Cladding Surface Temperature Research Articles

Numerical investigation on practicability of reducing MCST by using grid spacer in a tight rod bundle

The numerical investigation was carried out to reveal the practicability of reducing the maximum cladding surface temperature (MCST) within the inner sub-channel of a tight, hexagon rod bundle using commercial CFD code STAR CCM+ 6.04. The special heat transfer and pressure drop characteristics caused by four existing grid spacer designs were discussed in detail by analyzing the effects of grid strap length, different flow enhancing features and different Reynolds numbers. It was found that the local heat transfer within the grid strap is greatly enhanced due to the raised flow velocity. Both the standard grid spacer and the grid spacer with split-vanes cause decreased heat transfer in the downstream region. The friction drag is very influential in the tight rod bundle and can eliminate the positive effect of flow blockage on the heat transfer performance. The grid spacer with flow blockage discs induces relatively good heat transfer performance and higher pressure drop within sub-channels, indicating a tradeoff between the heat transfer augmentation and the pressure drop. The combination of multiple existing grid spacers can reduce the MCST to a certain level, but the corresponding disadvantages cannot be ignored. The improved grid spacer design was proposed based on the overall considerations of heat transfer and pressure drop characteristics and has been proved more suitable to widely reduce MCST for SCWR than any other grid spacer designs involved in present study.

Improvements of two-pass core design for super fast reactor

Improvements of two-pass core design of Super critical-water cooled fast reactor (Super FR) are proposed. The core height has been shortened; assembly pitch is enlarged to take Control Rod (CR) design into account. Well-coupled fuel loading pattern is proposed, number of control rods location are limited only in second flow pass assemblies due to current flow pattern. Fuel shuffling is also confined within each own flow pass because of the different structure of assemblies. CR withdrawal is considered during the operation, satisfactory radial and axial power distribution can be achieved by implementing appropriate CR withdrawal strategy. Results suggest that all the design target, criteria and limits are satisfied in terms of average coolant outlet temperature, maximum linear heat generation rate (MLHGR), maximum cladding surface temperature (MCST) as well as core shutdown margin, negative coolant void reactivity coefficient and positive coolant density reactivity coefficient in all coolant density range.

Coupled analysis and improvements for Canadian-SCWR core design

A coupled neutronics and thermal hydraulics analysis have been done for the recent Canadian SCWR, high efficiency re-entrant channel (HERC) design. The neutronics analysis was carried out by using a 3D fine mesh diffusion theory code and thermal hydraulics calculations were done by using single channel model. These two codes were coupled with each other by a link code. The axial variation of water density in the central flow tube has also been considered in the analysis. The maximum clad surface temperature for the original design was calculated to be very high, so some improvements in the channel design have also been proposed in this study. Moreover, a new core-loading scheme has been searched in order to get more flat radial power distribution and lower cladding surface temperature.

Steady-State Thermal-Hydraulic Analysis of TRIGA Research Reactor

The COOLOD-N2 and PARET computer codes were used for a steady-state thermal hydraulic and safety analysis of the 3 MW TRIGA Mark-II research reactor located at Atomic Energy Research Establishment (AERE), Savar, Dhaka, Bangladesh. The objective of the present study is to ensure that all important safety related thermal hydraulic parameters uphold margins far below the safety limits by steady-state calculations at full power. We, therefore, have calculated the hot channel fuel centreline temperature, fuel surface temperature, cladding surface temperature, the departure from nucleate boiling (DNB) heat flux and DNB ratio, axial fuel centreline temperature and compared. The comparison indicates that the calculated values are in satisfactory agreement between the codes. The data obtained in this investigation are largely far to compromise safety of the reactor. The results can also be used to upgrade the current core configuration of the TRIGA reactor.

超臨界流体中に存在するスペーサまわりの伝熱流動に関する数値解析

As one of key issues on the thermal design of supercritical water reactors (SCWRs), it is required that the maximum fuel cladding surface temperature becomes less than 700℃. Then, the authors propose a heat transfer enhancement method by spacers. In addition, the spacer has a couple of small wing-shape ribs in order to increase the turbulent heat transfer. Preliminary analyses were performed to investigate numerically the heat transfer performance of a currently proposed ribbed-spacer under the supercritical water condition. Numerical parameters are the inlet velocity and heat flux. Moreover, as attack angles of the ribs to the spacer, three conditions were provided; 0°, 15° and 30°. From the predicted results, it was confirmed that the ribbed-spacer is effective as a turbulence promoter.

Effect of Surface Modification on Candidate Alloys for Canadian SCWR Fuel Cladding

The Canadian Generation-IV supercritical water reactor (SCWR) requires peak cladding surface temperature of 800˚C for a core outlet temperature of 625˚C. Materials selection for high temperature fuel cladding is becoming one of the major challenging tasks. Austenitic stainless steels with excellent corrosion resistance are often susceptible to stress corrosion cracking upon SCW exposure. Low-Cr steels such as P91 exhibit good high-temperature mechanical properties, but the lack of sufficient Cr content makes this group of alloys corrode too fast. One possible solution is to use coatings or surface modification techniques to improve the surface resistance to corrosion. In this study, we investigated the effect of surface modification on commercial 316L stainless steel. Surface modification by mechanical deformation has marked improvement in corrosion resistance during SCW exposure. Possible mechanisms for such improvement are discussed.

Improved single pass core design for high temperature Super LWR

Core design of a supercritical-pressure light water cooled and moderated reactor (Super LWR) with single flow pass is carried out for achieving high outlet coolant temperature 500°C and simplifying upper core structure. Both the coolant in water rod and fuel channel flows up and is mixed at the upper plenum. The structure inside pressure vessel is simplified as a PWR by eliminating upper core moderator distribution/guide tubes applied in previous two pass core. The fuel assemblies loaded at peripheral core region are divided into four flow zones by using separation plates to adjust flow to the power for increasing average outlet temperature. The shuffling scheme is carried out separately for inner and peripheral core assemblies. Axial fuel enrichment is split into four segments to control the axial power peaking due to relatively large axial water density variation. Five fuel batches are adopted to increase the discharge burn-up. The equilibrium core is analyzed by employing neutronic/thermal-hydraulic coupled calculation. The numerical results show that all the design criteria are fulfilled by the maximum cladding surface temperature of 656°C with 500°C average core outlet temperature, maximum linear heat generation rate of 37.4kW/m and positive water density as well as shutdown margin of 1.45%dk/k.

Preliminary safety evaluation for CSR1000 with passive safety system

This paper describes the preliminary safety analysis of the Chinese Supercritical water cooled Reactor (CSR1000), which is proposed by Nuclear Power Institute of China (NPIC). The two-pass core design applied to CSR1000 decreases the fuel cladding temperature and flattens the power distribution of the core at normal operation condition. Each fuel assembly is made up of four sub-assemblies with downward-flow water rods, which is favorable to the core cooling during abnormal conditions due to the large water inventory of the water rods. Additionally, a passive safety system is proposed for CSR1000 to increase the safety reliability at abnormal conditions. In this paper, accidents of “pump seizure”, “loss of coolant flow accidents (LOFA)”, “core depressurization”, as well as some typical transients are analysed with code SCTRAN, which is a one-dimensional safety analysis code for SCWRs. The results indicate that the maximum cladding surface temperatures (MCST), which is the most important safety criterion, of the both passes in the mentioned incidents are all below the safety criterion by a large margin. The sensitivity analyses of the delay time of RCPs trip in “loss of offsite power” and the delay time of RMT actuation in “loss of coolant flowrate” were also included in this paper. The analyses have shown that the core design of CSR1000 is feasible and the proposed passive safety system is capable of mitigating the consequences of the selected abnormalities.

The relationship between air gap sizes and clothing heat transfer performance

The objective of the study is to establish a quantitative relationship between air gap sizes and clothing thermal performance. Using a three-dimensional human body scanner, the thicknesses and volumes of air gaps of 35 experimental shirts were measured. Relationships between the thermal insulations of clothing, measured on a thermal manikin, and air gap volumes were examined by regression analysis. Also, the regression model between clothing surface temperatures, measured by an infrared thermal camera, and air gap thicknesses was established. The results proved that the thermal insulation of experimental shirts increased with air gap sizes but began to decrease as a result of natural convection when the air gap thickness was higher than 1 cm or the air gap volume was greater than 6000 cm3. The 3D human body scanner can accurately measure air gaps under clothing, but it is expensive and is not available everywhere. A substitute method is to build a mathematical model for predicting air gap sizes. In this paper, regression models were established to estimate volumes and thicknesses of air gaps when the ease allowances of chest circumference and fabric properties were known. This study can be used to predict the thermal performance of clothing at the product’s design stage.

Single-pass core design of a low-temperature Super LWR

A supercritical-pressure light water cooled and moderated reactor (Super LWR) with a single-pass flow scheme is developed for simplifying upper core structures. Both coolant in the fuel channels and the water rods flow upward and are mixed in the upper plenum. It eliminates the moderator guide/distribution tubes in the upper core that were used in the previous Super LWR design adopting two-pass coolant flow scheme. This core design adopts a four-batch fuel management scheme and an out–in fuel loading pattern. One hundred and twenty-one fuel assemblies with an active height of 3.7 m are included. The flow rate fraction for water rods is 3.5%, and the thermal insulator is used to keep the moderator temperature below pseudocritical temperature. The equilibrium core is analyzed by using neutronic and thermal-hydraulic coupled calculation. The results show that the maximum cladding surface temperature (MCST) is limited to 485 °C with the average outlet temperature of 400 °C. The inherent safety is fulfilled by the positive water density reactivity coefficient and sufficient shutdown margin. On the other hand, the investigation of average outlet coolant temperature varying with MCST is carried out to explore the maximum outlet temperature by employing current MCST criterion and single-pass core design. The average outlet temperature increases with the MCST, and it achieves 465 °C with the thermal efficiency of 43.1% at the MCST criterion of 650 °C. The structure inside the reactor pressure vessel is simplified as a pressurized water reactor.

Thermo-Hydraulic Behavior of Water Cooling Channel Subjected to Constant Heat Flux during Pressure Reduction Transient in its Cooling System

Heat transfer parameters of concentric heat source cooling channel exposed to pressure reduction transient is studied experimentally and theoretically. The heat source is of constant heat flux cooled by upward flowing water in concentric channel. The heat source is located inside cylindrical shape tube which is fixed in an annular vertical channel. The cooling water pressure reduction transient is ensured by different shape disturbance functions. The theoretical investigation involved a mathematical modeling for axially, symmetric, simultaneously developing laminar water flow in a vertical annulus. The mathematical model is based on one dimensional flow. The boundary conditions of the studied case are based on adiabatic outer wall while the inner wall is subjected to a constant heat flux for upwards flow. The heat & mass balance equation derived for specified element of bulk water within the annulus, is solved to determine the variation of bulk water temperature, heat transfer coefficient, clad surface temperature and the boiling safety factor based on clad surface temperature versus length and time during transient course. The present theoretical work covers heat flux of 46345 W/m 2 , channel inner to outer diameter ratio of 0.8, water sub-cooled degree in the channel inlet ranging (20-30 O C), heat source length of 0.65, water pressure at channel inlet of 1.3 bars and pressure reduction transient according to step, ramp and sinusoidal shape disturbance function (1.3- 1.0) bars. The experimental investigation included a set of experiments carried out to investigate the temperature variation along the heat source for step, ramp and sinusoidal pressure reduction transients in cooling system during and after reaching the steady state condition.

Core design of super LWR with double tube water rods

Double tube water rods are employed in core design of super LWR to simplify the upper core structure and refueling procedure. The light water moderator flows up in the inner tube from the bottom of the core, then, changes the flow direction at the top of the core into the outer tube and flows out at the bottom of the core. It eliminates the moderator guide/distribution tubes into the single tube water rods from the top dome of the reactor pressure vessel of the previous super LWR design. Two rows of fuel rods are filled between the water rods in the fuel assembly. Out-in refueling pattern is adopted to flatten radial power distribution. The peripheral fuel assemblies of the core are divided into four flow zones by separation plates for increasing the average core outlet temperature. Three enrichment zones are used for axial power flattening. The equilibrium core is analyzed based on neutronic/thermal-hydraulic coupled model. The results show that, by applying the separation plates in peripheral fuel assemblies and low gadolinia enrichment, the maximum cladding surface temperature (MCST) is limited to 653°C with the average outlet temperature of 500°C. The inherent safety is satisfied by the negative void reactivity effects and sufficient shutdown margin.

Time dependent start-up thermal analysis of a Super Fast Reactor

The startup system of a supercritical pressure light water cooled fast reactor (Super FR) is studied by time dependent thermal-hydraulic analysis. The plant analysis code is developed based on an innovative upward flow pattern in all the assemblies of the Super FR. A recirculation system consisting of a steam drum, a circulation pump, and a heat exchanger is used for the startup. Detailed procedures are performed and the maximum cladding surface temperature (MCST) at rated power, 640°C, is used as the criterion. Firstly a small constant nuclear power is used for rising the core feed water temperature to be 280°C through the recirculation system. Secondly, pressurization is done in the recirculation system from atmospheric to operating pressure, 25MPa, by raising the power. Thirdly, line-switching from recirculation mode to once-through direct-cycle is performed while turbines are started by supercritical steam at supercritical pressure. Finally the power is raised to be 100% of power followed by raising the flow rate. During pressurization the heat flux margin is large due to low power used for pressurization and the MCST is much lower than the criterion. The MCST is not sensitive to the inlet temperature, the flow rate, and the gap volume of the core because of high flow to power ratio. Smaller dimension of steam drum can be used for pressurization stably. The MCST satisfies the criterion both during subcritical pressure and during power-raising.

Thermal hydraulic analysis of the AHWR—The Indian thorium fuelled innovative nuclear reactor

Analysis has been carried out for simulating loss-of-coolant accident (LOCA) at inlet header in a natural circulation type reactor developed as the advanced heavy water reactor (AHWR).The paper will cover a case of LOCA due to 200% break at inlet header which is double ended rupture. The maximum clad surface temperature has been predicted in different cases by using the thermal hydraulic safety code RELAP5/Mod4.0. The proposed reactor is a 920MWth vertical pressure tube type, boiling light water cooled and heavy water moderated reactor. One of the important passive design features of this reactor is heat removal through natural circulation of primary coolant (at all allowed power levels) with no primary coolant pumps. This reactor is equipped with emergency core cooling system (ECCS) and isolation condensers (ICs) to remove decay heat during LOCA. This ECCS provides cooling to fuel in passive mode during first fifteen minutes of LOCA and it is achieved by high pressure injection from advanced accumulator. Cooling is continued for Later for three days by the gravity driven water pool (GDWP). This paper investigates the impact of high pressure injection in this cooling process.

Study of successive ramp reactivity insertions in typical pool-type research reactors

A comparative safety assessment of both HEU and the proposed LEU1 & LEU2 cores having UAlx–Al, U3O8–Al and U3Si2–Al fuels respectively in a typical MTR system has been carried out using the PARET code. Super prompt-critical transients initiated with the insertion of single as well as double ramps at low power start-up have been studied in this work. The results for the standard HEU core with single reactivity ramps up to $2.2/0.5 s show excellent agreement with the already published data. According to these simulations, for large time gap values (Δt = 1 s) between double ramps, the first peak power dominates for all three cores and each system remains safe up to $2.2/0.5 s ramp rates. However, for closely spaced double ramps (Δt = 0.1 s), for both HEU and the LEU2 cores, the simulations indicate clad melting for $2.4/0.5 s ramp rates. The peak power, fuel centerline & clad surface temperature values remain within safe limits for single as well as for double reactivity ramps up to $2.2/0.5 s for all three cores.

A lumped parameter core dynamics model for MTR type research reactors under natural convection regime

On the basis of lumped parameter modeling of both the kinetic and thermal–hydraulic effects, a reasonably accurate simplified model has been developed to predict the dynamic response of MTR reactors following to an unprotected reactivity insertion under natural convection regime. By this model the reactor transient behavior at a given initial steady-state can be solved by a set of ordinary differential equations. The model predictions have an acceptable consent with corresponding results of reactivity insertion transients analyzed in the literature. The inherent safety characteristics of MTR research reactors utilizing natural convection is clearly demonstrated by the expanded model. The safety margin of reactor operating is selected ONB condition and thereby the proposed model determines that any slight increase in the value of $0.73 for inserted reactivity will cause the maximum cladding surface temperature to exceed the ONB condition.

Total loss of flow accident characteristics of Super FR with new coolant flow

Total loss of flow accident characteristics of the supercritical-pressure light water-cooled fast reactor (Super FR) with all upward flow core cooling in two paths is investigated at supercritical pressure. Upward flow cooling has advantage that flow stagnation does not occur at total loss of coolant flow accident due to the buoyancy of the coolant. It is found that maximum cladding surface temperature (MCST) is sensitive to the volume of the gap between two paths the. By open the automatic depressurization system (ADS), depressurization will induce flow in the RPV, and it can improve the MCST in this accident.

Conceptual design of a supercritical water reactor with double-row-rod assembly

A novel type of fuel assemblies with double rows of fuel rods between water rods is proposed and optimized for a supercritical light water reactor design. It brings improved neutron moderation and lower local power peak. Gadolinium is introduced as burnable poison to reduce excess reactivity at the beginning of the fuel cycle. The optimizations for the fuel rods with gadolinium are performed in the present paper. SS316L is used for fuel rod cladding and structural material. In order to reduce the amount of SS316L because of its high thermal neutron absorption, honeycomb structure filled with thermal isolation is introduced to replace the solid stainless steel. The two-pass water flow scheme is chosen with more fuel assemblies for downward flow. Fuel in-core loading pattern and control rod clusters pattern are designed to flatten power distribution at inner regions to enhance coolant outlet temperature. Axial fuel enrichment is zoned into three regions to control axial power peak, which might affect maximum cladding surface temperature. An equilibrium core is then analyzed based on neutronics/thermal-hydraulics coupling model. The numerical results indicate that a high average coolant outlet temperature of 500 °C is achieved with a maximum cladding surface temperature less than 650 °C. The void reactivity effects of moderator and coolant are negative throughout the cycle.

Coupled analysis for new fuel design using UN and UC for SCWR

All-ceramic bi-material coated fuel pellet design is proposed and analyzed for thermal Super Critical Water Reactor (SCWR). Uranium mono nitride (UN) with Zirconium Carbide (ZrC) coating and Uranium mono carbide (UC) with Silicon Carbide (SiC) coating are analyzed. Carbide and nitride ceramic fuels offer the advantage of high thermal conductivity as compared to UO2. Use of coating can solve the problems of harder spectrum and reactive nature for UN and UC which arise when these fuels are used in light water thermal reactors. Larger heavy metal density of UN and UC can lead to equivalent or even larger values of heavy metal loading using coated pellet concept. Coating can give the extra advantage of working as yet another barrier against release of radioactive fission gases in accidental scenarios. Due to large coolant density variation along active height of fuel, coupled neutronics/thermal hydraulic analyses are performed using MCNP4c/SACoS coupled system. Design proposed by Shanghai Jiao Tong University (SJTU) for thermal SCWR is chosen to perform the calculations. Coupled analyses show that coated fuel pellets lead to more uniform radial power density which leads to smaller value of hot channel factor. A significant decrease in fuel centerline temperature is seen due to high thermal conductivity of carbide and nitride ceramic fuels. A staggering difference of more than 500 °C is seen between the maximum average centerline temperatures for UO2 and coated fuel pellets. Slightly smaller values of average clad surface temperature are obtained using coated fuel pellets as compared to UO2. A maximum coating temperature of less than 800 °C is observed. Effect of varying coating thickness is studied by performing analyses for different coating thickness values. Keeping the same fuel pin radius and pitch, increasing the thickness of coating materials can lead to increased value of thermal neutrons but will have a negative effect on the burnup due to smaller heavy metal loading. Reactivity coefficients, i.e. Doppler, coolant and moderator, show negative values. Considering the favorable neutronic and thermal hydraulic properties along with a potential increase in cycle length, all-ceramic bi-material coated fuel pellets can be a very good fuel option for SCWR.

Methodology for estimation of conditional probability of stagnation channel break for PHWRs

A methodology to estimate the conditional probability of a break in inlet feeder being the stagnation channel break in Pressurized Heavy Water Reactors is presented in this paper. According to AERB Safety Guide AERB/SG/D-5, rupture at any location of PHT system up to size of double ended largest feeder pipe belongs to category-2 events which are events of moderate frequency. For a particular range of break sizes in inlet feeder, the flow in the corresponding reactor channel reduces almost immediately to stagnant/very low flow conditions and remains sustained due to high flow inertia of PHWR pipe structure. As it is a single channel event out of hundreds of channels, the signal for reactor trip is generated very late. Consequently the fuel temperature rises sharply and fuel failure may occur. The probability of break in a feeder leading to flow stagnation conditions in the corresponding channel is low but needs to be estimated. In this paper the conditional probability of stagnation channel break for feeder pipe is estimated deterministically for a reactor similar to Indian PHWR. In this work, the stagnation channel break is identified if minimum critical heat flux ratio (MCHFR) is sustained lower than one anywhere in the corresponding reactor channel. It leads to sudden rise in fuel clad temperature due to sudden significant reduction in heat removal. There is a range of break sizes for which the clad surface temperature rises above fuel failure limits. Depending on this range and break locations in the feeder, the conditional probability is estimated. It is assumed that conditional probability depends on the range of break sizes leading to stagnation. It is observed from the calculations that the conditional probability of stagnation channel break is significant. The analysis has been carried out using thermal hydraulic code RELAP5/Mod 3.4.