A modular approach to lead-cooled reactors modelling

After an overview of the lego plant simulation tools (LegoPST), the paper gives some details about the ongoing LegoPST extension for modelling lead fast reactor plants. It refers to a simple mathematical model of the liquid lead channel dynamic process and shows the preliminary results of its application in dynamic simulation of the BREST 300 liquid lead steam generator. Steady state results agree with reference data [IAEA-TECDOC 1531, Fast Reactor Database, 2006 Update] both for water and lead.     

Thermohydraulic studies of sodium-cooled reactor facilities

Thermohydraulic studies of reactor facilities with fast reactors are complex experimentally and computationally. Extensive experimental data are obtained on the velocity and temperature profiles, hydrodynamic resistance and heat emission, initial heat section, and interchannel mixing of the coolant in the fuel assemblies. These are used to develop engineering methods of performing thermohydraulic calculations of fuel assemblies as well as computational compute codes. The particulars of the hydrodynamics and heat transfer in intermediate heat exchangers and steam generators of reactor facilities with fast reactors are studied. This has made it possible to validate their thermohydraulic characteristics.     

Design of a fuel element for a lead-cooled fast reactor

The options of a lead-cooled fast reactor (LFR) of the fourth generation (GEN-IV) reactor with the electric power of 600 MW are investigated in the ELSY Project. The fuel selection, design and optimization are important steps of the project. Three types of fuel are considered as candidates: highly enriched Pu-U mixed oxide (MOX) fuel for the first core, the MOX containing between 2.5% and 5.0% of the minor actinides (MA) for next core and Pu-U-MA nitride fuel as an advanced option. Reference fuel rods with claddings made of T91 ferrite-martensitic steel and two alternative fuel assembly designs (one uses a closed hexagonal wrapper and the other is an open square variant without wrapper) have been assessed. This study focuses on the core variant with the closed hexagonal fuel assemblies. Based on the neutronic parameters provided by Monte-Carlo modeling with MCNP5 and ALEPH codes, simulations have been carried out to assess the long-term thermal-mechanical behaviour of the hottest fuel rods. A modified version of the fuel performance code FEMAXI-SCK-1, adapted for fast neutron spectrum, new fuels, cladding materials and coolant, was utilized for these calculations. The obtained results show that the fuel rods can withstand more than four effective full power years under the normal operation conditions without pellet-cladding mechanical interaction (PCMI). In a variant with solid fuel pellets, a mild PCMI can appear during the fifth year, however, it remains at an acceptable level up to the end of operation when the peak fuel pellet burnup ∼80 MW d kg⁻¹ of heavy metal (HM) and the maximum clad damage of about 82 displacements per atom (dpa) are reached. Annular pellets permit to delay PCMI for about 1 year. Based on the results of this simulation, further steps are envisioned for the optimization of the fuel rod design, aiming at achieving the fuel burnup of 100 MW d kg⁻¹ of HM.     

SSTAR: The US lead-cooled fast reactor (LFR)

It is widely recognized that the developing world is the next area for major energy demand growth, including demand for new and advanced nuclear energy systems. With limited existing industrial and grid infrastructures, there will be an important need for future nuclear energy systems that can provide small or moderate increments of electric power (10-700 MWe) on small or immature grids in developing nations. Most recently, the global nuclear energy partnership (GNEP) has identified, as one of its key objectives, the development and demonstration of concepts for small and medium-sized reactors (SMRs) that can be globally deployed while assuring a high level of proliferation resistance. Lead-cooled systems offer several key advantages in meeting these goals. The small lead-cooled fast reactor concept known as the small secure transportable autonomous reactor (SSTAR) has been under ongoing development as part of the US advanced nuclear energy systems programs. It is a system designed to provide energy security to developing nations while incorporating features to achieve nonproliferation goals, anticipating GNEP objectives. This paper presents the motivation for development of internationally deployable nuclear energy systems as well as a summary of one such system, SSTAR, which is the US Generation IV lead-cooled fast reactor system.     

Thermohydraulic Calculation of VVER and SCWR Reactor Channels

Atomic Energy - The results of a comparative thermohydraulic calculation of the core channels of two types of nuclear reactors – VVER and VVER-SKD, differing in the coolant parameters at the...     

HP-STMCs空间堆堆芯典型瞬态热工分析

以计算流体力学(CFD)为基础,编写HP-STMCs空间堆堆芯功率瞬变模型和反应性反馈模型的用户自定义函数(UDF),开发堆芯瞬态分析程序SNPS-FTASR。对程序的正确性进行验证并得到满意的结果后,用SNPS-FTASR分析1个控制鼓误动作向堆芯引入正反应性和堆芯1根热管失效时的瞬态响应特性。结果显示:在1个控制鼓误动作引入正反应性时,堆芯功率先迅速升高后因堆芯反应性负反馈而缓慢上升,最终堆芯功率稳定在额定功率的121.3%。在堆芯1根热管失效时,堆芯UN燃料芯块的温度先迅速升高后因反应性负反馈使得堆芯功率迅速下降,最终堆芯功率稳定在额定功率的88.7%,堆芯最高温度较稳定状态上升约140 K,表明热管冷却空间堆在一个控制鼓误动作和1根热管失效时热工方面是安全的。     

Core design investigation for a SUPERSTAR small modular lead-cooled fast reactor demonstrator

In this paper a preconceptual neutronics design study for a SUstainable Proliferation-resistance Enhanced Refined Secure Transportable Autonomous Reactor (SUPERSTAR) demonstrator is presented. The main goal of achieving the highest realistic power level limited by natural circulation and transportability, while providing energy security and proliferation resistance thanks to a long core lifetime design has been satisfactorily attained. A preliminary core configuration has been developed meeting the foremost requirements of limiting the reactivity swing over the core lifetime to about 1 $ and flattening the radial power profiles, as demanded by the choice of wrapper-less (i.e. without flow ducts) fuel assemblies and by the stringent technological constraints imposed by the requirement of short-term deployment. Reactivity coefficients and kinetic parameters have been evaluated for the reference beginning-of-life, middle-of-life and end-of-life core configurations. Furthermore, the results of thermal-hydraulic analyses of the primary loop have confirmed that the system can be effectively cooled by natural circulation heat transport, all the technological constraints being respected even when incorporating peaking factors.     

Comparative study of minor actinide transmutation in sodium and lead-cooled fast reactor cores

This paper shows that lead-cooled and sodium-cooled fast reactors (LFRs and SFRs) can preferentially consume minor actinides without burning plutonium, both in homogeneous and in heterogeneous mode. The former approach consists of admixing about 5% of minor actinides (MAs) into uranium–plutonium fuels in the core and using a limited number of thermalising pins consisting of UZrH_1.6. These are needed to keep the negative Doppler feedback larger than the positive coolant reactivity coefficient. Our Monte Carlo burn-up calculations showed that a 600 MW_e LFR self-breeder without blankets can burn an average of around 67 kg annually of MAs with a reactivity swing of only about −0.7$ per year. The reactivity swing of a corresponding 600 MW_e SFR is more than three times larger due to the poorer breeding and half the critical mass in comparison to the LFR. However, when axial and radial blankets loaded with 10% MAs are added, the SFR burns 25% more MAs (131 kg/yr) and breeds 30% more Pu (150 kg/yr) than an equally sized LFR. When only the blankets are loaded with MAs, the SFR breeds 30% more Pu (198 kg/yr) and still burns about 60 kg a year of MAs. However, in terms of severe accident behaviour, the LFR, with its superior natural coolant circulation and larger heat capacity, has definite advantages.     

Thermohydraulic Aspects of Deep Burnup of Nuclear Fuel in Fast Reactors

The results of thermomechanical and thermohydraulic studies showing the relative effect of the deformation of fuel-element claddings and lattices in fast-reactor fuel assemblies on their temperature regimes are presented. It is shown that the temperature nonuniformities in fuel assemblies largely determine the deformation of fuel assemblies and, in turn, the operating efficiency and, correspondingly, the degree of burnup of nuclear fuel in fast reactors. The increase in the efficiency of the fuel assemblies is largely due to temperature smoothing, including smoothing of local temperature nonuniformities. Various solutions to technical and structural problems can accomplish this.