Design of a supercritical water-cooled reactor with a three-pass core arrangement

The Supercritical Water-cooled Reactor (SCWR) is one of the six concepts of the Generation IV International Forum. In Europe, investigations have been integrated into a joint research project, called High Performance Light Water Reactor (HPLWR). Due to the higher heat up within the core and a higher outlet temperature, a significant increase in turbine power and thermal efficiency of the plant can be expected.Besides the higher pressure and higher steam temperature, the design concept of this type of reactor differs significantly from a conventional LWR by a different core concept. In order to achieve the high outlet temperature of over 500 °C, a core with a three-step heat up and intermediate mixing is proposed to keep local cladding temperatures within today's material limits. A design for the reactor pressure vessel (RPV) and the internals has been worked out to incorporate a core arrangement with three passes. All components have been dimensioned following the safety standards of the nuclear safety standards commission in Germany. Additionally, a fuel assembly cluster with head and foot piece has been developed to facilitate the complex flow path for the multi-pass concept. The design of the internals and of the RPV is verified using mechanical or, in the case of large thermal deformations, combined mechanical and thermal stress analyses. Furthermore, the reactor design ensures that the total coolant flow path remains closed against leakage of colder moderator water even in case of large thermal expansions of the components. The design of the RPV and internals is now available for detailed analyses of the core and the reactor.     

三环路压水堆压力容器上腔室交混矩阵数值研究

使用STAR-CCM+软件对三环路压水堆压力容器上腔室流场进行了大规模、精细化三维数值模拟，并采用组分跟踪方法分别对157个燃料组件出口冷却剂流动进行计算，构造了一个具有3×157个元素的“上腔室交混矩阵”，用该矩阵即可定量、精确地描述冷却剂从堆芯流出后，经上腔室内交混并再分配到各热管道的复杂流动过程。研究发现堆芯流出的冷却剂在压力容器上腔室内的交混是并不充分的，径向上不同位置燃料组件流出的冷却剂会在上腔室同热管道的接口区域存在明显的对应关系，而燃料组件径向功率分布的差异必然导致热管道中冷却剂热分层现象的产生。      

On the accuracy of reactor physics calculations for square HPLWR fuel assemblies

Although the supercritical-pressure or high-performance light water reactor (HPLWR) concept is largely based on the well-established technological experience available with conventional light water reactors, there is still no consensus on various key design features such as an optimal layout for the fuel assembly. This results mainly from the very large density variations of supercritical-pressure water in the core, which render it difficult to ensure reliable values for parameters such as power peaking factors and reactivity worths. The present paper describes studies carried out to compare deterministic and Monte Carlo codes for analysing a representative square HPLWR lattice with uniform 5%-enriched UO₂ fuel. The main purpose has been to assess the prediction accuracies achievable for integral parameters such as the multiplication factor, control absorber effectiveness, moderator/coolant density reactivity feedback and pin power distributions. The results show good agreement between the deterministic and stochastic calculations for the unperturbed lattice. However, for certain perturbed situations involving, for example, local coolant density changes in the assembly or control absorber insertion, the observed discrepancies are large enough to question the basic viability of the reactor physics design, e.g. with respect to the thermal performance.     

Three pass core design proposal for a high performance light water reactor

The paper describes a novel core concept for a nuclear reactor cooled with supercritical water, in which the coolant is heated up from 280 °C at the reactor inlet to 500 °C at the outlet in four steps: a first heat-up step is provided by heat transfer from fuel assemblies to the moderator water in gaps and moderator boxes, a second step is foreseen in a central “evaporator” and two further steps in a first and a second superheater surrounding it. The coolant flow scheme includes upward and downward flow through the core with intermediate mixing in chambers above and below the core to eliminate hot streaks. A preliminary single channel analysis, concentrating on an average flow channel and on the hottest one only, indicates that such core design can match the limits of cladding materials available today. Even though the resultant pressure drop of the coolant will be higher than usual, it is expected that the assembly boxes can be designed with acceptable deformations.     

European supercritical water cooled reactor

The High Performance Light Water Reactor (HPLWR), how the European Supercritical Water Cooled Reactor is called, is a pressure vessel type reactor operated with supercritical water at 25 MPa feedwater pressure and 500 °C average core outlet temperature. It is designed and analyzed by a European consortium of 10 partners and 3 active supporters from 8 Euratom member states in the second phase of the HPLWR project. Most emphasis has been laid on a core with a thermal neutron spectrum, consisting of small fuel assemblies in boxes with 40 fuel pins each and a central water box to improve the neutron moderation despite the low coolant density. Peak cladding temperatures of the fuel rods have been minimized by heating up the coolant in three steps with intermediate coolant mixing. The containment design with its safety and residual heat removal systems is based on the latest boiling water reactor concept, but with different passive high pressure coolant injection systems to cause a forced convection through the core. The design concept of the steam cycle is indicating the envisaged efficiency increase to around 44%. Moreover, it provides the constraints to design the components of the balance of the plant. The project is accompanied by numerical studies of heat transfer of supercritical water in fuel assemblies and by material tests of candidate cladding alloys, performed by the consortium and supported by additional tests of the Joint Research Centre of the European Commission. Besides the scientific and technical progress, the HPLWR project turned out to be most successful in training the young generation of nuclear engineers in the technologies of light water reactors. More than 20 bachelor or master theses and more than 10 doctoral theses on HPLWR technologies have been submitted at partner organizations of this consortium since the start of this project.     

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

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

HPLWR equilibrium core design with the KARATE code system

The High Performance Light Water Reactor (HPLWR) is the European version of the various supercritical water cooled reactor proposals. The paper presents the activity of KFKI-AEKI in the field of neutronic core design within the framework of the "HPLWR Phase 2" FP-6 and the Hungarian “NUKENERG” projects. As the coolant density along the axial direction shows remarkable change, coupled neutronic-thermohydraulic calculations are essential which take into account the heating of moderator in the special water rods of the assemblies. A parametrized diffusion cross section library was prepared for the HPLWR assembly with the MULTICELL neutronic transport code. The parametrized cross sections are used by the KARATE program system, which was verified by comparative Monte Carlo calculations. Preliminary loadings of the HPLWR core were assessed, which contain insulated assemblies with Gd burnable absorbers. The fuel assemblies have radial and axial enrichment zoning to reduce hot spots.     

Mixing of cooling water in the mixing chambers of the HPLWR—High Performance Light Water Reactor

The High Performance Light Water Reactor (HPLWR), a supercritical water cooled reactor concept with multiple heat-up steps, requires efficient mixing of the coolant between these steps to minimize hot spots in the core. Analyzing and improving the mixing in the mixing chamber above the core, situated between evaporator and superheater assemblies, and below the core, between the first and second superheater, is one of the challenges in the design process of the HPLWR. Different measures to enhance mixing have been studied with CFD analyses, in which a new design approach has been applied to the upper mixing chamber. It simplifies the complex structures and takes the effects of the disregarded structures into account by introducing source terms into the momentum equations.     

Studies on supercritical water reactor fuel assemblies using the sub-channel code COBRA-EN

In the Generation IV International Forum (GIF) program, the supercritical water reactor (SCWR) concept is among the six innovative reactor types selected for development in the near future. In principle the higher efficiency and better economics make the SCWR concept competitive with the current reactor design. Due to different technical challenges that, however exist, fuel assembly design represents a crucial aspect for the success of this concept. In particular large density variations, low moderation, heat transfer enhancement and deterioration have a strong effect on the core design parameters. Only a few computational tools are currently able to perform sub-channel thermal-hydraulic analysis under supercritical water conditions. At JRC-IE the existing sub-channel code COBRA-EN has been improved to work above the critical pressure of water. The water properties package of the IAPWS Industrial Formulation 1997 was integrated in COBRA-EN to compute the Thermodynamic Properties of Water and Steam. New heat transfer and pressure drop correlations more indicated for the supercritical region of water have also been incorporated in the code. As part of the efforts to appraise the new code capabilities, a code assessment was carried out on the hexagonal fuel assembly of a fast supercritical water reactor. COBRA-EN was also applied in combination with the neutronic code MCNP to investigate on the use of hydride fuel in the HPLWR supercritical water fuel assembly. The results showed that COBRA-EN was able to reproduce the results of similar studies with acceptable accuracy. Future activities will focus on the validation of the code against experimental data and the implementation of new features (counter-current moderator channel, wall, and wire-wrap models).     

Development of a coupled neutronic/thermal-hydraulic tool with multi-scale capabilities and applications to HPLWR core analysis

The High Performance Light Water Reactor (HPLWR) is a thermal spectrum nuclear reactor cooled and moderated with light water operated at supercritical pressure. It is an innovative reactor concept, which requires developing and applying advanced analysis tools as described in the paper.The relevant water density reduction associated with the heat-up, together with the multi-pass core design, results in a pronounced coupling between neutronic and thermal-hydraulic analyses, which takes into account the strong natural influence of the in-core distribution of power generation and water properties. The neutron flux gradients within the multi-pass core, together with the pronounced dependence of water properties on the temperature, require to consider a fine spatial resolution in which the individual fuel pins are resolved to provide precise evaluation of the clad temperature, currently considered as one of the crucial design criteria. These goals have been achieved considering an advanced analysis method based on the usage of existing codes which have been coupled with developed interfaces.Initially neutronic and thermal-hydraulic full core calculations have been iterated until a consistent solution is found to determine the steady state full power condition of the HPLWR core. Results of few group neutronic analyses might be less reliable in case of HPLWR 3-pass core than for conventional LWRs because of considerable changes of the neutron spectrum within the core, hence 40 groups transport theory has been preferred to the usual 2 groups diffusion theory. Successively, with the usage of a developed pin-power reconstruction technique capable to account for the innovative fuel assembly design, sub-channel investigations of the individual fuel assemblies have been performed evaluating pin-wise clad temperatures. Obtained results will be discussed giving a detailed insight of the revolutionary HPLWR 3 pass core concept and understanding the physical reasons, which influence the local clad temperatures.The obtained results represent a new quality in core analyses, which takes into full consideration the coupling between neutronics and thermal-hydraulics as well as the spatial effects of the fuel assembly heterogeneity in determining the local pin-power and the associated maximum clad temperature.     

Concept of a pressurized water reactor cooled with supercritical water in the primary loop

A novel concept of a pressurized water reactor with a primary loop cooled with supercritical water is introduced and analyzed in this work. A steam cycle analysis has been performed to illustrate the advantages of such a nuclear power plant with respect to specific power and thermal efficiency. Moreover, a reactor pressure vessel concept including its internals and a suitable core and fuel assembly design are presented overcoming the problems, which arise due to the high heat up of the coolant and the density change involved with it. The core power and coolant density distributions are predicted with coupled neutronic and thermal-hydraulic analyses. The method features the definition of inlet orifices for coolant mass flow adjustment within the core as well as an additional tool for the interpolation of local pin power data. The latter one has been used for a successive sub-channel analysis of the hottest fuel assembly of the core, which provides a more detailed spatial resolution and thus predicts peak cladding temperatures, the maximum linear pin power of fuel pins, and maximum fuel temperatures. It can be shown that maximum temperatures of claddings and fuel are well below the material limits. Thanks to an average core exit temperature below the pseudo-critical temperature, the core concept leaves enough margin for additional uncertainties and allowances for operation.     

Design and analysis of a thermal core for a high performance light water reactor

The High Performance Light Water Reactor is a Generation IV light water reactor concept, operated at a supercritical pressure of 25 MPa with a core outlet temperature of 500 °C. A thermal core design for this reactor has been worked out by a consortium of Euratom member states within the 6th European Framework Program. Aiming at peak cladding temperatures of less than 630 °C, including uncertainties and allowances for operation, the coolant is heated up in three steps with intermediate coolant mixing to eliminate hot streaks. Different from conventional reactors, the radial power profile is intended to be non-uniform, with the highest power in the first heat-up step in the core center and the lowest power in the second superheater step to result in the same peak cladding temperatures in each region. The concept has been studied with neutronic, thermal-hydraulic and structural analyses to assess its feasibility. Coupled neutronic/thermal-hydraulic analyses are defining the initial distribution of enrichment, control rod positions and the use of burnable poisons. Sub-channel analyses predict the coolant mixing inside assemblies, and a porous media approach simulates the flow of moderator water between assembly boxes. Finally, structural analyses of the assembly boxes are needed to minimize deformations during operation. Even though the core design cannot yet considered to be final, this state of the art review shall summarize the progress achieved so far and outline the remaining challenges.     

混合能谱超临界水堆堆芯设计分析

提出了一种新型的超临界水堆概念设计:混合能谱超临界水堆,它包括慢谱区和快谱区两部分.其慢谱区燃料组件采用双排燃料组件,快谱区采用简单的正方形栅元燃料组件.慢谱区与快谱区的燃料组件都采用同向流动方式来简化堆芯设计.慢谱区的冷却剂出口温度远低于整个堆芯的出口温度,这大大降低了慢谱区包壳的温度峰值.此外,由于快谱区冷却剂密度很小,流速很高,故可采用较大的栅元结构,这有效地降低了包壳周向局部传热不均匀性.所以混合堆在充分继承慢谱、快谱堆芯优点的基础上,弥补两者的不足.     

Thermal and stability considerations for a supercritical water-cooled fast reactor with downward-flow channels during power-raising phase of plant startup

This paper describes study on the procedure of raising the reactor thermal power and the reactor coolant flow rate during the power-raising phase of plant startup for the supercritical water-cooled fast reactor (SWFR), which is selected as one of the Generation IV reactor concepts. Since part of the seed fuel assemblies and all the blanket fuel assemblies of the SWFR are cooled by downward flow, the feedwater from the reactor vessel inlet nozzle to the mixing plenum located below the core is distributed among these fuel assemblies and the downcomer. The flow rate distribution as the function of both the reactor thermal power and the feedwater flow rate, which are the design parameters for the power-raising phase, is obtained by the thermal hydraulic calculations. Based on the flow rate distribution, thermal analyses and thermal-hydraulic stability analyses are carried out in order to obtain the available region of the reactor thermal power and the feedwater flow rate for the power-raising phase. The criteria for the “available” region are the maximum cladding surface temperature (MCST) and the decay ratio of thermal-hydraulic stability in three “hot” channels; two seed assemblies with upward/downward flow and a blanket assembly. The effects of various heat transfer correlations and axial power distributions are also studied.     

Experimental study of the coupled thermo-hydraulic–neutronic stability of a natural circulation HPLWR

The HPLWR (high performance light water reactor) is the European concept design for a SCWR (supercritical water reactor). This unique reactor design consists of a three pass core with intermediate mixing plena. As the supercritical water passes through the core, it experiences a significant density reduction. This large change in density could be used as the driving force for natural circulation of the coolant, adding an inherent safety feature to this concept design. The idea of natural circulation has been explored in the past for boiling water reactors (BWR). From those studies, it is known that the different feedback mechanisms can trigger flow instabilities. These can be purely thermo-hydraulic (driven by the friction – mass flow rate or gravity – mass flow rate feedback of the system), or they can be coupled thermo-hydraulic–neutronic (driven by the coupling between friction, mass flow rate and power production). The goal of this study is to explore the stability of a natural circulation HPLWR considering the thermo-hydraulic–neutronic feedback. This was done through a unique experimental facility, DeLight, which is a scaled model of the HPLWR using Freon R23 as a scaling fluid. An artificial neutronic feedback was incorporated into the system based on the average measured density. To model the heat transfer dynamics in the rods, a simple first order model was used with a fixed time constant of 6 s. The results include the measurements of the varying decay ratio (DR) and frequency over a wide range of operating conditions. A clear instability zone was found within the stability plane, which seems to be similar to that of a BWR. Experimental data on the stability of a supercritical loop is rare in open literature, and these data could serve as an important benchmark tool for existing codes and models.     

Pressurized water reactor (PWR) hot-leg streaming: Part 1: Computational fluid dynamics (CFD) simulations

A thermo-hydraulic analysis model was developed to analyze thermal stratification phenomena observed in the hot-legs of pressurized water reactors (PWR). The model uses VIPREW code to determine the flow field and temperature distribution in the reactor fuel region. The temperature readings from the thermal couples located at the exit of the reactor core were used to compare with the VIPREW computed results. The predicted values agree well with the measurements. The VIPREW results are then used as the boundary conditions for the CFD analysis. The CFD computational domain includes the upper plenum and hot-legs and the fifty two (52) control rod guiding tubes to properly include the additional obstructions imposed to the fluid. Different fuel loading patterns were studied to investigate the effects of different power distribution and fuel channel exit water temperature on hot-leg thermal stratification magnitude. The analysis results show that the 52 control rod guide tubes have major contribution to the mixing effect in the upper plenum. The sudden expansion of the cross sectional area in the upper plenum leads to the formation of recirculation vortex that prolongs the duration of coolant in the reactor vessel. The hotter coolant from the center portion tends to flow upwards to the top before exiting at the upper portion of the hot-leg pipes. It leads to higher temperature in the upper portion of the hot-legs. Water from the cooler outer fuel channels tends to trap in the recirculation region before exiting from the lower portion of the hot-legs.     

International Solvent Extraction Conference ISEC ′90

The supercritical-water-cooled power reactor (SCPR) is expected to reduce power costs compared with those of current LWRs because of its high thermal efficiency and simple reactor system. The high thermal efficiency is obtained by supercritical pressure water cooling. The fuel cladding surface temperature increases locally due to a synergistic effect from the increased coolant temperature, the expanded flow deflection due to coolant density change and the decreased heat transfer coefficient, if the coolant flow distribution is non-uniform in the fuel assembly. Therefore, the SCPR fuel assembly is designed using a subchannel analysis code based on the SILFEED code for BWRs.

The SCPR fuel assembly has many square-shaped water rods. The fuel rods are arranged around these water rods. The fuel rod pitch and diameter are 11.2 mm and 10.2 mm, respectively. Since coolant flow distribution in the fuel assembly strongly depends on the gap width between the fuel rod and the water rod, the proper gap width is examined. Subchannel analysis shows that the coolant flow distribution becomes uniform when the gap width is 1.0 mm. The maximum fuel cladding surface temperature is lower than 600°C and the temperature margin of the fuel cladding is increased in the design.     

超高通量快中子试验堆堆芯初步概念设计

针对先进核能系统发展需要,提出了超高通量堆的堆芯概念设计。本文采用板型燃料、正方形燃料组件设计,设置宽流道保证堆芯冷却剂占有较高的体积份额。堆芯采用52盒燃料组件,设置8盒控制棒组件和较厚的反射层。通过堆芯概念设计方案评价,结果表明堆芯循环长度可达100EFPD(等效满功率天),所提出的超高通量堆的最大中子注量率可达到1.08×10¹⁶ cm^-2·s^-1。     

超长寿命小型自然循环铅铋快堆堆芯概念设计研究

以提高铅铋快堆的经济性与固有安全性为目标，开展100 MWt超长寿命小型自然循环铅铋快堆SPALLER-100概念设计，在选用PuN-ThN燃料和²⁰⁸Pb-Bi冷却剂的基础上，提出了一种添加固体慢化剂BeO的燃料组件设计方案，开展了堆芯布置研究和控制棒系统设计，分析了堆芯物理特性与稳态自然循环特性。结果表明：在低燃料装载量和小堆芯体积条件下，SPALLER-100堆芯换料周期达32 a，平均卸料燃耗高达210.38 MW·d/kg(HM)，整个寿期内的反应性系数均为负值。稳态运行工况下燃料包壳、芯块最大温度均小于安全限值，反应堆具备一回路自然循环能力和一定流量自动分配能力。     

小型压水堆下腔室交混特性实验研究

为研究小型压水堆下腔室的交混特性,本文基于比例模化方法,开展小型压水堆1∶3比例模型水力学实验,通过测量溶液浓度变化,获得在冷管流量均衡和非均衡工况下堆芯入口的交混因子矩阵。研究结果表明,均衡流量工况下,冷管流量的变化对堆芯入口交混因子矩阵未产生明显影响;非均衡流量工况下,靠近出口管的燃料组件交混因子受流量不均衡的影响较大,而中心区域的交混因子变化幅度较小。由此可见,小型压水堆在均衡流量下具有较稳定的下腔室交混特性,而在非均衡工况下需要重点关注出口附近燃料组件交混特性的变化。