Water and hydrogen in heavy liquid metal coolant technology

Lead and lead-bismuth are currently considered as coolants in some reactor and accelerator driven system designs. Processes related to water steam and hydrogen supply to the primary circuit are very important for the coolant technology. The similar physical and chemical processes occur, for instance, in case of steam leaks to the coolant caused by failure of steam generator (SG). In this paper are designs of heat and mass exchangers, in which water or other liquid and gaseous media are in direct contact with heated lead-containing coolant. These heat and mass exchangers can be used for evaporation of liquids, as well as for hydrogen production technologies.     

Hydrodynamic characteristics of lead and lead-bismuth coolants

The results of an analysis of the effect of the physical properties of lead and lead-bismuth coolants on the hydrodynamic characteristics and the results of experimental investigations of the particulars of the hydrodynamic flows of these coolants in application to loops with fast reactors are presented. It is shown that cavitation, in the conventional meaning of this word, cannot arise in flow part of vane pumps pumping lead and lead-bismuth coolants in a reactor loop. It is confirmed that a gas gap can form between the surface of a heavy liquid-metal coolant flow and the channel walls not wetted by it. The results of experimental studies of the rupture of a column of heavy liquid-metal coolant and detachment of a centrifugal pump flow, probably because of the appearance of gas cavitation, are presented.     

Determining water chemistry conditions in nuclear reactor coolants

The chemistry of the process and coolant systems in water-cooled nuclear reactors is tightly controlled to minimise material degradation and, for some systems, to regulate reactor power. Tight control entails monitoring the systems and making appropriate adjustments. Online monitoring can be utilised where instruments are available but otherwise samples must be taken and measurements made offline. This paper reviews the current technologies for monitoring and sampling.     

Cross-comparison of fast reactor concepts with various coolants

Four fast reactor concepts using lead (LFR), liquid salt, NaCl-KCl-MgCl₂ (LSFR), sodium (SFR), and supercritical CO₂ (GFR) coolants are compared. Since economy of scale and power conversion system compactness are the same by virtue of the consistent 2400 MWt rating and use of the S-CO₂ power conversion system, the achievable plant thermal efficiency, core power density and core specific powers become the dominant factors. The potential to achieve the highest efficiency among the four reactor concepts can be ranked from highest to lowest as follows: (1) GFR, (2) LFR and LSFR, and (3) SFR. Both the lead- and salt-cooled designs achieve about 30% higher power density than the gas-cooled reactor, but attain power density 3 times smaller than that of the sodium-cooled reactor. Fuel cycle costs are favored for the sodium reactor by virtue of its high specific power of 65 kW/kgHM compared to the lead, salt and gas reactor values of 45, 35, and 21 kW/kgHM, respectively. In terms of safety, all concepts can be designed to accommodate the unprotected limiting accidents through passive means in a self-controllable manner. However, it does not seem to be a preferable option for the GFR where the active or semi-passive approach will likely result in a more economic and reliable plant. Lead coolant with its superior neutronic characteristics and the smallest coolant temperature reactivity coefficient is easiest to design for self-controllability, while the LSFR requires special reactivity devices to overcome its large positive coolant temperature coefficient. The GFR required a special core design using BeO diluent and a supercritical CO₂ reflector to achieve negative coolant void worth—one of the conditions necessary for inherent shutdown following large LOCA. Protected accidents need to be given special attention in the LSFR and LFR due to the small margin to freezing of their coolants, and to a lesser extent in the SFR.     

Scram and nonlinear reactor system seismic analysis for a liquid metal fast reactor

Earthquake vibrations cause large forces and stresses that can significantly increase the scram time required for safe shutdown of a nuclear reactor. The horizontal deflections of the reactor system components cause impact between the control rods and their guide tubes and ducts. The resulting frictional forces, in addition to other operational forces, delay the travel time of the control rods. To obtain seismic responses of the various reactor system components (for which a linear response spectrum analysis is considered inadequate) and to predict the control rod drop time, a non-linear seismic time history analysis is required. Nonlinearities occur due to the clearances or gaps between various components. When the relative motion of adjacent components is large enough to close the gaps, impact takes place with large impact accelerations and forces.This paper presents the analysis and results for a liquid metal fast rector system which was analyzed for both scram times and seismic responses such as bending moments, accelerations and forces. The reactor system was represented with a one-dimensional nonlinear mathematical model with two degrees of freedom per node (translational and rotational). The model was developed to incorporate as many reactor components as possible without exceeding computer limitations. It consists of 12 reactor components with a total of 71 nodes, 69 beam and pin-jointed elements and 27 gap elements. The gap elements were defined by their clearances, impact spring constants and impact damping constants based on a 50% co-efficient of restitution.The horizontal excitation input to the model was the response of the containment building at the location of the reactor vessel supports. It consists of a 10 sec safe shutdown eathquake (SSE) acceleration-time history at 0.005 sec intervals and with a maximum acceleration of 0.408 g. The analysis was performed with two Westinghouse special purpose computer programs. The first program calculated the reactor system seismic responses and stored the impact forces on tape. The impact forces on the control rod driveline were converted into vertical frictional forces by multiplying them by a coefficient of friction, and then these were used by the second program for the scram time determination.The results give time history plots of various seismic responses, and plots of scram times as a function of control rod travel distance for the most critical scram initiation times. The total scram time considering the effects of the earthquake was still acceptable but about four times longer than that calculated without the eathquake. The bending moment and shear force responses were used as input for the structural analysis (stresses, deflections, fatigue) of the various components, in combination with the other applicable loading conditions.     

Safety study of primary coolants in fusion power reactor first wall design

This paper examines potential safety problems associated with the various primary coolant candidates currently considered for the EPR fusion blanket designs. The basic concern is the possibility of overheating and melting of the first wall and the blanket, induced by a malfunction in the primary coolant system. These accidents include the loss-of-coolant flow, the loss-of-heat removal, overpower transients, and the loss of coolant. Following a mechanistic safety for these four types of accident sequences and comparing helium and liquid metal cooling, it was found that helium has a more adverse effect on the first-wall heat up in the event of a loss-of-heat removal or a loss-of-coolant because its lack of thermal inertia.     

Thermal-hydraulic performance of heavy liquid metal in straight-tube and U-tube heat exchangers

Motivated by an increased interest in heavy liquid metal (lead or lead alloy) cooled fast reactors (LFR) and accelerator-driven system (ADS), the present paper presents a study on resistance characteristics and heat transfer performance of liquid lead bismuth eutectic (LBE) flow through a straight-tube heat exchanger and a U-tube heat exchanger. The investigation is performed on the TALL test facility at KTH. The heat exchangers have counter-current flow arrangement, and are made from a pair of 1-m-long concentric ducts, with the LBE flowing in the inner tube of 10 mm I.D. and the secondary coolant flowing in the annulus. The inlet temperature of LBE into the heat exchangers is from 200 °C to 450 °C with temperature drops from 0 °C to 100 °C within the LBE flow range of Re = 10⁴-10⁵. Analysis of the experimental results obtained provides a basic understanding and quantification of the regimes of lead-bismuth flow and heat transfer through a straight tube and a U-shaped tube. The unique data base also serves as benchmark and improvement for system thermal-hydraulic codes (e.g. RELAP, TRAC/AAA) whose development and testing were dominantly driven by applications in water-cooled systems. Lessons and insights learnt from the study and recommendations for the heat exchanger selection are discussed.     

Breeding and void reactivity analysis on heavy metal closed-cycle water cooled thorium reactor

Design parameters of heavy water (D₂O) cooled thorium breeder reactors for actinides closed-cycle cases have been investigated to find a design feasible area of breeding and negative void reactivity. Heavy metals (HMs) closed-cycle shows narrower feasible area compared with feasible area of ²³³U closed-cycle. In thorium fuel cycle, the breeding capability of the reactors becomes worse when all HMs are recycled. The result shows an opposite profile of breeding capability compared with uranium fuel cycle which obtains higher breeding capability when more HMs are recycled. Feasible design area which has a breeding and negative void reactivity can be estimated for higher burnup, even higher than 60 GW d/t for ²³³U closed-cycle; however, it is limited to 36 GW d/t for HM closed-cycle. Contribution of capture ²³⁵U is more significant to reduce breeding capability and contribution of ²³⁴U is also more effective to make the reactor more positive or less negative void coefficient for HM closed-cycle case in thorium fuel cycle system.     

Model development for analysis of the Korea advanced liquid metal reactor

The SSC-K code is under development for analysis of the Korea Advanced LIquid MEtal Reactor (KALIMER) design adopting a pool-type reactor in Korea. The SSC-L code which was originally developed at Brookhaven National Laboratory for analysis of a loop-type liquid metal reactor, is its precursory code. The main reason for the development is that SSC-L cannot be applied directly to the KALIMER design because its application is limited to only a loop-type reactor. The SSC-K code represents the core with multiple coolant channels incorporated with a point kinetics model for calculation of the reactivity feedback. It can provide detailed one-dimensional thermal-hydraulic simulations not only for the primary and secondary sodium coolant circuits, but also the steam/water circuit of the balance-of-plant. This paper presents an overview of the recent developments on the physical models for SSC-K. Those developments are concerned with the two-dimensional hot pool model for analysis of the thermal stratification phenomena in the hot pool, the model for the passive decay heat removal system, the sodium boiling model in the core, and other physical models necessary for the KALIMER analysis. It also demonstrates the analysis results for the unprotected accidents like unprotected transient over power, unprotected loss of flow, and unprotected loss of heat sink postulated in the preliminary KALIMER design. The major focus of these analyses is made on confirmation of the inherent safety characteristics for the reactivity feedback in the core.     

Evolutionary design of reactor vessel assembly for liquid metal cooled battery

The Battery Omnibus Reactor Integral System (BORIS) is being developed as a multipurpose integral fast reactor at the Seoul National University. This paper focuses on developing design methodology for optimizing geometry of the liquid metal cooled reactor vessel assembly. The key design parameters and constraints are chosen considering technical specifications such as thermal limits and manufacturing difficulties. The evolution strategy is adopted in optimizing the geometry. Two objective functions are selected based upon economic and thermohydraulic reasons. Optimization is carried out in the following steps. First, selected design values are supplied to the momentum integral model code to evaluate steady-state mass flow rate and coolant temperature distribution of the reactor vessel assembly utilizing the thermodynamic boundary condition on heat exchanger calculated by the thermodynamics code. Second, the objective function values are calculated and compared against the previous results. The steps are repeated until an optimum value is obtained. Results of the improved design of the reactor vessel assembly are presented and their characteristics are discussed.     

Radiation stability of low-melting organic coolants in the liquid and solid phases

The radical yields G_r in gas oil, hydroterphenyl, and monoisopropyldiphenyl were measured after their irradiation by electrons at –170°C, as were the yields of high-boiling products G_h.b after irradiation of these same substances in a reactor at a temperature of 50°C. From a comparison of these quantities it follows that larger values of G_r correspond to larger values of G_{h. b}. It was shown that the calculated radiation yields of radicals in the liquid phase differ from the values measured in the solid phase by no more than four-fold. On this basis it is concluded that the radiation stability of organic coolants in the liquid phase can be predicted according to the radiation yield of radicals in the solid phase.Translated from Atomnaya Énergiya, Vol. 22, No. 1, pp. 27–30, January, 1967.     

Liquid metal cooled fast breeder reactor development and its mechanical-structural requirements and problems

After a brief survey about the history and the present status of liquid metal cooled fast reactors the paper explains the specific properties of liquid sodium and the principal layout of sodium-cooled fast reactor plants. The most important mechanical-structural requirements and problems are broken down into fields of sodium containment, core structure, and special mechanisms to be operated in sodium and the cover gas atmosphere.     

液态金属在堆芯子通道内的湍流换热

采用Speziale-Sarkar-Gatski(SSG)雷诺应力模型对液态金属在堆芯子通道内的流动、传热过程进行计算流体动力学(Computational Fluid Dynamics,CFD)模拟,研究雷诺数(Re)、分子普朗特数(Pr)、格拉晓夫数(Gr)、节径比(P/D)等无量纲参数对湍流换热的影响。比较无量纲对流换热系数(Nu)可以看出,CFD预测值与实验值及经验关系式符合得较好。对各种不同无量纲参数下的计算结果进行分析发现:在P/D和Re数相同条件下,三角形子通道的壁面温度分布比方形更均匀,换热情况更好;提高Re数,增大P/D,选用Pr数大的冷却剂,可有效改善温度和换热的周向分布不均情况;在Re数大于10 000的条件下,浮力对液态金属换热的影响可忽略不计。     

Investigation on heavy liquid metal cooling of ADS fuel pin assemblies

In the framework of accelerator driven sub-critical reactor systems heavy liquid metals are considered as coolant for the reactor core and the spallation target. In particular lead or lead bismuth eutectic (LBE) exhibit efficient heat removal properties and high production rate of neutrons.However, the excellent heat conductivity of LBE-flows expressed by a low molecular Prandtl number of the order 10⁻² requires improved modeling of the turbulent heat transfer. Although various models for thermal hydraulics of LBE flows are existing, validated heat transfer correlations for ADS-relevant conditions are still missing. In order to validate the sub-channel codes and computational fluid dynamics codes used to design fuel assemblies, the comparison with experimental data is inevitable.Therefore, an experimental program composed of three major experiments, a single electrically heated rod, a 19-pin hexagonal water rod bundle and a LBE rod bundle, has been initiated at the Karlsruhe Liquid metal Laboratory (KALLA) of the Karlsruhe Institute of Technology, in order to quantify and separate the individual phenomena occurring in the momentum and energy transfer of a fuel assembly.     

The detection of sodium vapor bubble collapse in a liquid metal fast breeder reactor

Sodium boiling detection utilizing the sound pressure emanated during the collapse of a sodium vapor bubble in a subcooled media is discussed in terms of the sound characteristic, the reactor ambient noise background, transmission loss considerations and performance criteria. Data obtained in several loss of flow experiments on Fast Test Reactor Fuel Elements indicate that the collapse of the sodium vapor bubble depends on the presence of a subcooled structure or sodium. The collapse pressure pulse was observed in all cases to be on the order of a kPa, indicating a soft type of cavitational collapse. Spectral examination of the pulses indicates the response function of the test structure and geometry is important. The sodium boiling observed in these experiments was observed to occur at a low (<50°C) liquid superheat with the rate of occurrence of sodium vapor bubble collapse in the 3 to 30 Hz range. Reactor ambient noise data were found to be due to machinery induced vibrations, flow induced vibrations, and flow noise. These data were further found to be weakly stationary enhancing the possibility of acoustic surveillance of an operating Liquid Metal Fast Breeder Reactor. Based on these noise characteristics and extrapolating the noise measurements from the Fast Flux Test Facility Pump (FFTP), one would expect a signal to noise ratio of up to 20 dB in the absence of transmission loss. The requirement of a low false alarm probability is shown to necessitate post detection analysis of the collapse event sequence and the cross correlation with the second derivative of the neutronic boiling detection signal. Sodium boiling detection using the sounds emitted during sodium vapor bubble collapse are shown to be feasible but a need for in-reactor demonstration is necessary.