Thermal–hydraulic modeling of water/Al2O3 nanofluid as the coolant in annular fuels for a typical VVER-1000 core

In this paper, a thermal–hydraulic analysis of nanofluid as the coolant is performed in a typical VVER-1000 reactor with internally and externally cooled annular fuel. The fuel assembly for annular case with 8 × 8 arrays is considered for annular pin configuration. The considered nanofluid is a mixture composed of water and particles of Al₂O₃ with various volume percentages. The fuel rod is modeled using a CFD code. To validate the calculated results, the present results of solid fuel with nanofluid and pure water are compared with other studies which have been done with visual FORTRAN language, DRAGON/DONJON code, COBRA-EN code and the mentioned analytical approaches have been validated by comparing with the final safety analysis report (FSAR). The comparison of the calculated results shows that the results are in good agreement with other studies. Thus, the accuracy of the validation is satisfactory. Moreover, the temperature distributions of the fuel, clad and coolant are described for water/Al₂O₃ nanofluid in solid fuel and annular fuel. It is observed that as the concentration of Al₂O₃ nanoparticles increases, due to higher heat transfer coefficient of Al₂O₃ nanofluid, the temperature of the coolant is increased and the central fuel temperature is reduced. Thus, it improves margin from peak fuel temperature to melting. Finally, it is illustrated the use of the annular fuel instead of solid fuel in core of the reactor, security and efficiency of the nuclear power plant will be increased.     

基于MCNP和ORIGEN的熔盐快堆燃耗分析计算

熔盐堆是6种第4代先进核能系统中唯一使用液态燃料设计的反应堆型,其堆芯一回路中循环流动的熔盐既是燃料,也是冷却剂。这一特征在省去燃料元件加工制造步骤的同时,也使得熔盐堆能进行在线处理和在线添料的操作。因此,传统固态反应堆燃耗分析程序不再适用于熔盐堆。本文以熔盐快堆(MSFR)为分析对象,基于物理分析程序MCORE(MCNP+ORIGEN),将上述熔盐堆特点考虑进去,开发出能进行熔盐堆燃耗分析的MCORE-MS。初步分析表明,233 U-started模式下,熔盐在线处理可有效降低堆芯熔盐中裂变产物的含量,提高中子经济性。MSFR运行过程中能够一直保持负的温度反应性系数。     

Performance of TRISO particles fueled with plutonium and minor actinides in a PBMR-400 core design

The TRISO particle design of high temperature reactors fueled with plutonium (Pu) and/or minor actinides (MAs) is investigated by calculating the failure fraction of TRISO particles during irradiation. For this purpose, a fuel depletion, neutronics and thermal-hydraulics code system, which delivers the fuel temperature, fast neutron flux and power density profiles, is coupled to an analytical stress analysis code. The latter is being further developed for the calculation of a reliable and realistic failure fraction. The code system has been applied to a PBMR-400 design containing TRISO particles fueled with 1st and 2nd generation plutonium and with a target burn-up of 700 and 600 MWd/kgHM, respectively. It is shown that the pebble-bed type high temperature reactor under consideration is a promising option for burning Pu and MAs if very high burn-ups can be achieved. The TRISO particle failure fraction is also calculated for both Pu and MA fuels, and compared to U-based fuel. It is shown by the present stress analysis code that the Pu-based fuel particles need a better design and this has been achieved for the MA-based fuel, in which helium gas atoms have a significant contribution to the buffer pressure.     

Thermo-fluid investigation on a double-side-cooled annular fuel for the prismatic very high temperature gas-cooled reactor

One of the key issues in the development of the very high temperature gas-cooled reactor (VHTR) is an undesirably high temperature of the nuclear fuels. An increased reactor coolant outlet temperature directly drives an increase of the nuclear fuel temperature. Therefore, a special measure has to be taken to overcome the issue of a fuel temperature in the VHTR. In this paper, a double-side-cooled annular fuel concept for a prismatic type reactor is proposed to solve the fuel temperature issue. A detailed thermo-fluid analysis using a computational fluid dynamics (CFD) code was carried out to investigate the thermo-fluid performances of the proposed fuel design. The CFD results show that the proposed design has superior thermo-fluid characteristics to the existing prismatic fuel assembly designs.     

Neutronic feasibility design of a small long-life HTR

Small high temperature gas-cooled reactors (HTRs) have the advantages of transportability, modular construction and flexible site selection. This paper presents the neutronic feasibility design of a 20 MWth U-Battery, which is a long-life block-type HTR. Key design parameters and possible reactor core configurations of the U-Battery were investigated by SCALE 5.1. The design parameters analyzed include fuel enrichment, the packing fraction of TRISO particles, the radii of fuel compacts and kernels, and the thicknesses of top and bottom reflectors. Possible reactor core configurations investigated include five cylindrical, two annular and four scatter reactor cores for the U-Battery. The neutronic design shows that the 20 MWth U-Battery with a 10-year lifetime is feasible using less than 20% enriched uranium, while the negative values of the temperature coefficients of reactivity partly ensure the inherent safety of the U-Battery. The higher the fuel enrichment and the packing fraction of TRISO particles are, the lower the reactivity swing during 10 years will be. There is an optimum radius of fuel kernels for each value of the fuel compact design parameter (i.e., radius) and a specific fuel lifetime. Moreover, the radius of fuel kernels has a small influence on the infinite multiplication factor of a typical fuel block in the range of 0.2–0.25 mm, when the radius of fuel compacts is 0.6225 cm and the lifetime of the fuel block is 10 years. The comparison of the cylindrical reactor cores with the non-cylindrical ones shows that neutron under-moderation is a basic neutronic characteristic of the reactor core of the U-Battery. Increasing neutron moderation by replacing fuel blocks with graphite blocks and dispersing the graphite blocks in the reactor core are two effective ways to increase the fuel burnup and lifetime of the U-Battery. Water or steam ingress may induce positive reactivity ranging from 0 to 0.16 Δk/k, which further demonstrates that the U-Battery is under-moderated.     

CATHENA: A thermalhydraulic code for CANDU analysis

This paper describes the Canadian algorithm for thermal hydraulic network analysis (CATHENA) transient, thermalhydraulics code developed for the analysis of postulated upset conditions in CANDU®¹ reactors. The core of a CANDU reactor consists of a large number of horizontal pressure tubes containing fuel bundles. As a result of the unique design of the CANDU reactor, the CATHENA thermalhydraulic code has been developed with a number of unique modelling capabilities. The code uses a one-dimensional, two-fluid, nonequilibrium representation of two-phase flow. Some of the unique features of the CATHENA code are the one-step semi-implicit numerical method used and the solid heat transfer modelling capability that allows horizontal fuel bundles to be represented in detail. The code has been used in the design and analysis of CANDU-3, CANDU-6 and CANDU-9 reactors. The code has also been used for the design and analysis of the multiple applied lattice experimental (MAPLE) class of reactors and for the analysis of thermalhydraulic experimental programs conducted by Atomic Energy of Canada Limited (AECL).     

Two safety coefficients for a typical 13 × 13 annular fuel assembly

Recently, annular fuel rods are proposed for both PWRs and BWRs to achieve more average power density and therefore many neutronic as well as thermal-hydraulics calculations have been made to find more performance in the future reactors. Also, some safety margins are studied for the proposed uranium-nitride (and/or other fissile materials) annular pins. Our aim herein is to study two important safety coefficient of the annular fuel core. These are “prompt reactivity” and “power coefficients”, where all investigations are made using MCNP-5 code. Also, a thermal resistance model for annular fuel heat transferring were given and then we have calculated its (a case) thermal resistance numerically.     

基于CMS程序包的环形燃料堆芯物理计算分析

为研究在大型商用压水堆中采用环形燃料元件的可能性,需分析环形燃料的堆芯物理性能。本文研究了CASMO5程序计算环形燃料组件物理参数出现偏差的原因及其处理方法,分析了4组环形燃料先导组件加入秦山二期核电站平衡循环堆芯之后的堆芯物理参数。计算结果表明,装载的环形燃料先导组件对堆芯物理性能影响较小,基于CMS程序包开展环形燃料堆芯物理性能分析计算是可行的。     

Physics design of experimental metal fuelled fast reactor cores for full scale demonstration

Fast breeder reactors based on metal fuel are planned to be in operation for the year beyond 2025 to meet the growing energy demand in India. A road map is laid towards the development of technologies required for launching 1000 MWe commercial metal breeder reactors with closed fuel cycle. Construction of a test reactor with metallic fuel is also envisaged to provide full-scale testing of fuel sub-assemblies planned for a commercial power reactor. Physics design studies have been carried out to arrive at a core configuration for this experimental facility. The aim of this study is to find out minimum power of the core to meet the requirements of safety as well as full-scale demonstration. In addition, fuel sustainability is also a consideration in the design. Two types of metallic fuel pins, viz. a sodium bonded ternary (U-Pu-6% Zr) alloy and a mechanically bonded binary (U-Pu) alloy with 125 μm thickness zirconium liner, are considered for this study. Using the European fast reactor neutronics code system, ERANOS 2.1, four metallic fast reactor cores are optimized and estimated their important steady state parameters. The ABBN-93 system is also used for estimating the important safety parameters. Minimum achievable power from the converter metallic core is 220 MWt. A 320 MWt self-sustaining breeder metal core is recommended for the test facility.     

The feasibility of cooling heavy-water reactors with supercritical fluids

It appears technically feasible to use supercritical carbon dioxide as a coolant for a CANDU-type reactor. A new supercritical loop is proposed in which the reactor is cooled by a single-phase fluid pumped in a high density liquid-like state. The supercritical fluid-cooled reactor has the advantage of gas-cooled reactors of avoiding dryout, and of liquid-cooled reactors of low coolant-circulation power. By eliminating dryout, the maximum operating temperature of the fuel sheath can be increased to 1021°F (550°C) for existing Canadian fuel bundles, with a coolant exit temperature of 855°F (458°C) producing steam comparable to that of conventional fossil-fuel plants. Since the reactor coolant exit temperature from the steam generator may be as high as 280°F (138°C) low-pressure steam may also be produced. A new dual-reheat cycle is proposed with an ideal overall plant efficiency of 33%, comparable to the Pickering generating station.     

Subchannel analysis of fuel assemblies of a lead–alloy cooled fast reactor

Lead–alloy cooled fast reactor is one of the six Gen-IV reactors. It has many attractive features such as excellent natural circulation performance, better shielding against gamma rays or energetic neutrons and potentially reduced capital costs. A natural circulation lead–alloy cooled fast reactor with 10 MWth is under design in China (hereafter called LFR-10MW). Fuel assemblies thermal hydraulic analysis is of vital importance for a successful design. A subchannel analysis code with flow distribution model was used to carry out the thermal hydraulic analysis. This work briefly gave the thermal-hydraulic design for the LFR-10MW and analyzed the thermal-hydraulic characteristics under steady-state condition using the subchannel analysis code. Whole core analysis was performed to locate the hottest fuel assembly using the code. The hottest fuel assembly was analyzed to obtain the cladding temperature, fuel temperature and coolant velocity. The maximum cladding temperature, the maximum fuel center temperature and the maximum coolant velocity are all below the design constraints. These results imply that the thermal-hydraulic design of LFR-10MW is feasible.     

Static design of a liquid-salt-cooled pebble bed reactor (LSPBR)

A renewed interest has been raised for liquid-salt-cooled nuclear reactors. The excellent heat transfer properties of liquid-salt coolants provide several benefits, like lower fuel temperatures, higher average coolant temperature, increased core power density and better decay heat removal, and thus higher achievable core power. In order to benefit from the on-line refueling capability of a pebble bed reactor, the liquid salt pebble bed reactor (LSPBR) is proposed. This is a high temperature pebble bed reactor with a fuel design similar to existing HTRs, but using a liquid-salt as coolant. In this paper, the selection criteria for the liquid-salt coolant are described. Based on its neutronic properties, LiF–BeF₂ (flibe) was selected for the LSPBR. Two designs of the LSPBR were considered: a cylindrical core and an annular core with a graphite inner reflector. Coupled neutronic thermal-hydraulic calculations were performed to obtain the steady state power distribution and the corresponding fuel temperature distribution. Calculations were performed to investigate the decay heat removal capability in a protected loss-of-forced cooling accident. The maximum allowable power that can be produced with the LSPBR is hereby determined.     

Visual Study of Flow Behavior in Upper Plenum during Simulated Reflood Phase of PWR-LOCA

Fuel temperature is an important parameter in reactor safety. However, temperatures of fuel sub-assemblies in power reactors usually cannot be directly measured. A means therefore has been devised to use heat removed by coolant from fuel sub-assemblies following a reactor trip to estimate the pre-trip average fuel temperatures and fuel-to-coolant heat transfer coefficients. Sub-assembly coolant mass flow and outlet temperature measurements are used to determine the heat removed. After accounting for the contribution from release of stored heat from non-fuel components and from decay heat, the contribution due to release of fuel-stored heat alone during the trip transient can be inferred. Average fuel temperatures and fuel-to-coolant heat transfer coefficients of the FBR “MONJU” at 45% thermal power were estimated using this energy balance approach. The trip-test data derived estimates compared favorably with plant simulation code calculated values.     

CANDU fuel management optimization using alternative gradient methods

CANDU fuel management can be optimized for efficient reactor operation and reduced fueling costs. The quasi-linear programming approach developed in the OPTEX code for CANDU fuel management optimization has been implemented in the multipurpose multigroup diffusion code DONJON. With new reactor designs and requirements for advanced reactors, alternative gradient methods are presented and tested to address more complex CANDU fuel management problems.     

高通量研究堆堆芯热工水力分析程序THAS-PC4

针对高通量研究堆的热工水力特点和设计准则开发了堆芯子通道分析微机程序ＴＨＡＳ－ＰＣ４。该程序能够计算稳态和瞬态工况下堆芯内冷却剂的流量、温度、密度和压力等参量的分布以及燃料元件的温度场。它还可以确定临界热流密度和偏离泡核沸腾比（ＤＮＢＲ），泡核沸腾开始时的壁面温度以及流动开始不稳定时的热流密度等，以评价反应堆的安全性。     

双面慢化环形燃料反应堆物理实验

环形燃料零功率反应堆是首个双面慢化环形燃料作为核燃料的反应堆。本文采用周期法、落棒法获取环形燃料零功率反应堆的临界参数、控制棒价值、元件价值、含Gd元件的反应性效应等关键参数,对环形燃料零功率反应堆的物理性能进行实验研究,验证环形燃料反应堆堆芯物理设计计算程序。结果表明：根据外推过程确定堆芯临界装载环形燃料元件96根,实心燃料元件172根,此时k_eff为1.000 40,堆芯调节棒价值为-247.5 pcm,安全棒价值为-1 358.4 pcm;元件价值与理论值平均偏差为1.3 pcm,含Gd元件反应性效应与理论值平均相对偏差为8.8%。本文结果为环形燃料的工程化设计程序提供关键数据支撑。     

Irradiation swelling, creep and thermal-stress analysis of LWR fuel elements, computer code isune-2

The irradiation swelling, creep, and thermal-stress analysis of light-water reactor (LWR) oxide (UO₂) fuel elements is analysed. The analysis is based on the basic physical and mathematical assumptions and the experimental data of the fuel and cladding (or canning) materials. In the analysis, the nuclear, physical, metallurgical, and thermo-mechanical properties of the fuel and cladding materials under irradiation environment are examined carefully. The objectives of the paper are mainly (1) to formulate and carry out the irradiation swelling, irradiation creep, and thermal-stress analysis of fuel elements for LWR power reactors, and (2) to develop a computer code which will facilitate the computations for fuel element design, safety analysis, and economic optimization of the power reactors. In a general procedure of the analysis, the irradiation swelling, irradiation creep, temperature distribution, etc. in the fuel and cladding of the oxide fuel elements during the reactor in operation are studied. Some theoretical models and empirical relations (on the basis of accepted experimental data) for irradiation swelling and creep in the fuel and irradiation creep in cladding materials are postulated and developed. Some analytical and empirical relations (based on test results) for heat generation and temperature distribution in the fuel during fuel restructuring are derived. The fuel restructure is, in general, divided into the central void, columnar grain, equiaxed grain, and unaffected grain zones (or regions) after a sufficiently long period for the fuel elements to be irradiated (or operated). From these relations derived for irradiation swelling, irradiation creep, and temperature distribution in the fuel and cladding, together with the well-known strain-stress, incompressibility, compatibility, and stress equilibrium equations, the irradiation swelling, creep, and thermal-stress analysis for the LWR fuel elements can be carried out.From the analytical results obtained, a computer code, ISUNE-2 (which is in the sequence of computer code ISUNE-1 and -1A developed and used previously for liquid-metal fast breeder reactor fuel element design and safety and economic analysis), can be developed. With some reliable experimental data (measured during fuel elements in operation) as input, the computer code may predict various cases of LWR (oxide or carbide) fuel elements in operation. The general scope and resulting contribution of this paper is to provide a realistic analysis and a reliable operating LWR fuel element code for use by nuclear power utilities to predict the fuel element behavior in power reactors. The fuel element design, safety analysis, and economic optimization depend largely on the fuel element behavior in the power reactors.     

Comparative study of fast critical burner reactors and subcritical accelerator driven systems and the impact on transuranics inventory in a regional fuel cycle

In the frame of Partitioning and Transmutation (P&T) strategies, many solutions have been proposed in order to burn transuranics (TRU) discharged from conventional thermal reactors in fast reactor systems. This is due to the favourable feature of neutron fission to capture cross section ratio in a fast neutron spectrum for most TRU. However the majority of studies performed use the Accelerator Driven Systems (ADS), due to their potential flexibility to utilize various fuel types, loaded with significant amounts of TRU having very different Minor Actinides (MA) over Pu ratios. Recently the potential of low conversion ratio critical fast reactors has been rediscovered, with very attractive burning capabilities. In the present paper the burning performances of two systems are directly compared: a sodium cooled critical fast reactor with a low conversion ratio, and the European lead cooled subcritical ADS-EFIT reactor loaded with fertile-free fuel. Comparison is done for characteristics of both the intrinsic core and the regional fuel cycle within a European double-strata scenario. Results of the simulations, obtained by use of French COSI6 code, show comparable performance and confirm that in a double strata fuel cycle the same goals could be achieved by deploying dedicated fast critical or ADS-EFIT type reactors. However the critical fast burner reactor fleet requires ∼30-40% higher installed power then the ADS-EFIT one. Therefore full comparative assessment and ranking can be done only by a parametric sensitivity study of both the fuel cycle and the electricity generating costs.     

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

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

IAEA GT-MHR benchmark calculations by using the HELIOS/MASTER physics analysis procedure and the MCNP Monte Carlo code

The IAEA's gas-cooled reactor program has coordinated international cooperation for an evaluation of a high temperature gas-cooled reactor's performance, which includes a validation of the physics analysis codes and the performance models for the proposed GT-MHR. This benchmark problem consists of the pin and block calculations and the reactor physics of the control rod worth for the GT-MHR with a weapon grade plutonium fuel. Benchmark analysis has been performed by using the HELIOS/MASTER deterministic code package and the MCNP Monte Carlo code. The deterministic code package adopts a conventional 2-step procedure in which a few group constants are generated by a transport lattice calculation, and the reactor physics analysis is performed by a 3-dimensional diffusion calculation.In order to solve particular modeling issues in GT-MHR, recently developed technologies were utilized and new analysis procedure was devised. Double heterogeneity effect could be covered by using the reactivity-equivalent physical transformation (RPT) method. Strong core–reflector interaction could be resolved by applying an equivalence theory to the generation of the reflector cross sections. In order to accurately handle with very large control rods which are asymmetrically located in a fuel and a reflector block, the surface dependent discontinuity factors (SDFs) were considered in applying an equivalence theory. A new method has been devised to consider SDFs without any modification of the nodal solver in MASTER.All computational results of the HELIOS/MASTER code package were compared with those of MCNP. The multiplication factors of HELIOS for the pin cells are in very good agreement with those of MCNP to within a maximum error of 693 pcm Δρ. The maximum differences of the multiplication factors for the fuel blocks are about 457 pcm Δρ and the control rod worths of HELIOS are consistent with those of MCNP to within a maximum error of 3.09%. On considering a SDF in the core calculations, the maximum differences of the control rod worths are significantly decreased to be 7.7% from 21.5%. It is showed that there are good consistencies between the deterministic code package and the Monte Carlo code from the results of these benchmark calculations. Therefore, the HELIOS/MASTER 2-step procedure can be used as a standard reactor physics analysis tool for a prismatic VHTR.