[en] After the severe accident inside a nuclear reactor, the IVR (In-vessel retention) management strategy is an effective way to keep the integrity of pressure vessel and reduce risk of radioactive leakage by holding the damaged core materials through External Reactor Vessel Cooling (ERVC). One of the most important topics is the characteristics of the CHF (Critical Heat Flux) on the outside wall of RPV lower head. In this work, several influence factors on saturated pool boiling CHF for downward facing heating surface were studied, including the type and concentration of working fluid, orientation angle. Experimental results indicated that CHF increases with increase of orientation angle of heating surface. It also indicated that both Al₂O₃/H₂O and SiO₂/H₂O nanofluids could improve CHF dramatically compared to deionized water, but the CHF value of two types of nanofluids is similar to each other, which means type of nanoparticles has just little influence on CHF. (author)

[en] Reactor fuel's power distribution plays a vital role in designing the new generation thermionic Space Reactor Power Systems (SRPS). In this paper, the 1/12th SPACE-R's full reactor core was numerically analyzed with two kinds of different axial power distribution, to identify their impacts on thermal-hydraulic and thermoelectric characteristics. In the benchmark study, the maximum error between numerical results and existing data or design values ranged from 0.2 to 2.2%. Four main conclusions were obtained in the numerical analysis: a) The axial power distribution has less impact on coolant temperature. b) Axial power distribution influenced the emitter temperature distribution a lot, when the core power was cosine distributed, the maximum temperature of the emitter was 194 K higher than that of the uniform power distribution. c) Comparing to the cosine axial power distribution, the uniform axial power distribution would make the maximum temperature in each component of the reactor core much lower, reducing the requirements for core fuel material. d) Voltage and current distribution were similar to the axial electrode temperature distribution, and the axial power distribution has little effect on the output power

[en] Highlights: • An alkali metal heat pipe radiator is proposed for TOPAZ-II power system. • A numerical calculation is performed to predict the heat rejection characteristics. • Heat transfer performances of heat pipe radiator and original radiator are compared. - Abstract: An alkali metal heat pipe radiator is proposed for the Russian TOPAZ-II space reactor power system to replace the original pumped loop radiator with three advantages. First, the single point failure problem of the original radiator can be avoided. Second, because only the NaK manifold needs to be armored, the specific mass would be reduced and the reliability and survivability can be improved. Third, the heat pipe has nearly isothermal characteristics and high heat transfer capabilities. In the present paper, the high temperature heat pipes using potassium as working fluid, wire screen mesh as wick layer, made of stainless steel are adopted to remove waste heat by radiation. An integral carbon-carbon fin covering and connecting heat pipes as whole heat transfer radiator is selected for its higher thermal conductivity and lower self-weight. A detailed steady state numerical calculation using the finite element method coupled with finite difference method is performed to predict the heat rejection characteristics of the heat pipe radiator. Heat transfer performances of the new concept heat pipe radiator and the pumped loop radiator are compared with key parameters. The results show that the designed heat pipe radiator satisfies the waste heat rejection requirements of the TOPAZ-II power system under the normal operating conditions and has an ideal redundancy. The isothermal and safety of heat pipe radiator is also better than that of pumped loop radiator.

[en] As the key component of the reactor, the life of the reactor pressure vessel directly determines the life of the reactor. Therefore, the life extension of the reactor pressure vessel requires the improvement of the material performance, especially the radiation resistance. At present, some data about the radiation hardening and embrittlement behavior of reactor pressure vessel steel have been obtained, mainly the standard tensile properties, standard impact properties and so on. However, to understand the service behavior of materials at higher doses, higher neutron fluence irradiation tests are needed. The problem is that because of the large size of the standard sample, if the irradiation dose continues to increase, the surface dose of the irradiated sample will be larger, and it is difficult to test after irradiation. In addition, if the irradiation sample can be replaced by smaller sample, the utilization space of the irradiation pore will be greatly improved. The high-throughput characterization for irradiation performance of reactor pressure vessel steel can be achieved. Therefore, in this work the small punch test of reactor pressure vessel material A508-3 steel will be carried out at several temperatures (Room temperature ∼ -150 ℃). Combining with the observation of fracture morphology, the fracture mode of A508-3 steel is analyzed. Comparing the results of small punch test with the data obtained by standard impact test, the correlation on the performance data and fracture behavior will be established for A508-3 steel. (authors)

[en] Small nuclear reactor features higher power capacity, longer operation life than conventional power sources. It could be an ideal alternative of existing power source applied for special equipment for terrestrial or underwater missions. In this paper, a 25kWe heat pipe cooled reactor power source applied for multiple use is preliminary designed. Based on the design, a thermal-hydraulic analysis code for heat pipe cooled reactor is developed to analyze steady and transient performance of the designed nuclear reactor. For reactor design, UN fuel with 65% enrichment and potassium heat pipes are adopted in the reactor core. Tungsten and LiH are adopted as radiation shield on both sides of the reactor core. The reactor is controlled by 6 control drums with B4C neutron absorbers. Thermoelectric generator (TEG) converts fission heat into electricity. Cooling water removes waste heat out of the reactor. The thermal-hydraulic characteristics of heat pipes are simulated using thermal resistance network method. Thermal parameters of steady and transient conditions, such as the temperature distribution of every key components are obtained. Then the postulated reactor accidents for heat pipe cooled reactor, including power variation, single heat pipe failure and cooling channel blockage, are analyzed and evaluated. Results show that all the designed parameters satisfy the safety requirements. This work could provide reference to the design and application of the heat pipe cooled nuclear power source.

[en] Highlights: • Two fluid model combining with RPI wall boiling model is employed. • Coupled heat transfer between SG primary side and secondary side is obtained. • Subcooled flow boiling phenomenon in secondary side of SG is simulated. • TSP impact on thermal–hydraulic characteristics of secondary side flow is analyzed. - Abstract: Steam generator (SG) is the heat exchanger used to transfer heat generated by the nuclear reactor core to the secondary side. In this paper, the two-fluid model combining with inter-phase heat and mass transfer model, inter-phase momentum transfer model, and RPI wall boiling model proposed by Kurul and Podowski (1991) was applied to solve the local flow and heat transfer of subcooled flow boiling in the secondary side of SG tube bundles coupled with the primary side by using ANSYS CFX 12.0. The subcooled nucleate boiling phenomenon and the coupled heat transfer between the SG primary side and secondary side were obtained. Also, the effects of tube support plate (TSP) and the different inlet subcooling on the thermal–hydraulic characteristics of SG were studied. The model adopted here was validated by using the published experiment and was in reasonable agreement with experimental data. The numerical results reasonable revealed the subcooled flow boiling occurred in the SG secondary side and the distributions of key parameters around TSP, elucidating that this model can provide useful information to the design of the steam generator

[en] In order to study the phosphoric acid immersion electrolysis decontamination method of radioactive contamination 17-4PH stainless steel, a phosphoric acid immersion electrolysis experimental device was designed, the simulation experiment was carried out with the uncontaminated valve stem of the nuclear power plant as the research material and the best electrolytic decontamination parameters, including electrolyte concentration, current density and electrolyte temperature, were obtained. According to the best experimental parameters, the stained stem material was decontaminated. The results show that the test device is simple, safe and reliable. When the sample test time is 15 s, the sample surface β count has dropped from the initial 93/s to 2/s. It has been below the environmental background value of 9/s, the decontamination efficiency has reached 98%, and the decontamination requirements for the stained samples have been achieved. (authors)

[en] The free-piston Stirling engine (FPSE) has been widely used in aerospace owing to its advantages of high efficiency, high reliability, and self-starting ability. In this paper, a 20-kW FPSE is proposed by analyzing the requirements of space nuclear power reactor. A code was developed based on an improved simple analysis method to evaluate the performance of the proposed FPSE. The code is benchmarked with experimental data, and the maximum relative error of the output power is 17.1%. Numerical results show that the output power is 21 kW, which satisfies the design requirements. The results show that: a) reducing the pressure shell’s thickness can improve the output power significantly; b) the system efficiency increases with the wire porosity, while the growth of system efficiency decreases when the porosity is higher than 80%, and system efficiency exhibits a linear relationship with the temperatures of the cold and hot sides; c) the system efficiency increases with the compression ratio; the compression ratio increases by 16.7% while the system efficiency increases by 42%. This study can provide valuable theoretical support for the design and analysis of FPSEs for space nuclear power reactors

[en] The Free-Piston Stirling engine (FPSE) is of interest for many research in aerospace due to its advantages of long operating life, higher efficiency, and zero maintenance. In this study, a 1-kW FPSE was proposed by analyzing the requirements of Space Reactor Power Systems (SRPS), of which performance was evaluated by developing a code through the Simple Analysis Method. The results of SAM showed that the critical parameters of FPSE could satisfy the designed requirements. The heater of the FPSE was designed with the copper rectangular fins to enhance heat transfer, and the parametric study of the heater was performed with Computational Fluid Dynamics (CFD) software STAR-CCM^þ. The Performance Evaluation Criteria (PEC) was used to evaluate the heat transfer enhancement of the fins in the heater. The numerical results of the CFD program showed that pressure drop and Nusselt number ratio had a linear growth with the height of fins, and PEC number decreased as the height of fins increased, and the optimum height of the fin was set as 4 mm according to the minimum heat exchange surface area. This paper can provide theoretical supports for the design and numerical analysis of an FPSE for SRPSs

[en] Highlights: • Heat Pipe Radiator (HPR) unit for space nuclear power reactor is numerically studied. • Heat transfer performance of HPR unit is investigated under transient conditions. • Several heat transport limits are analyzed during potassium heat pipe startup. - Abstract: Heat pipe radiator, featured with remarkable advantages in heat transfer efficiency and inherent safety with small specific mass and little weight, are widely adopted to the heat-rejection system for space nuclear power reactors. In this paper, physical and numerical models are developed to obtain the startup and transient behaviors of radiator unit with a potassium (K) heat pipe covered by fin under space environment. The heat transfer limit theory is adopted as criteria for heat pipe operation success. Numerical results indicated that according to the internal vapor flow regimes, the K heat pipe startup could be divided into three distinct stages. The K heat pipe started up from frozen state successful and rapidly until the expected operation state is reached. Among the different heat transfer limits, only the sonic limit due to choked flow restricts the K heat pipe during the second stage. Overall, for the heat-rejection system of space nuclear power reactor, the heat pipe radiator unit can effectively radiated waste heat to the space environment in 5 min and responses fast under transient conditions.