the loop will be investigated. The detailed thermal-hydraulic analysis of the feed through of the hot helium into the low temperature pressure vessel using ANSYS CFX will be presented. The impact of the design choices on the overall energy budget of the loop will be analyzed using RELAP5-3D.     

Neon implantation and the radiation enhanced diffusion of platinum for the local lifetime control in high-power silicon diodes

About 10 MeV helium and 120 MeV neon implantations were used for the local lifetime control of silicon power diodes with subsequent annealing at 200 °C. DLTS measurements show that the concentration ratio between VO^(–/0) pairs and divacancies after the implantation of neon is close to one in agreement with the data published for other heavy ions. The implantation dose to achieve the same point at the technology curve of the diodes under test was found about 10 times lower for the neon compared to helium. The radiation enhanced diffusion (RED) of platinum at 725 °C was evaluated both for the enhancement by implantation of helium and neon. The electrical parameters of silicon diodes (carrier lifetime, voltage drop, leakage current and reverse recovery) were compared. One order lower implantation dose of the neon compared to that of the helium was found necessary to obtain the same improvement of electrical parameters. The RED of Pt using the neon implantation was found functional in a similar way to that of the helium. The reduction of carrier lifetime, which would be normally sufficient for robust diodes, was found for the doses of neon at about 1 × 10¹³ cm⁻². However, the simultaneous increase of background doping concentration at the end of range of neon, which increases electric field, was found responsible for the decreased static breakdown voltage, decreased turn-off ruggedness and increased leakage current.     

Simulating test for thermal mixing in the hot gas chamber of the HTR-10

The helium coolant at the outlet of the pebble bed core of the 10 MW High Temperature Gas-cooled Reactor-Test Module exhibits a severe radial temperature deviation. In order to avoid damages at the downstream components due to alternating thermal loads such as the steam generator, a hot gas chamber is especially designed to solve the problem. Thermal mixing performance of the coolant in the hot gas chamber is experimentally investigated on a 1:1.5 scale model by air. The experimental result shows that within the Reynolds number range of 1.4×10⁵–5.8×10⁵, the hot gas chamber with a radial mixer reaches excellent thermal mixing of the coolant of about 94%. The flow resistance coefficient for the hot gas chamber is also presented.     

Experimental study on the effect of angles and positions of mixing vanes on CHF in a 2 × 2 rod bundle with working fluid R-134a

In this paper, the CHF experiment on the effect of angles and position of mixing vanes was performed in a 2 × 2 rod bundle. The test section had rectangular geometry in which four rod, each with a diameter of 9.5 mm, were inserted. The rod-to-rod gap was 3.15 mm, and the rod-to-wall gap was 1.575 mm. It was vertically installed in the test loop and was uniformly heated by electricity. The heating length was 1.125 m. The working fluid was R-134a. The mass flux ranged from 1000 to 1800 kg/m². The test pressure ranged from 14.67 to 25.67 bar. CHF data in the 2 × 2 rod bundle without a mixing vane were compared to the Bowring correlation and a CHF look-up table at equivalent hydraulic diameter. For this comparison, Katto's fluid-to-fluid model is applied. The results had a good agreement with error rates of 16 and 20%. In the CHF experiment with the mixing vanes with various angles, the angles of the mixing vanes were 20–40°. The CHF enhancement ratio (CER) was largest at 30°. CHF was enhanced up to 19%. A CHF experiment on the position of the mixing vane was also performed. In the experiment on the position of mixing vane, CER was reduced with increasing distance between grid and CHF location because swirl flow decayed. We also performed the CHF experiment on mixing vane developed by KAIST.     

Conceptual design of helium cooling circuit for irradiation target

Small metal specimens of about 20 mm × 20 mm and 0.4 mm thick are irradiated in cyclotron facilities for radiation damage studies. Cooling of these specimens is an important factor which decides the intensity of irradiation. In this paper helium is used for the cooling of irradiation target specimen. In order to have enhanced heat removal from the specimen jet cooling is employed. The cooling scheme and the conceptual helium cooling circuit has been arrived at based on the empirical correlation available in the literature. The heat removal rate has been estimated for various jet velocities. Experiments with impinging air jets have been carried out to compare the empirical predictions. Numerical predictions have also been carried out using commercial Computational Fluid Dynamic (CFD) code. Experimental predictions are 35%–55% higher compare to empirical correlation. The empirical correlation is 30% higher compare to CFD predictions.     

Model validation of GAMMA code with heat transfer experiment for KO TBM in ITER

By considering the requirements for a DEMO-relevant blanket concept, Korea (KO) has proposed a He cooled molten lithium (HCML) test blanket module (TBM) for testing in ITER. A performance analysis for the thermal–hydraulics and a safety analysis for the KO TBM have been carried out using a commercial CFD code, ANSYS-CFX, and a system code, GAMMA (GAs multicomponent mixture analysis), which was developed by the gas cooled reactor in Korea. To verify the codes, a preliminary study was performed by Lee using a single TBM first wall (FW) mock-up made from the same material as the KO TBM, ferritic martensitic steel, using a 6 MPa nitrogen gas loop. The test was performed at pressures of 1.1, 1.9 and 2.9 MPa, and under various ranges of flow rate from 0.0105 to 0.0407 kg/s with a constant wall temperature condition. In the present study, a thermal–hydraulic test was performed with the newly constructed helium supplying system, in which the design pressure and temperature were 9 MPa and 500 °C, respectively. In the experiment, the same mock-up was used, and the test was performed under the conditions of 3 MPa pressure, 30 °C inlet temperature and 70 m/s helium velocity, which are almost same conditions of the KO TBM FW. One side of the mock-up was heated with a constant heat flux of 0.3–0.5 MW/m² using a graphite heating system, KoHLT-2 (Korea heat load test facility-2). Because the comparison result between CFX 11 and GAMMA showed a difference tendency, the modification of heat transfer correlation included in GAMMA was performed. And the modified GAMMA showed the strong parity with CFX 11 calculation results.     

Direct contact condensation of steam on slowly moving water

A study of direct contact condensation of stagnant saturated steam on slowly moving subcooled water has been performed with reference to a horizontal flat geometry. Inlet water mass flowrate and temperature together with inlet steam temperature have been investigated, as experimental variables, in the following ranges:

1. (a) pressure up to 6 bar,

2. (b) inlet steam temperature up to 160°C

3. (c) inlet water mass flowrate up to 120 kg/h,

4. (d) inlet water temperature up to 70°C,

5. (e) available steam mass flowrate up to 20 kg/h.

Condensation heat transfer coefficients have been determined as functions of inlet water mass flowrate, inlet water and steam temperature. Heat transfer coefficient does not show, practically, dependence either on inlet water temperature or inlet steam temperature but only on inlet water mass flowrate. Correlations are given for the Nusselt number, as a function of Reynolds and Prandtl numbers.An evaluation of thermal non-equilibrium degree between the phases is also presented, together with a correlation for its prediction.     

Hydraulic flow tests of APWR reactor internals for safety analysis

A concept of radial neutron reflector of APWR brings about safety problems relevant to the flow induced vibration and thermal deformation. The CFD code has been expected to solve them by calculating pressure fluctuations of turbulent flow in the downcomer and the flow distribution into the neutron reflector. A series of hydraulic flow tests was conducted by NUPEC from 1998 to 2002 to demonstrate the new design of the neutron reflector and to obtain test data for validating the CFD code. The measured pressure fluctuations in the downcomer and their statistics were utilized for validating the specific turbulent model to be able to calculate a spectrum of pressure fluctuation such as the LES model. The measured flow rates at inlet holes of the lower core plate were utilized for validating for the general turbulent model, for example, the k– turbulent model. The calculated results with the LES model agreed well with the measured pressure fluctuations and their spectrum, but did not agree with the correlation between adjacent pressure fluctuations. On the other hand, the calculation results with the k– turbulent model agreed well with the measured flow rates at inlet holes of the lower core plate.     

Current status of He-cooled divertor development for DEMO

A He-cooled divertor concept for DEMO is being investigated at the Forschungszentrum Karlsruhe within the framework of the EU power plant conceptual study. The design goal is to resist a heat flux of 10 MW/m² at least. The major R&D areas are design, analyses, fabrication technology, and experimental design verification. A modular design is preferred for thermal stress reduction. The HEMJ (He-cooled modular divertor with multiple-jet cooling) was chosen as reference concept. It employs small tiles made of tungsten, which are brazed to a thimble made of tungsten alloy W-1%La₂O₃. The W finger units are connected to the main structure of ODS Eurofer steel by means of a copper casting with mechanical interlock. The divertor modules are cooled by helium jets (10 MPa, 600 °C) impinging onto the heated inner surface of the thimble.In cooperation with the Efremov Institute a combined helium loop & electron beam facility (60 kW, 27 keV) was built in St. Petersburg, Russia, for experimental verification of the design. It enables mock-up testing at a nominal helium inlet temperature of 600 °C, an internal pressure of 10 MPa, and a pressure difference in the mock-up of up to 0.5 MPa. Technological studies were performed on manufacturing of the W finger mock-ups. Several high heat flux tests were successfully performed till now. Post-examination and characterisation of the mock-ups subjected to the high heat flux tests were performed in collaboration with Forschungszentrum Jülich. Altogether, the test results confirm the divertor performance required. The helium-cooled divertor concept was demonstrated to be feasible. The knowledge gained from these experiments and some aspects on the design improvement are discussed in this contribution.