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.     

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.     

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.     

Thermoelectric characteristics analysis of thermionic space nuclear power reactor

A thermoelectric analysis of space nuclear power reactor is of great importance to the space reactors development. In this paper, the reactor core model of TOPAZ-II designed by Soviet Union, including neutronics model, thermal-hydraulic model, and electrical circuit model, is established based on reasonable assumptions. A system analysis code is developed to analyze the thermoelectric characteristics of the TOPAZ-II under the condition of steady-state operation, start-up procedure, power change mode, and reactor shutdown. The code has been benchmarked with experimental data, and the maximum relative error is 16.6%. Numerical results show that for the steady state, the simulated electrical power is 5.2 kW within the design value range. For transient state, the reactor thermoelectric characteristics are mainly affected by the electrode temperature: (a) For the start-up procedure, when the emitter temperature increases above 1700 K, electrical system begins to work and reach full power in 5 minutes. (b) For power change mode, the emitter temperature decreases by 5%, while the electrical power decreases by 67%; (c) For reactor shutdown, electrical power reduces to 0 kW as the emitter temperature decreases to 1400 K in 100 seconds. This study provides valuable theoretical supports for the design and analysis of the thermionic space nuclear power reactor.     

Dynamic simulation of a space gas-cooled reactor power system with a closed Brayton cycle

Space nuclear reactor power (SNRP) using a gas-cooled reactor (GCR) and a closed Brayton cycle (CBC) is the ideal choice for future high-power space missions. To investigate the safety characteristics and develop the control strategies for gas-cooled SNRP, transient models for GCR, energy conversion unit, pipes, heat exchangers, pump and heat pipe radiator are established and a system analysis code is developed in this paper. Then, analyses of several operation conditions are performed using this code. In full-power steady-state operation, the core hot spot of 1293 K occurs near the upper part of the core. If 0.4 $ reactivity is introduced into the core, the maximum temperature that the fuel can reach is 2059 K, which is 914 K lower than the fuel melting point. The system finally has the ability to achieve a new steady-state with a higher reactor power. When the GCR is shut down in an emergency, the residual heat of the reactor can be removed through the conduction of the core and radiation heat transfer. The results indicate that the designed GCR is inherently safe owing to its negative reactivity feedback and passive decay heat removal. This paper may provide valuable references for safety design and analysis of the gas-cooled SNRP coupled with CBC.     

Mathematical modeling and transient thermal-hydraulic analysis of boiling water reactors

This paper presents the development and practical application of a dynamic mathematical model for the simulation of the transient thermal-hydraulic analysis of Boiling Water Reactors (BWRs). In the development of the mathematical model, the reactor core thermal-hydraulic model for the coolant flow along the fuel pin channel is based on the moving boundary concept. Some results from the transient analysis are examined to evaluate the thermal-hydraulic performance of the mathematical model using moving boundaries. These analyses show that the predictions of the mathematical model based on the moving boundary formulation provide a suitable analytical tool for the transient thermal-hydraulic analysis of BWRs.     

Sub‐channel analysis for Pb‐Bi‐cooled direct contact boiling water fast reactor

Pb‐Bi‐cooled direct contact boiling water fast reactor (PBWFR) featured with a direct‐contact heat exchanger between lead‐bismuth eutectic coolant and water could significantly simplify the primary system and enhance the natural circulation capability, meeting the potential needs for small modular reactor design. It is of great importance to conduct thermal‐hydraulic analysis of the PBWFR core in detail. In this paper, a self‐developed SUB‐channel AnalysiS code SUBAS is adopted to study the thermal hydraulic characteristics of the PBWFR core. The fidelity and the reliability of the code have been preliminarily benchmarked. With SUBAS, the space grid is studied to figure out its impact on the temperature and flow distributions in each sub‐channel. Besides, the application of space grids would increase the pressure drops and decreases the cross flow between adjacent sub‐channels. To study the transient performance of the PBWFR core, the power transient and the inlet blockage accident are calculated by SUBAS. The results of the power transient show the cross‐flow effect would be weakened in the sub‐channel which has higher coolant temperature and larger mass flow rate. For the inlet blockage accident, the results indicate the influence of the small area blockage is relatively weak on the overall performance of the assembly but is significant on the local parameters. With consideration of time and space, the blockage influence only exists in a certain area. This research may provide contribution to the design of PBWFR.     

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.     

Development of a steady thermal-hydraulic analysis code for the China Advanced Research Reactor

A multi-channel model steady-state thermal-hydraulic analysis code was developed for the China Advanced Research Reactor (CARR). By simulating the whole reactor core, the detailed mass flow distribution in the core was obtained. The result shows that structure size plays the most important role in mass flow distribution, and the influence of core power could be neglected under single-phase flow. The temperature field of the fuel element under unsymmetrical cooling condition was also obtained, which is necessary for further study such as stress analysis, etc. of the fuel element. At the same time, considering the hot channel effect including engineering factor and nuclear factor, calculation of the mean and hot channel was carried out and it is proved that all thermal-hydraulic parameters satisfy the “Safety design regulation of CARR”. Translated from Atomic Energy Science and Technology, 2006, 40(1): 51–55 [译自: 原子能科学技术]     

The multiscale thermal-hydraulic simulation for nuclear reactors: A classification of the coupling approaches and a review of the coupled codes

Simulation is becoming increasingly important in the safety analysis of nuclear reactors nowadays. The physical phenomena in a nuclear power plant happen on three classified scales: system scale (phenomenon over the whole plant is concerned), component scale (phenomenon in specific component is concerned), and mesoscale (phenomenon in a small part of a component is concerned). Owing to the particular emphases, various codes are developed to simulate particular problems. System codes intend to predict the behavior of the whole power plant during normal or accidental phases (system scale). Subchannel codes are for core behavior predictions (component scale). CFD codes can simulate the thermal-hydraulic in a fixed part of the plant (mesoscale). Those codes are coupled together to better predict the conditions in a nuclear reactor in last the two decades, which is the multiscale thermal-hydraulic simulation approach for nuclear power systems. Diverse coupling approaches are developed and various coupling codes are implemented. This paper first proposes a classification of those approaches. It tells that a multiscale coupling is composed of five items: coupling architecture, operation mode, domain coupling, field mapping, and temporal coupling. Numbers of options are available for each item. For coupling architecture, it can be internal coupling, via-IO coupling, server-client, or serverless coupling. For operation mode, it can be either parallel or serial. For domain coupling, it can be either domain-decomposition or domain-overlapping coupling. For field mapping, it can be manual-definition, processed by user-developing toolkit, or handled by third-party libraries. For temporal coupling, it can be explicit coupling, semi-implicit coupling, or implicit coupling. An evaluation of the approaches is performed based on new-proposed criterion. A general review of the multiscale thermal-hydraulic coupled codes is made based on the classification. Especially, a review of the domain-overlapping approach is present considering it is the most promising but challenging method for multiscale thermal-hydraulic simulation.     

Assessment of transient characteristics of fast reactors and influences of minor actinides using neutron transport method

Partitioning and transmutation of the minor actinides (MAs) from nuclear power plants are of importance for the nuclear‐energy sustainable development. Fast reactors are applied to transmutation for the hard neutron spectrum and high neutron flux. However, the safety‐related neutronic parameters will become worse when large amounts of MAs are loaded. In this paper, transients of a 600 MWe sodium‐cooled fast reactor for MA transmutation are analyzed by using neutron transport simulation. The control rod withdrawal transients are calculated. Two cases are compared to investigate the influence of loading MAs into the fast reactor core. One is the common core loaded with mixed oxide (MOX) fuel, and the other one is the transmutation core loaded with MOX fuel and MAs. The results indicate that in order to apply similar operation criterion with the common core, the transmutation core with 6% weight fraction of MAs should be operated with more than 30% power reduction. In addition, the results of the transport‐based transient analysis and the point kinetics transient analysis are compared. There are noticeable differences between them, which indicate that the usual way based on the point kinetics calculation is not suitable well for simulating the control rod introduced transients. Copyright © 2017 John Wiley & Sons, Ltd.     

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).     

Predictability of safety analysis codes for departure from nucleate boiling in bundle for safety evaluation of massive hydrogen production systems

The advanced water reactor was indicated as a candidate for massive hydrogen production system using the water electrolysis method. In order to utilize the advanced water reactor system for hydrogen production, it is crucial to demonstrate the safety of the nuclear system during normal operations and accidents. Departure from nucleate boiling (DNB) is a critical phenomenon in the reactor core which should be addressed to demonstrate the integrity of the nuclear core during normal operations and accidents. Therefore, DNB has a particular importance to the reactor safety and precise prediction has been required for thermal-hydraulic analysis codes including subchannel and safety analysis codes. In this study it has been assessed the DNB prediction capability of thermal-hydraulic safety analysis codes used for the safety evaluation of nuclear reactor system against experimental data. For the assessment, thermal-hydraulic safety analysis codes, MARS-KS and TRACE, have been utilized. The DNB experiments conducted at the NUPEC experimental facility have been employed as a reference experiment for assessment. All experiments with bundle geometries under various steady-state conditions have been analyzed. The results show that both safety analysis codes generally predict the DNB power lower than the experimental database by 20% and the under-prediction occurs systematically with a linear characteristic. It is found that no significant difference in predictability of the DNB occurrence is observed between MARS-KS and TRACE. Therefore, it is concluded that both codes predict DNB conservatively, and MARS-KS and TRACE have almost identical predictability for the DNB occurrence.     

Determination of the liquid eutectic metal fuel dual fluid reactor (DFRm) design – steady state calculations

The dual fluid reactor (DFR) is a novel concept of a very high‐temperature (fast) reactor that falls off the classification of generation 4 international forum (GIF). DFR makes best of the two previous designs: molten salt reactor (MSR) and lead‐cooled fast reactor (LFR). In this paper, we present a new reactor design Dual Fluid Reactor metallic (DFRm) with the liquid eutectic U‐Cr metal fuel composition and the lead coolant of which general idea was patented recently. By performing the first steady state neutronic calculations for such a reactor (the neutron flux density as a function of energy, the burnup, the effective multiplication factor/reactivity), we show that this 250‐MWth reactor is critical, and that it can operate almost 20 years without refuelling. We also optimise the geometry (reflector thickness, fuel tubes pin pitch) with respect to the multiplication factor. The optimisation together with some other opportunities for the liquid metal fuel design (eg, the use of electromagnetic pumps to circulate the medium) allows DFRm to be of a small size. This rises economy of the construction as expressed nicely in terms of the energy return on invested (EROI) factor, which is even higher than for the molten salt fuel design (DFRs). Last but not least, we show that DFRm has all the (fuel, coolant, reflector) temperature coefficients negative, which is an important factor of the passive safety.     

Steady-state thermal-hydraulic analysis of SCWR assembly

Among the six gen-IV reactor concepts recommended by the gen-IV international forum (GIF), supercritical water-cooled reactor (SCWR), the only reactor with water as coolant, achieves a high thermal efficiency and, subsequently, has economic advantages over the existing reactors due to its high outlet temperature. A thermal-hydraulic analysis of the SCWR assembly is performed in this paper using the modified COBRA-IV code. Two approaches to reduce the hot channel factor are investigated: decreasing the moderator mass flow and increasing the thermal resistance between moderator channel and its adjacent sub-channels. It is shown that heat transfer deterioration cannot be avoided in SCWR fuel assembly. It is, therefore, highly required to calculate the cladding temperature accurately and to preserve the fuel rod cladding integrity under heat transfer deterioration conditions.     

Steady-state thermal-hydraulic analysis of SCWR assembly

Among the six gen-IV reactor concepts recommended by the gen-IV international forum (GIF), supercritical water-cooled reactor (SCWR), the only reactor with water as coolant, achieves a high thermal efficiency and, subsequently, has economic advantages over the existing reactors due to its high outlet temperature. A thermal-hydraulic analysis of the SCWR assembly is performed in this paper using the modified COBRA-IV code. Two approaches to reduce the hot channel factor are investigated: decreasing the moderator mass flow and increasing the thermal resistance between moderator channel and its adjacent sub-channels. It is shown that heat transfer deterioration cannot be avoided in SCWR fuel assembly. It is, therefore, highly required to calculate the cladding temperature accurately and to preserve the fuel rod cladding integrity under heat transfer deterioration conditions. __________ Translated from Nuclear Power Engineering, 2007, 28(5): 18–21, 58 [译自: 核动力工程]     

Generation-IV nuclear power: A review of the state of the science

The Generation-IV consortium seeks to develop a new generation of nuclear energy systems for commercial deployment by 2020–2030. These systems include both the reactors and their fuel-cycle facilities. The aim is to provide significant improvements in economics, safety, sustainability, and proliferation resistance. The systems selected for development are the very high-temperature gas-cooled reactor (VHTR), the sodium-cooled fast reactor (SFR), the gas-cooled fast reactor (GFR), the lead-cooled fast reactor (LFR), the molten salt reactor (MSR) and the super-critical water-cooled reactor (SCWR). UK organisations plan to contribute to the first three of these systems because of its existing capabilities and experience with gas-cooled systems, graphite cores, and SFRs. The science base for the VHTR and SFR systems is reasonably established, although there are gaps. For the VHTR, these include the performance of graphite at high neutron doses, and the performance of the fuel. For the SFR, the behaviour of fuels containing minor actinides, and processes for their recycling and refabrication into new fuel, must be established. The GFR presents many technical challenges, because it would need fuel and structural materials capable of withstanding extremes of fast neutron flux and high temperatures. Adequate heat removal from the core under fault conditions is likely to determine its feasibility.     

Assessment of boiler tubes overheating mechanisms during a postulated loss of feedwater accident

Due to their simplicity and passive nature, natural circulation loops have many industrial applications and being increasingly used in many innovative designs as the new generation of nuclear reactor cooling systems. Consequently, special attention is increasingly considered towards their safety issues. In the current framework, a safety aspect related to the loss of feedwater in an industrial D-type steam boiler is assessed. Indeed, loss of feedwater event occurs frequently during steam generator facilities lifetime. Under such conditions, the system integrity is in jeopardy and serious hazards and economic losses could occur. Up to now, specific numerical models have been used to simulate the thermal-hydraulic phenomena occurring under such circumstances. Such models are generally based upon simplified assumptions, and if complex model are performed, their applications are generally restricted to the facility they have been developed for. In the current framework, an attempt is made to apply a multipurpose best estimate (BE) thermal-hydraulic system code, namely, RELAP5/Mod3.2 to simulate steady-state and transient dynamic behaviors of a two-phase natural circulation steam boiler. For this purpose, two loss of feedwater scenarios, with and without the actuation of the boiler control system, have been performed. The analysis results show on one hand the achievement in applying RELAP5 for a conventional boiler system, and on the other hand the efficiency of the control systems to mitigate the accident consequences.     

Neutronics optimization and characterization of a long‐life SCO2‐cooled micro modular reactor

This paper presents a neutronics optimization study of a supercritical CO₂‐cooled micro modular reactor (MMR). The MMR is a fast‐spectrum reactor designed to be an extremely compact, integrated, and truck‐transportable reactor with 36.2‐MWth power and a 20‐year lifetime without refueling. The reactor uses a drum‐type primary control system and a single absorber rod located at the core center as the secondary ultimate shutdown system. In order to maximize the fuel inventory in a compact fast reactor, hexagonal fuel assemblies are adopted in this work. We compare two types of MMR: One is using U¹⁵N fuel, and the other one is based on UC fuel. In addition, the minimization of the core excess reactivity to less than 1 dollar is also achieved in this study by a unique application of a replaceable fixed absorber in order to enhance safety of the MMR core by preventing the possibility of a prompt criticality accident. Moreover, the required number of primary control drums is also reduced through minimization of the excess reactivity. Several important safety parameters such as control rod/drum worth, reactivity coefficients, and power peaking factors are also characterized as a function of core burnup. The neutronics analyses and depletion calculations are all performed using the continuous‐energy Monte Carlo Serpent code with the latest evaluated nuclear data file (ENDF/B‐VII.1) library. Copyright © 2016 John Wiley & Sons, Ltd.     

Development of effective thermal conductivity models for Reserve Shutdown Control fuel block of prismatic HTGR for hydrogen production

A hydrogen production system coupled to High Temperature Gas-cooled nuclear Reactor (HTGR) is considered to be one of the most promising ways for massive hydrogen production. For the reliability of the coupled system, the safety analysis on the HTGR is to be conducted by a system-scale analysis code. The system-scale analysis code adopts an effective thermal conductivity (ETC) model for a fuel block due to its complex geometry containing large number of coolant holes and nuclear fuel rods. The ETC of the fuel block is crucial to calculate the heat transfer inside the reactor core and prediction of thermal distribution over the reactor core is the most significant for the safety analysis of HTGR. Therefore, the verification of the ETC model that contributes to the prediction is essential. This ETC model based on Maxwell's theory shows an inaccurate prediction when the configuration of the composite materials is not homogeneous. Since the geometry of Reserve Shutdown Control (RSC) fuel block of HTGR is not homogeneous due to a large RSC hole, the ETC model for RSC fuel block should be developed to improve the accuracy and reliability of the reactor system analysis code. In this study, the two ETC models for the RSC fuel block have been developed by the thermal network modeling. Computational fluid dynamic simulations with a real geometry were performed to evaluate the accuracy of the ETC models for the RSC fuel block. The comparative result between CFD analysis and the ETC model shows that the newly developed model predicts the effective thermal conductivity of RSC fuel block more accurately than the previous model.