Cladding Surface Temperature Research Articles

Coupled Three-Dimensional Neutronics and Thermal-Hydraulics Analysis for SCWR Core Typical Transients

Transient performance of China supercritical water-cooled reactor (SCWR) with the rated electric power of 1000 MWel (CSR1000) core during some typical transients, such as control rod (CR) ejection and uncontrolled CR withdrawal, is analyzed and evaluated with the coupled three-dimensional neutronics and thermal-hydraulics SCWR transient analysis code. The 3D transient analysis shows that the maximum cladding surface temperature (MCST) retains lower than safety criteria 1260 °C during the process of CR ejection accident, and the MCST retains lower than safety criteria 850 °C during the process of uncontrolled CR withdrawal transient. The safety of CSR1000 core can be ensured during the typical transients under the salient fuel temperature and water density reactivity feedback and the essential reactor protection system.

Startup Thermal Analysis of a Supercritical-Pressure Light Water-Cooled Reactor CSR1000

Supercritical-pressure light water-cooled reactors (SCWR) are the only water cooled reactor types in Generation IV nuclear energy systems. Startup systems, and their associated startup characteristic analyses, are important components of the SCWR design. To analyze the entire startup system, we propose a wall heat transfer model in a paper, based on the results from a supercritical transient analysis code named SCTRAN developed by Xi’an Jiao tong Tong University. In this work, we propose a new heat transfer mode selection process. Additionally, the most appropriate heat transfer coefficient selection method is chosen from existing state-of-the-art methods. Within the model development section of the work, we solve the problem of discontinuous heat transfer coefficients in the logic transformation step. When the pressure is greater than 19 Mpa, a look-up table method is used to obtain the heat transfer coefficients with the best prediction accuracy across the critical region. Then, we describe a control strategy for the startup process that includes a description of the control objects for coolant flow rate, heat-exchange outlet temperature, system pressure, core thermal power, steam drum water-level and the once-through direct cycle loop inlet temperature. Different control schemes are set-up according to different control objectives of the startup phases. Based on CSR1000 reactor, an analytical model, which includes a circulation loop and once-through direct cycle loop is established, and four startup processes, with control systems, are proposed. The calculation results show that the thermal parameters of the circulation loop and the once-through direct cycle meets all expectations. The maximum cladding surface temperature remains below the limit temperature of 650℃. The feasibility of the startup scheme and the security of the startup process are verified.

The impact of azimuthally asymmetric carbon deposition upon pellet-clad mechanical interaction in advanced gas reactor fuel

Cracked nuclear fuel pellets were modelled in the r-θ plane with an azimuthally varying clad surface temperature boundary condition using the PELICAN set of fuel performance models for the commercial finite element software, Abaqus. The temperature boundary condition was assumed to represent heat transfer impairment due to an azimuthally asymmetric carbon deposit on advanced gas-cooled reactor pins. The model predicts the radial and azimuthal displacement of the idealised fuel fragments, together with the resulting elastic, creep and plastic strains in the cladding. These were compared to simulations assuming a uniformly hot or cold boundary condition. Apart from a short period during the return to power from reduced power (70%) operation and outages, the hoop stress in the simulation with an azimuthally varying clad surface temperature was bounded by that of models with a uniform hot or cold surface temperature. The reduced stress was proposed to be due to the greater ability of the fuel fragments to relocate in order to accommodate changes to the power level. As a result, the creep strains in the model with an azimuthally varying clad surface temperature were lower than assuming either a uniform hot or cold boundary condition.

A two-phase model for simulation of MTR type research reactor during protected and unprotected LOFA

An optimized, accurate and fast model is investigated to simulate the transient behavior of a research reactor during fast and slow LOFA (loss of flow accident). Simulation is performed in two states when the reactor scram system is initiated at critical core power and when it doesn't work properly. The latter case leads to boiling condition and hence the reactor core will be encountered with two-phase flow condition. A lumped heat conduction model for fuel and clad accompanied with a two-phase flow model for coolant are coupled with point kinetic equations to simulate reactor behavior during the accident. A combination of Runge-Kutta method and up-wind finite difference technique are adopted for the solution to the set of differential equations. The calculation performed for IAEA 10 MW benchmark problem in HEU and LEU fuel. Some important parameters such as the peak temperatures of fuel and clad are obtained for both HEU and LEU conditions during transients. Results of numerical simulations, at the first, agree well with the general physical rules and also verified through comparison with the other computer code results. According to present simulation, under both protected and unprotected LOFA transient, the peak clad surface temperatures remain in safe mode.

The approximation between thermal sensation votes (TSV) and predicted mean vote (PMV): A comparative analysis

The Fanger's predicted mean vote (PMV) model is used to evaluate thermal comfort. However, when PMV is compared to people's real thermal sensations, collected in field studies, some discrepancies are verified. One of the components for the calculation of PMV is clothing surface temperature (tcl), which can be a factor that contributes towards these discrepancies. The aim of this study was to propose alternative methods for predicted mean vote, seeking to reduce these discrepancies. The mathematical Newton's method was applied to obtaining tcl values. The PMV1 was determined by replacing the tcl values in the traditional equation of PMV as described by ISO 7730 (2005). The second model of thermal prediction, named as PMV2, was obtained by a multiple linear regression considering the thermal sensation votes, the metabolic rate and the six heat exchange mechanisms. Two groups (welders and army officers) were used to verify the accuracy of the methods used in this research. The results show that both methods were able to describe the thermal sensation votes. For the welder group, both PMV1 and PMV2 overestimated the results: when people voted TSV = 0, PMV1 = 0.64 and PMV2 = 0.23. In the case of the army officers group, applying PMV1, when TSV = 0, PMV1 = 1.47. The application of the multiple regression increased the potential of PMV2 to obtain responses closer to those provided by the occupants of the thermal environment studied: when TSV = 0, PMV2 = 0.0068, demonstrating a greater effectiveness of this method.

The effect of thermal insulation pads on heat flux, physical effort and perceived exertion during endurance exercise in cool environments

To determine the effects of thermal insulation pads on clothing surface temperature, physical effort and perceived exertion during endurance exercise in cool environments two different pants (PCOOL, PINSUL) were designed: PCOOL (Insulation: 0.055 clo) was made of a thin base material while PINSUL (Insulation: 0.131 clo) featured additional insulation pads covering ~ 30% of its surface, which were placed over the working leg muscles. Two sets of experiments were performed to compare both pants: Study A was completed in 10 active sportsmen who were instructed to run on a treadmill for 45 min at 60% of their maximal running velocity (vmax) at 7 °C. In study B, 8 endurance athletes completed the run with 70% vmax at 0 °C. Lower and upper body clothing surface temperatures (TLB, TUB), auditory canal temperature, blood lactate, heart rate, subjects´ loss of body mass and perceived exertion were measured. In both studies TLB was found to be lower with PINSUL, reflecting smaller heat loss due to the better thermal insulation. However no significant differences between pants were found for auditory canal temperature, blood lactate, heart rate, subjects´ loss of body mass or perceived exertion. Inserting insulation pads into sports apparel is a practicable approach to limit heat emission from working muscles during endurance exercise in the cold without impairing overall body-heat dissipation. However, under the environmental conditions and exercise intensities tested in this study, the thermal insulation of leg muscles failed to significantly affect parameters reflecting physical effort or perceived exertion.

ROSA/LSTF test and RELAP5 code analyses on PWR hot leg small-break LOCA with accident management measure based on core exit temperature and PKL counterpart test

An experiment was performed for the OECD/NEA ROSA-2 Project using the large scale test facility (LSTF), which simulated a hot leg small-break loss-of-coolant accident with steam generator (SG) secondary-side depressurization as an accident management measure based on core exit temperature in a pressurized water reactor (PWR). This experiment was conducted under two conditions of high-pressure to meet the PWR pressure condition and of low-pressure to meet the Primӓrkreislӓufe Versuchsanlage (PKL) condition. Core uncovery took place by core boil-off with no reflux coolant from the SGs in the LSTF test. The increase rate of the cladding surface temperatures from top to center of the core relative to the core exit temperature increased according to the linear heat rate in the LSTF test. Some discrepancies appeared between the LSTF low-pressure phase and PKL test results for the core exit temperature increase due to differences in low-temperature structures around the core exit. The RELAP5/MOD3.3 code indicated a remaining problem in the prediction of the core exit temperature due to pseudo coolant mixing. Results of uncertainty analysis for the LSTF low-pressure phase test clarified influences of the combination of the multiple uncertain parameters on peak cladding temperature within the defined uncertain ranges.

Conceptual core design study of an innovative small transportable lead-bismuth cooled fast reactor (SPARK) for remote power supply

An innovative small transportable lead-bismuth cooled fast reactor, named SPARK, with rated power of 20 MWth is proposed to operate for 20 years without refueling as a remote power supply. The SPARK core neutronics and thermal-hydraulics design and preliminary safety analysis were performed in the current study. In order to achieve a compact and light-weight core design with enhanced transportability and passive safety, the selection of reflector materials, the optimization of fuel assembly design and radial core zoning loading, and the reactivity control system design were accomplished. MgO was selected as the optimal reflector material due to its good neutron reflecting characteristics and low density. The fuel assembly design was optimized to obtain a long lifetime of core and low peak cladding surface temperature. To flatten radial power distribution, 3 radial zones were designed with different fuel pin diameters. A liquid absorber control system was implemented using 6Li-enriched liquid lithium as the neutron absorber, which significantly reduces the core height. To reduce the initial excess reactivity, fixed absorbers were installed in the scram assemblies for the first half life and then replaced by fixed reflectors for the second half life. Based on the parametric study, the optimized core design was determined, and the core neutronics and thermal-hydraulics performances were evaluated. The objective core lifetime of 18 effective full power years was fulfilled with the compact and light-weight core design, and the thermal design constraints were satisfied during the whole life. Both the control and scram systems proved to independently provide sufficient shutdown margins. Using the quasi-static reactivity balance method, the passive safety characteristics of the optimized core design were analyzed based on 5 anticipated transients without scram. Passive shutdown was achieved due to the negative reactivity feedback. The critical design constraint of the peak cladding surface temperature was satisfied for all transients.

Uncertainty analysis of ROSA/LSTF test by RELAP5 code and PKL counterpart test concerning PWR hot leg break LOCAs

Abstract An experiment was conducted for the OECD/NEA ROSA-2 Project using the large-scale test facility (LSTF), which simulated a 17% hot leg intermediate-break loss-of-coolant accident in a pressurized water reactor (PWR). In the LSTF test, core uncovery started simultaneously with liquid level drop in crossover leg downflow-side before loop seal clearing, and water remaining occurred on the upper core plate in the upper plenum. Results of the uncertainty analysis with RELAP5/MOD3.3 code clarified the influences of the combination of multiple uncertain parameters on peak cladding temperature within the defined uncertain ranges. For studying the scaling problems to extrapolate thermal-hydraulic phenomena observed in scaled-down facilities, an experiment was performed for the OECD/NEA PKL-3 Project with the Primarkreislaufe Versuchsanlage (PKL), as a counterpart to a previous LSTF test. The LSTF test simulated a PWR 1% hot leg small-break loss-of-coolant accident with steam generator secondary-side depressurization as an accident management measure and nitrogen gas inflow. Some discrepancies appeared between the LSTF and PKL test results for the primary pressure, the core collapsed liquid level, and the cladding surface temperature probably due to effects of differences between the LSTF and the PKL in configuration, geometry, and volumetric size.

Fifty years of Fanger's equation: Is there anything to discover yet?

This short communication to the Editor is completely devoted to a recent paper published by Broday et al. (2017) in which a comparative analysis of methods for determining the clothing surface temperature is reported. Contrarily to what our colleagues have found, we will demonstrate that the algorithms reported in ISO 7730 and ASHRAE 55 Standards from more than 30 years are reliable and consistent with other home-made codes based on different numerical techniques.

Optimization of the Canadian SCWR Core Using Coupled Three-Dimensional Reactor Physics and Thermal-Hydraulics Calculations

The Canadian supercritical water-cooled reactor (SCWR) design is part of Canada's Generation IV reactor development program. The reactor uses batch fueling, light water above the thermodynamic critical point as a coolant and a heavy water moderator. The design has evolved considerably and is currently at the conceptual design level. As a result of batch fueling, a certain amount of excess reactivity is loaded at the beginning of each fueling cycle. This excess reactivity must be controlled using a combination of burnable neutron poisons in the fuel, moderator poisons, and control blades interspersed in the heavy water moderator. Recent studies have shown that the combination of power density, high coolant temperatures, and reactivity management can lead to high maximum cladding surface temperatures (MCST) and maximum fuel centerline temperatures (MFCLT) in this design. This study focuses on improving both the MCST and the MFCLT through modifications of the conceptual design including changes from a 3 to 4 batch fueling cycle, a slightly shortened fuel cycle (although exit burnup remains the same), axial graded fuel enrichment, fuel-integrated burnable neutron absorbers, lower reactivity control blades, and lower reactor thermal powers as compared to the original conceptual design. The optimal blade positions throughout the fuel cycle were determined so as to minimize the MCST and MFCLT using a genetic algorithm and the reactor physics code PARCS. The final design was analyzed using a fully coupled PARCS-RELAP5/SCDAPSIM/MOD4.0 model to accurately predict the MCST as a function of time during a fueling cycle.

Corrosion-based operational limit for aluminum clad monolithic U-Mo research reactor fuel plates

Aluminum clad research reactor fuel plates undergo corrosion during their service. The corrosion product accumulates on the surface of the cladding and may cause fuel failure if it reaches a critical thickness. To avoid the failure, the fuel plate power, cladding surface temperature and irradiation time must be limited.This study discusses corrosion-induced failure mechanisms and applicability of the corrosion-based operational limits to the U-Mo monolithic fuel that is proposed for conversion of some research reactors to the low enriched uranium fuel.

An experimental platform for real-time measurement of the deformation of nuclear fuel rod claddings submitted to thermal transients.

We report on experimental development and qualification of a system developed to detect and quantify the deformations of the cladding surface of nuclear fuel pellet assemblies submitted to heat transient conditions. The system consists of an optical instrument, based on 2 wavelengths speckle interferometry, associated with an induction furnace and a model pellet assembly used to simulate the radial thermal gradient experienced by fuel pellets in pressurized water reactors. We describe the concept, implementation, and first results obtained with this system. We particularly demonstrate that the optical system is able to provide real time measurements of the cladding surface shape during the heat transients from ambient to high temperatures (up to a cladding surface temperature of 600 °C) with micrometric resolution.

Fuel performance simulation of iron-chrome-aluminum (FeCrAl) cladding during steady-state LWR operation

Alternative cladding materials have been proposed to replace the currently used zirconium (Zr)-based alloys, in order to improve the accident tolerance of light water reactor (LWR) fuel. Of these materials, there is a particular focus on iron-chromium-aluminum (FeCrAl) alloys that exhibit much slower oxidation kinetics in high-temperature steam than Zr-alloys. This behavior should decrease the energy release due to oxidation and allow the cladding to remain integral longer in the presence of high temperature steam, making accident mitigation more likely. Within the development of these alloys, suitability for normal operation must also be demonstrated. This article is focused on modeling the integral thermo-mechanical performance of FeCrAl clad UO2 fuel during normal reactor operation. Finite element analysis has been performed to assess commercially available FeCrAl alloys (namely Alkrothal 720 and APMT) as a candidate fuel cladding replacement for Zr-alloys, using the MOOSE-based fuel performance code BISON. These simulations identify the effects of the mechanical-stress and irradiation responses of FeCrAl and provide a comparison with Zr-alloys. In comparing these cladding materials, fuel rods have been simulated for normal reactor operation and simple steady-state operation. Normal reactor operating conditions target the cladding performance over the rod lifetime (∼4 cycles) for the highest-power rod in the highest-power fuel assembly under reactor power maneuvering. These power histories and axial temperature profiles input into BISON were generated from a neutronics study on full-core reactivity equivalence for FeCrAl using the 3D full core simulator NESTLE. The fuel rod designs and operating conditions used here are based on the Peach Bottom BWR with representative GE-12/14 fuel geometries, and design consideration was given to minimize the neutronic penalty of the FeCrAl cladding by changing fuel enrichment and cladding thickness. Individual sensitivity analyses of the fuel and cladding creep responses were also performed, which indicated the influence of compliance for each material, separately, on the stress state of the fuel cladding. These parametric analyses are performed using steady-state operating conditions such as a simple axial power profile, a constant cladding surface temperature, and a constant fuel power history.

Burnable Poison Design for Supercritical Water Cooled Reactor With Annular Fuel

Burnable poison (BP) is used to control excess reactivity in supercritical water cooled reactor (SCWR). It helps reduce the number of control rods. Over all BP designs, the design in which rare-earth oxide mixes with fuel is widely used in SCWR. BP has influence on fuel assembly neutronics performance. After comparing four kinds of rare-earth oxide, Er2O3 is chosen as BP for the annular fuel assembly. The effect of different BP loading patterns on assembly power distribution is analyzed. The safety of annular fuel assembly is estimated with different BP contents. Core performance with and without BP is compared. The results had shown that the core radial power peaking factor decreased after introducing BP. It was also shown that the core axial power peaking factor increased, and the power peak moved toward the top of the core. The reason of this effect was studied. Two optimizations were given based on this study: decreasing the temperature of lower plenum and increasing the gradients of axial enrichments. By applying these optimizations, core axial power peaking factor and maximum cladding surface temperature decreased.

Improved Core Design of a High Breeding Fast Reactor Cooled by Supercritical Pressure Light Water

The authors look for an attractive light water reactor (LWR) concept, which achieves high breeding performance with respect to the compound system doubling time (CSDT). In the preceding study, a high breeding fast reactor concept, cooled by supercritical pressure light water (Super FBR), was developed using tightly packed fuel assembly (TPFA) concept, in which fuel rods were arranged in a hexagonal lattice and packed by contacting each other. However, the designed concept had characteristics, which had to be improved, such as low power density (7.4 kW/m), large core pressure loss (1.02 MPa), low discharge burnup (core average: 8 GWd/t), and low coolant temperature rise in the core (38 °C). The aim of this study is to clarify the main issues associated with improvement of the Super FBR with respect to these design parameters and to show the improved design. The core design is carried out by fully coupled three-dimensional neutronics and single-channel thermal-hydraulic core calculations. The design criteria are negative void reactivity, maximum linear heat generation rate (MLHGR) of 39 kW/m, and maximum cladding surface temperature (MCST) of 650 °C for advanced stainless steel. The results show that significant improvement is possible with respect to the core thermal-hydraulic characteristics with minimal deterioration of CSDT by replacing TPFA with the commonly acknowledged hexagonal tight lattice fuel assembly (TLFA). Further design studies are necessary to improve the core enthalpy rise by reducing the radial power swing and power peaking.

Thermal-hydraulic analysis of research reactor core with different LEU fuel types using RELAP5

Abstract In the current work, comparisons between the core performances when using different LEU fuels are done. The fuels tested are UA1X-A1, U3O8-Al, and U3Si2-Al fuels with 19.7 % enrichment. Calculations are done using RELAP5 code to evaluate the thermal-hydraulic performance of the IAEA benchmark 10 MW reactor. First, a reassessment of the slow reactivity insertion transient with UA1X-A1 LEU fuel to compare the results with those reported in the IAEA TECDOC [1]. Then, comparisons between the thermal-hydraulic core performances when using the three LEU fuels are done. The assessment is performed at initial power of 1.0 W. The reactor power is calculated using the RELAP5 point kinetic model. The reactivity feedback, from changes in water density and fuel temperature, is considered for all cases. From the results it is noticed that U3Si2-Al fuel gives the best fuel performance since it has the minimum value of peak fuel temperature and the minimum peak clad surface temperature, as operating parameters. Also, it gives the maximum value of the Critical Heat Flux Ratio and the lowest tendency to flow instability occurrence.

CFD analysis of heat transfer in hexagonal subchannels of super fast reactor in upward flow

AbstractA super fast reactor is a fast spectrum, supercritical, water‐cooled reactor. This paper represents CFD analysis of heat transfer in hexagonal subchannels of super fast reactor using FLUENT in ANSYS. The numerical simulation of grid stability was done by considering different mesh sizes and the turbulence model for heat transfer of supercritical water was also carried out and compared with the experimental data. RNG k‐ϵ turbulence model with enhanced wall treatment was considered for simulations. Heat transfer and heat generation rate analysis of the outer surface rod wall is carried out with different subchannels by changing various parameters like boundary conditions and pitch‐to‐diameter ratio. The analyses reveal that the outer surface of the rod wall temperature decreases with increase in pitch‐to‐diameter ratio. Maximum coolant temperature rises in edge subchannels more than corner subchannels. Further analysis is carried out with different mass fluxes. Increases in mass flux has minimal effect on the maximum rod wall surface temperature. Maximum cladding surface temperature for the corner subchannel is less compared to the edge subchannel. Heat generation rate also decreases with increase in pitch‐to‐diameter ratio. This paper also investigates the buoyancy effect on subchannels with varying heat flux as boundary conditions considering constant mass flux.

Neutronics and thermal hydraulic analysis of TRIGA Mark II reactor using MCNPX and COOLOD-N2 computer code

The neutronic analysis of TRIGA Mark II reactor has been performed. A detailed model of the reactor core was conducted including standard fuel elements, fuel follower control rods, and irradiation devices. As the approach to safety nuclear design are based on determining the criticality (keff), reactivity worth, reactivity excess, hot rod power factor and power peaking of the reactor, the MCNPX code had been used to calculate the nuclear parameters for different core configuration designs. The thermal-hydraulic model has been developed using COOLOD-N2 for steady state, using the nuclear parameters and power distribution results from MCNPX calculation. The objective of the thermal-hydraulic model is to determine the thermal safety margin and to ensure that the fuel integrity is maintained during steady state as well as during abnormal condition at full power. The hot channel fuel centerline temperature, fuel surface temperature, cladding surface temperature, the departure from nucleate boiling (DNB) and DNB ratio were determined. The good agreement between experimental data and simulation concerning reactor criticality proves the reliability of the methodology of analysis from neutronic and thermal hydraulic perspective.

ROSA/LSTF Test and RELAP5 Analyses on PWR Cold Leg Small-Break LOCA with Accident Management Measure and PKL Counterpart Test

Abstract An experiment using the Primӓrkreislӓufe Versuchsanlage (PKL) was performed for the OECD/NEA PKL-3 Project as a counterpart to a previous test with the large-scale test facility (LSTF) on a cold leg small-break loss-of-coolant accident with an accident management (AM) measure in a pressurized water reactor. Concerning the AM measure, the rate of steam generator (SG) secondary-side depressurization was controlled to achieve a primary depressurization rate of 200 K/h as a common test condition; however, the onset timings of the SG depressurization were different from each other. In both tests, rapid recovery started in the core collapsed liquid level after loop seal clearing, which caused whole core quench. Some discrepancies appeared between the LSTF and PKL test results for the core collapsed liquid level, the cladding surface temperature, and the primary pressure. The RELAP5/MOD3.3 code predicted the overall trends of the major thermal-hydraulic responses observed in the LSTF test well, and indicated a remaining problem in the prediction of primary coolant distribution. Results of uncertainty analysis for the LSTF test clarified the influences of the combination of multiple uncertain parameters on peak cladding temperature within the defined uncertain ranges.