Development and application of CFD and subchannel coupling analysis code for lead‐cooled fast reactor

It is of necessity and importance for the simulation of the three‐dimensional thermal hydraulics problem of the pool type fast reactor. However, because of current computing power limitations and the complexity of the reactor core structure, for conventional reactor applications, it is still not possible to directly simulate the entire reactor flow with sufficient fine meshes for detailed pin geometry. Until now, there is a multiscale coupling method which is suitable to deal with this type of simulation challenge. Through the user‐defined function (UDF) of FLUENT, the coupling code FLUENT/KMC‐sub for thermal hydraulic (TH) analysis by coupling the dynamic link library (DLL) complied by the transient subchannel code KMC‐sub is developed by University of Science and Technology of China (USTC). As a code validation case, the steady‐state simulation of a 19‐rod assembly has been carried out by using coupling codes of FLUENT/KMC‐sub, FLUENT and KMC‐sub, and consequently good consistency has been achieved by comparison with experiment results. And coupled code is further tested by comparison with the transient‐state 19‐pin assembly test results of KMC‐sub and FLUENT simulation. This coupling code is then used for TH of M2LFR‐1000 (medium‐size modular lead‐cooled fast reactor) in unprotected loss of flow (ULOF) accident. The transient temperatures of coolant and fuel and multidimensional TH phenomena and safety analysis are presented and discussed in this article.     

Preliminary design of a medium‐power modular lead‐cooled fast reactor with the application of optimization methods

Based on research and development experience from Gen III, Gen III+, and Gen IV reactor concepts, a 1000‐MWt medium‐power modular lead‐cooled fast reactor M²LFR‐1000 was developed by University of Science and Technology of China (USTC), aiming at achieving a reactor design fulfilling the Gen IV nuclear system requirements and meanwhile emphasizing application of optimization methods in preliminary design phase. By using the optimization methods presented, primarily considering the safety design limits (the maximum coolant velocity, the maximum cladding temperature, and the maximum burn‐up limited by the cladding radiation damage permitted), the preliminary design of 1000‐MWth medium‐power modular lead‐cooled fast reactor M²LFR‐1000 was carried out, including the design of fuel rods, fuel assemblies, reactivity control system, primary system, secondary system, decay heat removal system, and so on. The analysis of neutron characteristics (including reactivity feedback coefficients) and thermal hydraulics characteristics (the maximum fuel temperature and the maximum cladding outer surface temperature) of the core under normal steady‐state condition was carried out to evaluate the core design. Also, the analysis of 2 typical protected transients (protected transient over power accident and protected loss of flow accident) was conducted. Other analysis work of the reactor is to be done, such as the transient analysis via computational fluid dynamic codes and the seismic response analysis of the reactor. But the preliminary analysis results obtained so far under normal steady state and transient conditions confirm the inherent safety characteristics of the reactor design.     

Thermal‐hydraulic design and analysis of a small modular molten salt reactor (MSR) with solid fuel

A 20 MWth, 540 EFPD once through fuel cycle small modular molten salt reactor with solid fuel is proposed by Massachusetts Institute of Technology for off‐grid applications. In this paper, various thermal‐hydraulic analysis methods including computational fluid dynamics, Reactor Excursion Leak Analysis Program (RELAP5), and DAKOTA are adopted step‐by‐step for the reactor design based on the neutronic analysis results. First, 1/12th full core thermal hydraulic analysis is performed by using STAR CCM⁺ with most conservative considerations. Second, the transient safety behaviors of reactor system with risky assumptions are conducted by using REALP5. Finally, due to the unknown factors affecting reactor thermal‐hydraulic characteristics, the uncertainty quantification and sensitivity analysis for the designed reactor is performed with DAKOTA code coupled with RELAP5. Numerical results show that a more uniform temperature distribution with reduced peak temperatures of fuel and coolant across the reactor core has been achieved. Enough safety margin is maintained even under most severe transient accident. The uncertainties in the heat transfer coefficient and helium gap conductivity factor are the most remarkable contributors to the statistical results of peaking fuel temperature. All above results preliminarily indicate the feasibility of the current small modular molten salt reactor design and provide the further optimization direction from reactor thermal‐hydraulic prospective.     

Neutronic/thermal‐hydraulic design features of an improved lead‐bismuth cooled small modular fast reactor

Lead‐based fast reactors (LFRs) have unique advantages in the development of a SMR, which has attracted a lot of attention in recent years. In this paper, an optimized design for a lead‐bismuth small modular reactor was studied on the basis of the design of SUPERSTAR. This paper aims to propose an improved LFR core scheme to enhance the neutronic performance as well as the thermal‐hydraulic safety of the reference reactor. Advanced nitride fuel is adopted in which the plutonium is used as the driven fuel, while thorium is used as the fertile fuel. Subchannel analysis was performed in the assembly design using an in‐house subchannel code, SUBAS, and an 11 × 11 scheme with a pitch‐to‐diameter (P/D) ratio of 1.4 was chosen. Using the modified assembly, the core was redesigned using the coupled code MCORE. The active core was divided into four zones with different enrichment of ²³⁹Pu to extend the core lifetime and flatten the power distribution. The main kinetic parameters and reactivity coefficients were obtained. Neutronic performance at different operation times was also studied. The maximum radial power peak factor was 1.28, while the maximum total power peak factor was 1.737. During the whole lifetime, the reactivity swing was 0.926$, which was below the limit of 1$. The subchannel study of the core flow distribution showed that a flow distributor is needed to further improve the flow distribution capability. The peaking cladding temperature was 508.7°C, and the maximum fuel center temperature was 723.4°C, both of which do not exceed the limit temperature. Compared with features of SUPERSTAR, the peaking cladding temperature was well improved and the lifetime extended.     

Thermal-hydraulic modeling and transient analysis of pressurized water reactors

This paper examines a new dynamic moving boundary thermal-hydraulic fuel pin model (FUELPIN) for the transient analysis of a pressurized water reactor (PWR). FUELPIN is developed to accommodate the reactor core thermal-hydraulic model of the fuel pin and adjacent coolant flow channel, with detailed thermal conduction in fuel elements. Transient analyses using a known thermal-hydraulic analysis code, COBRA, and FUELPIN linked with a PWR system analysis code show that the thermal margin gains more by a transient MDNBR approach than the traditional quasi-steady methodology for a PWR. The studies of the nuclear reactor system show that moving boundary formulation is highly suitable for the transient thermal-hydraulic analysis of PWRs.     

Analysis of load‐following characteristics for an integrated pressurized water reactor

Integrated pressurized water reactor (IPWR) usually be equipped with once‐through steam generators (OTSGs). The OTSG has many advantages such as simple mechanical structure, smaller size, and higher heat transfer efficiency. It produces superheated steam but with less inventory in its secondary side. The steam pressure is easily affected by steam flow rate or feed water flow rate. This draws more attention to design advanced reactor control system. In this paper, a study has been carried out to analyze the thermal hydraulic performance of an advanced IPWR under steady‐state and transient conditions by using a thermal hydraulic safety analysis code Relap5. An effective load‐following control system is proposed. The steady‐state operating characteristics of IPWR at different load conditions show that the average primary coolant temperature, steam pressure, and coolant mass flow rate are the most important control parameters. Pump frequency conversion strategy and OTSG grouping run strategy are used to study the transient operating characteristics of IPWR. Simulation results of the control system demonstrate its capability in regulating feedwater flow rate and reactor power to follow the change of steam flow rate. According to the results, the OTSG grouping run strategy is optimized to ensure the OTSG operates safely under low load conditions. Copyright © 2013 John Wiley & Sons, Ltd.     

Core design of long‐cycle small modular lead‐cooled fast reactor

A core design of small modular liquid‐metal fast reactor (SMLFR) cooled by lead‐bismuth eutectic (LBE) was developed for power reactors. The main design constraint on this reactor is a size constraint: The core needs to be small enough so that (1) it can be transported in a spent nuclear fuel (SNF) cask to meet the electricity demands in remote areas and off‐grid locations or so that (2) it can be used as a power source on board of nuclear icebreaker ships. To satisfy this design requirement, the active core of the reactor is 1 m in height and 1.45 m in diameter. The reactor is fueled with natural and 13.86% low‐enriched uranium nitride (UN), as determined through an optimization study. The reactor was designed to achieve a thermal power of 37.5 MW with an assumption of 40% thermal efficiency by employing an advanced energy conversion system based on supercritical carbon dioxide (S‐CO₂) as working fluid, in which the Brayton cycle can achieve higher conversion efficiencies and lower costs compared to the Rankine cycle. The outer region of the core with low‐enriched uranium (LEU) performs the function of core ignition. The center region plays the role of a breeding blanket to increase the core lifetime for long cycle operation. The core working fluid inlet and outlet temperatures are 300°C and 422°C, respectively. The primary coolant circulation is driven by an electromagnetic pump. Core performance characteristics were analyzed for isotopic inventory, criticality, radial and axial power profiles, shutdown margins (SDM), reactivity feedback coefficients, and integral reactivity parameters of the quasi‐static reactivity balance. It is confirmed through depletion calculations with the fast reactor analysis code system Argonne Reactor Computation (ARC) that the designed reactor can be operated for 30 years without refueling. Preliminary thermal‐hydraulic analysis at normal operation is also performed and confirms that the fuel and cladding temperatures are within normal operation range. The safety analysis performed with the ARC code system and the UNIST Monte Carlo code MCS shows that the conceptual core is favorable in terms of self‐controllability, which is the first step towards inherent safety.     

Transient characteristics of small molten salt reactor during blockage accident

This paper reports the results from a transient core analysis of a small molten salt reactor (MSR) when a duct blockage accident occurred. The focus of this study is a numerical model employed in order to consider the interaction among fuel salt flow, heat transfer, and nuclear reactions. The numerical model comprises continuity and momentum conservation equations for fuel salt flow, two‐group neutron diffusion equations for fast and thermal neutron fluxes, transport equations for six‐group delayed neutron precursors, and energy conservation equations for fuel salt and graphite moderators. The analysis results show the following: (1) the effect of the self‐control performance of the MSR on the effective multiplication factor and thermal power output of the reactor after the blockage accident is insignificant, (2) fuel salt and graphite moderator temperatures increase drastically but locally at the blockage area and its surroundings, (3) the highest fuel salt temperature after the blockage accident is 1,363 K; this value is lower than the boiling point of fuel salt and the melting temperature of the reactor vessel, (4) the change in the distributions of fast and thermal neutron fluxes after the blockage accident when compared with the distributions at the rated condition is very slight, and (5) delayed neutron precursors, especially the first delayed neutron precursor, accumulate at the blockage area due to its large decay constant. These results imply that the safety of the MSR is assured in the case of a blockage accident. © 2006 Wiley Periodicals, Inc. Heat Trans Asian Res, 35(6): 434–450, 2006; Published online in Wiley InterScience ( www.interscience.wiley.com ). DOI 10.1002/htj.20123     

Thermal‐hydraulic design features of a micronuclear reactor power source applied for multipurpose

Microheat pipe cooled reactor power source (HRP) designed for space or underwater vehicles meets the future demands, such as safer structure, longer operating time, and fewer mechanical moving parts. In this paper, potassium heat pipe cooled reactor power source system which generates 50 kWe electricity is proposed. The reactor core using uranium nitride fuel is cooled by 37 potassium high‐temperature heat pipes. The shields are designed as tungsten and water, and reactor reactivity is controlled by control drums. The thermoelectric generator (TEG) consists of thermoelectric conversion units and seawater cooler. The thermoelectric conversion units convert thermal energy to electric energy through the high‐performance thermoelectric material. A code applied for designing and analyzing the reactor power system is developed. It consists of multichannel reactor core model, heat pipe model using thermal resistance network, thermoelectric conversion, and thermal conductivity model. Then, the sensitivity analysis is performed on two key parameters including the length of the heat pipe condensation section and the cold junction temperature of the TE cell. Meanwhile, the steady‐state calculations are conducted. Results show that the maximum fuel temperature is 938 K located in the center of reactor core and the outlet temperature of coolant reaches 316 K. Both of them are within the limitation. It is concluded that the preliminary design of HPR design is reasonable and reliable. The designed residual heat removal system has sufficient safety margin to release the decay heat of the reactor. This research provides valuable analysis for the application of micronuclear power source.     

Modeling of steady‐state and transient thermal performance of a Li‐ion cell with an axial fluidic channel for cooling

Thermal management of Li‐ion cells is an important technological problem for energy conversion and storage. Effective dissipation of heat generated during the operation of a Li‐ion cell is critical to ensure safety and performance. In this paper, thermal performance of a cylindrical Li‐ion cell with an axial channel for coolant flow is analyzed. Analytical expressions are derived for steady‐state and transient temperature fields in the cell. The analytical models are in excellent agreement with finite‐element simulation results. The dependence of the temperature field on various geometrical and thermal characteristics of the cell is analyzed. It is shown that coolant flow through even a very small diameter axial channel results in significant thermal benefit. The trade‐off between thermal benefit and reduction in cell volume, and hence capacity due to the axial channel, is analyzed. The effect of axial cooling on geometrical design of the cell, and transient thermal performance during a discharge process, is also analyzed. Results presented in this paper are expected to aid in the development of effective cooling techniques for Li‐ion cells based on axial cooling. Copyright © 2014 John Wiley & Sons, Ltd.     

Improvement of solidification model and analysis of 3D channel blockage with MPS method

In a severe accident of a nuclear power reactor, coolant channel blockage by solidified molten core debris may significantly influence the core degradations that follow. The moving particle semi-implicit (MPS) method is one of the Lagrangian-based particle methods for analyzing incompressible flows. In the study described in this paper, a novel solidification model for analyzing melt flowing channel blockage with the MPS method has been developed, which is suitable to attain a sufficient numerical accuracy with a reasonable calculation cost. The prompt velocity diffusion by viscosity is prioritized over the prompt velocity correction by the pressure term (for assuring incompressibility) within each time step over the “mushy zone” (between the solidus and liquidus temperature) for accurate modeling of solidification before fixing the coordinates of the completely solidified particles. To sustain the numerical accuracy and stability, the corrective matrix and particle shifting techniques have been applied to correct the discretization errors from irregular particle arrangements and to recover the regular particle arrangements, respectively. To validate the newly developed algorithm, 2-D benchmark analyses are conducted for steady-state freezing of the water in a laminar flow between two parallel plates. Furthermore, 3-D channel blockage analyses of a boiling water reactor (BWR) fuel support piece have been performed. The results show that a partial channel blockage develops from the vicinity of the speed limiter, which does not fully develop into a complete channel blockage, but still diverts the incoming melt flow that follows to the orifice region.     

EFFECT OF BUOYANCY ON FORCED CONVECTION IN A CUSPED DUCT

ABSTRACT

In the event of a loss of coolant accident in a pressurized water reactor, swelling of the fuel rod cladding will lead to reduction of the subchannel flow area and worsening of the core heat transfer in the region of the blockage. The four-cusped duct is an ideal geometry for the simulation of such a channel blockage. Understanding the characteristics of flow and heat transfer in the cusped duct is essential for better design of the emergency core cooling system. Thus, in this paper, combined natural and forced convection in a vertical cusped duct has been investigated in the region of both hydrodynamically and thermally fully developed flow. The thermal boundary condition imposed on the cusped duct is the axial uniform heat flux with peripheral uniform temperature. The results indicate that the fluid flow and heal transfer in the comer region of the cusped duct are improved because of the influence of natural convection. As the Rayleigh number increases, the friction factor and Nusselt number increase accordingly. It was also found that the critical Rayleigh number is 1200, at which flow reversal occurs in the buoyancy-assisted flow ( heated upflow). The velocity, temperature, and local Nusselt number distribution are presented for a range of Rayleigh numbers.     

One-dimensional experiments of a natural circulation two-phase flow under an external reactor vessel cooling

During a severe accident in a nuclear power plant, a decay heat from a molten corium should be removed to maintain an integrity of the reactor vessel. This feasible strategy can be achieved by a External Reactor Vessel Cooling (ERVC) which requires a coolant to be circulated sufficiently between the reactor vessel and its insulation. For this reason, one-dimensional experiments were conducted to estimate the natural circulation flow under the ERVC condition of the APR1400. The experimental facility is one-dimensional and scaled down to be a half height and a 1/238 channel area of the APR1400 reactor vessel. The natural circulation mass flow rates were measured with the various coolant inlet/outlet areas, heights of the supplied water level and the coolant outlet, and steam generation rates. In results, the natural circulation mass flow rates mainly depended on the inlet/outlet area, and the natural circulation mass flow rate increased, as the outlet height as well as the supplied water level increased.     

Thermal Feasibility Analyses for the 1356MWe High Conversion-Type Innovative Water Reactor for Flexible Fuel Cycle

A new reactor concept of innovative water reactor for flexible fuel cycle (FLWR) is under development at Japan Atomic Energy Agency in cooperation with Japanese reactor suppliers. A design of 1,356 MWe high conversion boiling water reactor-type FLWR core, which has an instantaneous conversion ratio of 1.04, negative void coefficient, high burnup of 65 GWd/t, and 15-month operational cycle length, has been constructed. So far, studies on thermal-hydraulic characteristics have been performed for tight lattice core. Evaluation methods for the critical power and the pressure drop under both the steady and the transient states have been established, and a modified TRAC-BF1 code has been developed for the thermal-hydraulic design of the FLWR. In this paper, the thermal feasibility of the designed 1356MWe FLWR core is analyzed by using the modified TRAC-BF1 code. The analysis is first carried out for the current core design. It is confirmed that no boiling transition (BT) occurs under the steady state. However, the minimum critical power ratio (MCPR) is only about 1.08, and the BT is confirmed occurring under the postulated abnormal transient processes. Therefore, concretizations of the conditions that ensure the thermal feasibility of a natural circulation-type FLWR and a forced circulation-type FLWR are performed. As for the results, for a forced circulation-type FLWR, the operation-limited MCPR (OLMCPR) is 1.32, and the necessary minimum core coolant flow rate is 640 kg/(m²s). For a natural circulation-type FLWR, the OLMCPR is 1.19, and the necessary minimum core coolant flow rate is 560 kg/(m²s).     

Investigation of core meltdown phenomena and radioactive materials release in VVER-1000/V446 nuclear reactor at severe accident condition due to LBLOCA along SBO

In this study, VVER-1000/V446 nuclear reactor modeling using RELAP5/SCDAP3.4 code to investigate the reactor core behavior during severe accident conditions in a LBLOCA scenario along with station-black-out (SBO) is carried out. The analyses are performed in two stages, before and after the core heat up, without considering operator’s action on the accident management. Fuel assemblies in the core are grouped into five based on average power peaking factors and are modeled in SCDAP code. For each group a corresponding channel is modeled in RELAP5 code plus a bypass channel. In the first stage, the study of ECCS and KWU tanks efficiency to keep the reactor core in the safe condition, the calculation of the elapsed time before the reactor core heat up and the estimation of available time for operator’s action to avoid core degradation, are investigated. In the second part the results of Hydrogen production rate, cladding oxide thickness, cladding damage level, release of fission products into the coolant are studied. Analysis of the scenario by the code shows the production of around 350kg Hydrogen with the maximum rate of about 1kg/s and releasing a large amount of FPs in the order of 10kg. The results also demonstrate that the operators have ∼3 h before the fuel rod cladding rupture and ∼2.5 h before the inception of exothermic steam-zirconium reaction. Finally, using a geometric mesh for the lower plenum and applying the COUPLE code, the results show that the core slumping into the lower plenum and the lower plenum rupture occur at 17561 and 18370 seconds after the onset of accident, respectively.     

Comparative study of high concentrating photovoltaics integrated with phase‐change liquid film cooling system

In high concentrating photovoltaic systems, thermal regulation is of great importance to the conversion efficiency and the safety of solar cells. Direct‐contact liquid film cooling technique is an effective way of thermal regulation with low initial investment. Tilt of solar cells is common in concentrating solar systems. An evaluation of direct‐contact liquid film cooling technique behind tilted high concentration photovoltaics was performed using both experimental and computational approaches. In the experiment, deionized water was used as the coolant at the back of simulated solar cells. Solar cell inclination of 0° to 75° with inlet water flow rate of 100–300 L/hour and inlet temperature of 30°C to 75°C were experimentally investigated. A two‐dimensional model was developed using computational fluid dynamics technique and validated by experimental results. The effects of inclination on average temperature, temperature uniformity, and heat transfer coefficient were discovered in this paper. The results indicated that 20° is the optimum angle for liquid film cooling. In addition, optimum inlet width, temperature, and velocity for inclination over 30° are 0.75 mm, 75°C, and 0.855 m/s, respectively.     

Loop analysis of a natural circulation two-phase flow under an external reactor vessel cooling

Under ERVC (External Reactor Vessel Cooling) conditions in a severe accident, a natural circulation two-phase flow is driven and a decay heat from the reactor vessel wall can be removed for an integrity of a reactor vessel wall. In this study, to estimate a natural circulation mass flow rate and to analyze the major factor determining the natural circulation mass flow rate, a loop analysis using the drift flux model was carried out and the calculation results were compared with experimental ones. From the results, the calculated circulation mass flow rate was similar to the experimental results within about a 15% error bound. And it is estimated that the shape factor of the coolant inlet and outlet is dominant for the calculation of the natural circulation mass flow rate and that a modeling of the coolant inlet and outlet should be improved to predict an accurate natural circulation mass flow rate.     

Heat transfer in high‐aspect‐ratio rectangular passage with skewed ribs

The heat transfer characteristics and flow behavior in a rectangular passage with two opposite 45° skewed ribs for turbine rotor blade have been investigated for Reynolds numbers from 7800 to 19,000. In this blade, the spanwise coolant passage at the trailing edge region whose thickness is very thin is chosen, so the channel aspect ratio (=width/height of channel) is extremely high, 4.76. Therefore the heat transfer experiment in the high‐aspect‐ratio cooling channel was performed using thermochromic liquid crystal and thermocouples. Furthermore, the calculation of flow and heat transfer was carried out using CFD analysis code to understand the heat transfer experimental results. The enhanced heat transfer coefficients on the smooth side wall at the rib's leading end were the same level as those on the rib‐roughened walls. © 2002 Scripta Technica, Heat Trans Asian Res, 31(2): 89–104, 2002; DOI 10.1002/htj.10018     

Thermal analysis of high‐power lithium‐ion battery packs using flow network approach

High‐power applications of lithium‐ion batteries require efficient thermal management systems. In this work, a lumped capacitance heat transfer model is developed in conjunction with a flow network approach to study performance of a commercial‐size lithium‐ion battery pack, under various design and operating conditions of a thermal management system. In order to assess the battery thermal management system, capabilities of air, silicone oil, and water are examined as three potential coolant fluids. Different flow configurations are considered, and temperature dispersions, cell‐averaged voltage distributions, and parasitic losses due to the fan/pump power demand are calculated. It is found that application of a coolant with an appropriate viscosity and heat capacity, such as water, in conjunction with a flow configuration with more than one inlet will result in uniform temperature and voltage distributions in the battery pack while keeping the power requirement at low, acceptable levels. Simulation results are presented and compared with literature for model validation and to show the superior capability of the proposed battery pack design methodology. Copyright © 2014 John Wiley & Sons, Ltd.     

Sub-channel-based multiscale thermal-hydraulic analysis of a medium-power,lead-cooled fast reactor

The lead-cooled fast reactor (LFR) offers enhanced safety and reliability with the fine properties of liquid lead and lead alloy. To study accurately the thermal characteristics of fast reactors, the multiscale thermal-hydraulic coupling simulation is an effective way. Multiscale coupling based on the sub-channel code has evident advantages on the analysis of fuel assemblies. In this study, a multiscale thermal-hydraulic analysis of a forced-circulation, medium-power LFR under steady-state and transient conditions is performed with the system code ATHLET and sub-channel code KMC-SUBtraC which was developed based on the previous version by modifying the pressure drop correlations and adding the assembly-level calculation. The codes are one-way-coupled, with good efficiency and precision. Transient verification of the sub-channel code is conducted with the CFD code. In the steady-state analysis of M²LFR-1000, mass flow and temperature distributions of the assemblies, sub-channels, and fuel rods in the hottest assembly are analyzed and the safety performance is investigated. In the transient analysis, two typical DECs (unprotected overpower transient and ULOF+ULOHS) are simulated and the multiscale thermal-hydraulic characteristics are analyzed. With the negative reactivity feedback, the variations of the temperatures of the coolant and fuel rods are within the safe limits, which shows the inherent safety of the reactor. And the results indicate that the loss of primary flow could increase the risk of cladding corrosion.