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.     

Analysis of Ledinegg flow instability in natural circulation at supercritical pressure

Since the thermophysical properties of water change dramatically at near-critical and pseudocritical point, the instability of natural circulation at supercritical pressure may occur in a loop. To predict the region of instability of natural circulation at supercritical pressure, a test loop was built at Tsinghua University. The paper presents the information of the test loop and a numerical analysis model for the loop. The paper verified the numerical analysis code by experiment results and using the code to analyze the instability of the loop. The paper concludes conclusion that there will be no Ledinegg instability occurring at supercritical pressure in the loop.     

两相自然循环流动不稳定性脉动周期的实验和理论研究

通过实验研究两相自然循环流动不稳定性脉动周期,建立理论模型,用理论分析法得到脉动周期的理论公式。用该公式计算的结果与实验值符合得很好。     

Effect of steam quality on two-phase flow in a natural circulation loop

Ｅｆｆｅｃｔｏｆｓｔｅａｍｑｕａｌｉｔｙｏｎｔｗｏ－ｐｈａｓｅｆｌｏｗｉｎａｎａｔｕｒａｌｃｉｒｃｕｌａｔｉｏｎｌｏｏｐＪｉａＨａｉ－Ｊｕｎ（贾海军）;ＷｕＳｈａｏ－Ｒｏｎｇ（吴少融）;ＷａｎｇＮｉｎｇ（王宁）ａｎｄＹａｏＳｉ－Ｍｉｎ（姚思民）（Ｉｎ...     

Theoretical investigation on the steady-state natural circulation characteristics of a new type of pressurized water reactor

1 Introduction With respect to the inherent safety of nuclear re- actors, application of passive systems/components including natural circulation phenomena as the main mechanism is intended to simplify the safety-related systems and to improve their reliability, to reduce the effect of human errors and equipment failures, and to provide more time to enable the operators to prevent or mitigate serious accidents. Natural circulation is the main mode of heat removal for removing decay heat from t…     

Flashing and flashing coupled density wave oscillation in natural circulation system

The phenomeon and mechainsm of different kinds of two-phase flow instabilities,namely geysering,flashing instability and flashing coupled density wave instability are firstly well interpreted by the experiment performed on the test loop(HRTL-5) simulating the 5-MW reactor.The flashing coupled density wave instability is analyzed by using an onedimensional non-thermoequilibrium two-phase flow drift model computer code.Calculation results are in good agreement with the experimetal.     

The development and validation of a natural circulation analysis code for marine reactors

A code PNCMC (Program for Natural Circulation under Motion Conditions) has been developed for natural circulation simulation of marine reactors. The code is based on one-dimensional two-fluid model in noninertial frame of reference. The body force term in the momentum equation is considered as a time dependent function, which consists of gravity and inertial force induced by three-dimensional ship motion. Staggered mesh, finite volume method, semi-implicit first order upwind scheme and Successive Over Relaxation (SOR) method are used to discretize and solve two-phase mass, momentum and energy equations. Single-phase natural circulation experiments under rolling condition performed in Institute of nuclear and new energy technology of Tsinghua University and two-phase natural circulation experiments under rolling condition performed by Tan and colleagues are used to validate PNCMC. The validation results indicate that PNCMC is capable to investigate the single-phase and two-phase natural circulation under rolling motion.     

自然循环两相漂移流理论分析模型的建立

以5MW核供热堆试验回路（HRTL－5）为物理原型,从理论上推导了加热段欠热沸腾,上升段入口附近的冷凝,上升段闪蒸以及气空间压力平衡等关系式,并考虑了热力不平衡等因素的影响,建立了一个完整的四方程漂移流模型,该模型适用于分析低压低干度自然循环系统的流动问题,通过引入凝边界层,首次解决了两相流动过程中汽泡在过冷水中的冷凝问题,对汽空间则推导了压力及液位的约束方程,方程组可采用积分方法求解,并利用Rung-Kutta-Verner方法求解所得的微分方程组。     

Mass flow rate sensitivity and uncertainty analysis in natural circulation boiling water reactor core from Monte Carlo simulations

Our aim was to evaluate the sensitivity and uncertainty of mass flow rate in the core on the performance of natural circulation boiling water reactor (NCBWR). This analysis was carried out through Monte Carlo simulations of sizes up to 40,000, and the size, i.e., repetition of 25,000 was considered as valid for routine applications. A simplified boiling water reactor (SBWR) was used as an application example of Monte Carlo method. The numerical code to simulate the SBWR performance considers a one-dimensional thermo-hydraulics model along with non-equilibrium thermodynamics and non-homogeneous flow approximation, one-dimensional fuel rod heat transfer. The neutron processes were simulated with a point reactor kinetics model with six groups of delayed neutrons. The sensitivity was evaluated in terms of 99% confidence intervals of the mean to understand the range of mean values that may represent the entire statistical population of performance variables. The regression analysis with mass flow rate as the predictor variable showed statistically valid linear correlations for both neutron flux and fuel temperature and quadratic relationship for the void fraction. No statistically valid correlation was observed for the total heat flux as a function of the mass flow rate although heat flux at individual nodes was positively correlated with this variable. These correlations are useful for the study, analysis and design of any NCBWR. The uncertainties were propagated as follows: for 10% change in the mass flow rate in the core, the responses for neutron power, total heat flux, average fuel temperature and average void fraction changed by 8.74%, 7.77%, 2.74% and 0.58%, respectively.     

Investigation on two-phase flow instability in steam generator of integrated nuclear reactor

In the pressure range of 3-18MPa,high pressure steam-water two-phase flow density wave instability in vertical upward parallel pipes with inner diameter of 12mm is studied experimentally.The oscillation curves of two-phase flow instability and the effects of several parameters on the oscillation threshold of the system are obtained.Based on the small pertubation linearization method and the stability principles of automatic control system,a mathematical model is developed to predict the characteristics of density wave instability threshold.The predictions of the model are in good agreement with the experimental results.     

低温堆和沸水堆流动不稳定因素比较分析

在实验的基础上分析了低压低干度自然循环系统中欠热沸腾、冷凝、闪蒸、以及汽空间压力等因素对两相流动不稳定的影响,并与沸水堆条件下的情况作了比较。分析表明：①自然循环系统中,欠热沸腾、冷凝、闪蒸是影响流动不稳定的重要因素,很多不稳定问题与此有关,这跟以沸水堆为背景的强迫循环系统有很大不同;②汽空间大小对系统稳定性有重要作用,在压力允许条件下,应尽量减小汽空间,提高系统稳定性。     

钠冷快堆中熔融池模型的建立与验证

快堆在超设计基准事故下运行时,会导致钠沸腾和干涸,如果不能及时停堆,接着就会产生燃料元件的熔化坍塌,在组件盒下部形成熔融池.为了对熔融池给出合理的安全分析,采用机理建模的方法,建立了完整的熔融池模型,并在法国的SCARABEE系列实验中的BF1三种功率的实验上进行了验证,和实验吻合较好,通过和所验证过的GEYSER及BF2等实验模型进行比较,得出了有关熔融池机理的相关结论.通过排热和温升等相关数据的比较,对熔融池向外的排热形式给出了合理分析,并得出了相关结论.     

Steam drum level dynamics in a multiple loop natural circulation system of a pressure-tube type boiling water reactor

Advanced Heavy Water Reactor (AHWR) is a pressure tube type boiling water reactor employing natural circulation as the mode of heat removal under all the operating conditions. Main heat transport system (MHTS) of AHWR is essentially a multi-loop natural circulation system with all the loops connected to each other. Each loop of MHTS has a steam drum that provides for gravity based steam–water separation. Steam drum level is a very critical parameter especially in multi-loop natural circulation systems as large departures from the set point may lead to ineffective separation of steam–water or may affect the driving head. However, such a system is susceptible to steam drum level anomalies under postulated asymmetrical operating conditions among the different quadrants of the core like feedwater flow distribution anomaly among the steam drums or power anomaly among the core quadrants. Analyses were carried out to probe such scenarios and unravel the underlying dynamics of steam drum level using system code RELAP5/Mod3.2. In addition, a scheme to obviate such problem in a passive manner without dependence on level controller was examined. It was concluded that steam drums need to be connected in the liquid as well as steam space to make the system tolerant to asymmetrical operating conditions.     

Investigations on the role of mixed convection and wall friction factor in single-phase natural circulation loop dynamics

The objective of the present paper is to present a 1-D model for simulating the startup from rest of water cooled single-phase natural circulation loops having horizontal heaters. The starting point of analysis is the inability of the 1-D codes to account for natural convection in the heater. Present 1-D models are unable to account for axial diffusion in the fluid caused by natural convection. Start-up from rest and many other characteristics cannot be simulated using classical 1-D models because of the inherent tendency of the predicted dynamics to be attracted by zero flow condition. The paper presents an elegant approach for taking into account both natural and forced convection. The enhancement of fluid motion and thermal mixing by natural convection is an important consideration in the design of nuclear reactors. Hence, the model developed is of direct relevance to nuclear reactor thermal hydraulics. The model developed for natural convection has been validated against the CFD simulations. The model developed has been incorporated in a classical 1-D model developed by the authors previously. The application of model to a rectangular single-phase natural circulation loop show that the model can predict the loop behavior from start-up with fidelity. The model reproduces most of the characteristics like unidirectional oscillation, bidirectional oscillations and chaotic switching reasonably well. Finally, model has been used to investigate the phenomenon of hysteresis observed in experimental loop. The paper also brings out the role of constitutive laws for wall friction in predicting the loop dynamics.     

Design and safety optimization of ship-based small nuclear power reactors

Design and safety optimization of ship-based nuclear power reactors have been performed. The neutronic and thermo-hydraulic programs of the three-dimensional X–Y–Z geometry have been developed for the analysis of ship-based nuclear power plant. Quasi-static approach is adopted to treat seawater effect and quasi-static approach is also employed to treat neutronic aspect during safety analysis.

The reactors are loop type lead–bismuth-cooled fast reactors with nitride fuel and with relatively large coolant pipe above reactor core, the heat from primary coolant system is directly transferred to water–steam loop through steam generators. The power level is 100–200 MW th and excess reactivity is about 1$. Three types of core were investigated in the optimization process: balance, tall, and pancake with five values of Z–Y size ratio.

As the optimization results, the core outlet temperature distribution is changing with the elevation angle of the reactor system. The pancake core type has larger temperature distribution change as the elevation angle changes due to the sea wave. The natural circulation capability is good for safety. However, large driving head of natural circulation may cause large temperature fluctuation as the elevation angle changes.     

Natural circulation experiments in a non-uniform diameter lead bismuth loop and validation of LeBENC code

Experimental studies are carried out on natural circulation in a Lead Bismuth Eutectic (LBE) loop. The loop mainly consists of a heated section, air heat exchanger, valves, various tanks and argon gas control system. All the components and piping are made of SS316L. The dissolved oxygen in the LBE is monitored online by an Yttria Stabilised Zirconia (YSZ) oxygen sensor and controlled during the operation of the loop. In this paper the details of the loop and experimental studies carried out with heater power levels varying from 900 W to 5000 W are described. The temperature range of LBE during the experiments was 200 °C–500 °C. The maximum heat loss in the piping is kept less than 20% of the main heater power. Steady state experimental studies are carried out at different power levels and the LBE flow rate was found to be varying from 0.095 kg/s to 0.135 kg/s. The analysis and results of the performance of the heat exchanger with air and water as the secondary coolants are also discussed in the paper. Transient studies were carried out to simulate various events like heat sink loss, step power change and secondary side coolant flow rate change and reported in the paper. In the start up experiments, where the flow is started from stagnant condition of LBE, the time required for starting of natural circulation is found to be 600 s, 400 s and 240 s with power level of 1200 W, 2400 W and 3000 W respectively. The results are compared with available correlation and prediction of computer code LeBENC.     

Prevention and mitigation of severe accident developments and recriticalities in advanced fast reactor systems

GEN-IV nuclear systems, especially advanced sodium-cooled fast reactors (SFRs) are on the horizon and a key issue of their success is the promise of a higher and improved safety level. In Europe safety investigations are currently under way e.g. in the collaborative CP-ESFR project of the EU. Both on the prevention and mitigation side significant efforts are invested to fulfill the high safety goals. One route of assurance concentrates on the mitigation or even elimination of specific severe accident routes leading to core disruption and recriticalities. The accident phase with larger disrupted and molten fuel regions is coined the transition phase. A competition between fuel losses and in-pool material motion exists deciding over recriticalities and energetics potentials in this phase. To get a control of the transition phase recriticalities and energetics, ideas have been developed to install dedicated means in the core that enhance and guarantee a sufficient and timely fuel discharge - a controlled material relocation (CMR). Several proposals are under way to accomplish this CMR and especially in Japan significant theoretical and experimental work has been performed. In Europe the path to develop CMR measures was driven in the past by the development of the CAPRA reactors with a high Pu enrichment. In the current paper the status of analyses is described and some new concepts are discussed. The CMR measures are discussed along two accident scenarios, the unprotected loss of flow (ULOF) and the instantaneous blockage accident (TIB).     

Co-processing of uranium and plutonium for sodium-cooled fast reactor fuel reprocessing by acid split method for plutonium partitioning without reductant

A solvent extraction flowsheet for Pu partitioning, based on the acid split method without reductant, originally proposed by the Oak Ridge National Laboratory (ORNL), was tested for sodium-cooled fast reactor fuel reprocessing. To enhance resistance to nuclear proliferation, a flowsheet for co-processing was developed that controls Pu content in the products while avoiding Pu polymerization and formation of a third phase during extraction. In this method, Pu is partitioned using the difference in distribution coefficients of U and Pu. It is effective for selective Pu stripping from U at low temperatures and HNO₃ concentrations. The flowsheet with a supply of 0.15 mol/dm³ HNO₃ solution at 21°C for Pu partitioning was tested experimentally using miniature centrifugal contactors and a highly radioactive solution. Neither a Pu(IV) polymer nor a third phase was observed during the experiment. The Pu content in the U/Pu product increased to 2.28 times that in the feed solution. The leakage ratio of Pu to the U product was slightly less in the U stripping section. Some fission products (FPs) were effectively decontaminated; e.g., decontamination factors (DFs) of Cs in U/Pu and U products were 4.51×10⁵ and 2.42×10⁵, respectively.     

Design and preliminary test of an annular linear induction electromagnetic pump for a sodium-cooled fast reactor thermal hydraulic experiment

An annular linear induction electromagnetic pump (ALIP) with a flow rate of 2265 L/min and a developed pressure of 4 bar was designed and fabricated to test the performance of the components of a sodium-cooled fast reactor (SFR) in a sodium thermal hydraulic experimental loop. The design characteristic of the ALIP was calculated using the electrical equivalent circuit method typically used for analyzing linear induction machines. Preliminary tests, such as verification of the moving function using an annular Al pipe, were carried out. The linearity between the input voltage, current, and magnetic flux density was verified. The developed force demonstrated an increase proportional to the square of the input current, whereas the velocity was linearly proportional to the input current. The main design variables of the pump were calculated theoretically for the SFR thermal hydraulic experimental loop. The pump was optimized for the design variables including input frequency, and the characteristics of the optimized pump were compared with those of the pump at the commercially used frequency of 60 Hz.