Annals of Nuclear Energy最新文献

A transient model for thermal–hydraulic characteristics of helical coiled once-through steam generator (HCOTSG) of liquid metal fast reactor was established based on global discrete grid method. The model meshed the liquid metal circuit and water-steam circuit. Meanwhile, unsteady heat conduction equations were built to accurately simulate coupled heat transfer between two sides. Four-equation drift-flux method was adopt for water-steam flow. Taking lead–bismuth fast reactor as example, thermal–hydraulic characteristics of HCOTSG were first calculated under steady conditions. It was found that heat flux distribution along steam generator is extremely uneven, and the difference between maximum and minimum wall heat flux reaches hundreds of times. Then, the transient response was simulated when inlet conditions of primary side were perturbed by step changes, and it was found that maximum wall heat flux increases from 1175.5 kW/m² to 1365 kW/m², increasing by nearly 16 %, when inlet lead–bismuth temperature step increases from 450°C to 480 °C.

Supercritical water-cooled reactors (SCWR) experience pressure reductions below the critical point during a loss of coolant accident (LOCA). During the reactor depressurization, boiling crises may occur, leading to pressure fluctuations and significant increases in wall temperatures, posing a severe threat to the cladding. In this study, a one-dimensional transient analysis code has been developed to incorporate one-dimensional steady-state flow equations in the fluid domain and transient thermal conductivity equations in the solid domain. The code also includes a wall heat transfer model and a model for the velocity of the moving quench front to simulate transcritical depressurization transient processes. The simulation successfully predicts the typical experimental phenomena. It is reasonable to use the critical temperature as the demarcation point between the dry and wet conditions of the axial wall of the rod bundle under transcritical conditions. By combining the experimental data with the program calculations, the critical interface for the occurrence of boiling crisis under transcritical conditions is determined using mass flow rate, heat flow density and fluid temperature as inputs. The criterion for the occurrence of CHF under transcritical conditions is obtained, which provides a reference to the safety analysis of SCWRs.

The lead–bismuth eutectic cooled fast reactor (LFR) emerges as a cutting-edge technology in nuclear reactor design. This study introduces a compact, passive, and long-life LFR reactor LFR-180. The LFR-180 incorporates modular helical-coiled once-through steam generators (H-OTSG) and direct reactor auxiliary cooling systems. This integration enables the generation of superheated steam and passive safety features of the LFR-180, eliminating the necessity for a water-steam separation device and contributing to enhanced power generation efficiency. A detailed transient thermal hydraulic analysis is conducted with tailored models for LBE solidification and H-OTSG thermal hydraulics. Results demonstrate that LFR-180 is a fully passive reactor, requiring no intervention up to 149.1 h after shutdown. The H-OTSG can serve as the heat exchangers of the active safety systems that can effectively reduce the peak core outlet temperature from 1003.6 K under the SBO accident to 724.2 K after shutdown.

The Kindai University Reactor, UTR-KINKI, is a research reactor with a rated thermal power of 1 W. The reactor is designed especially for education and training at universities and has been utilized for the education and training of nuclear engineering students in Japan since its first criticality in 1961. It is also used to educate the general public about nuclear science and technology, such as in training workshops for science teachers and high school students. Recently, it has been used to train nuclear engineers from overseas. Under the framework of allowing researchers from all over Japan to use the reactor, researches are conducted such as detector testing, reactor physics experiments, biological irradiations, and so on.

In order to improve the 1200°C steam oxidation properties of Cr-coated zirconium alloy accident tolerant fuel cladding tubes, chromium coatings with various micro-structures were fabricated by bipolar high-power impulse magnetron sputtering technology. The microstructures of chromium coatings transforms from unconsolidated column structure to dense & column-oblate structure with the bias voltage increase. Hardness and elastic modulus values of optimal chromium coatings are approximately 19.1 GPa and 374.8 GPa, respectively, which are 2.09 and 1.60 times higher than those of chromium coatings with unconsolidated column structures. Optimal chromium coatings with dense & column-oblate structures display preferable 1200°C steam oxidation resistance because of possession lower weight gain value and thicker Cr₂O₃ oxide layer after 2 h exposure periods. The fabrication strategy of chromium coatings with dense & column-oblate structure is expected to pave the way for enhancing 1200 °C steam oxidation resistance for Cr-coated zirconium alloys.

The Monte Carlo (MC) code RMC and sub-channel code CTF are coupled for Xi’an Pulsed Reactor (XAPR) pulse and depletion simulation. A python script is developed to handle data exchange through files. $3.45 pulse and $2 pulse are simulated with the pulsed state core. The pulse peak power and the Full Width at Half-Maximum (FWHM) results are compared with experiments as a validation and good agreement is achieved. Detailed 3-D power and temperature distributions are also obtained. Results show that the core power peak is close to the pulse rod where the reactivity is introduced. The radial power peak of each rod is at the boundary because of the self-shielding effect. The rod temperature distribution follows the same trend with the power, and the coolant temperature is not changed during the pulse period of about 0.12 s, which suggests that the heat transfer plays a negligible role in such a short time. For the steady state core, depletion simulation is carried out for the lifetime of the first fuel cycle of 120 Effective Full Power Days (EFPD). Validation is done by calculating the temperature distribution and the differential worth of the regulating rod at 0 EFPD, which both agree well with experiments. Results show that the power distribution is almost unchanged over time, only slightly more flat, indicating the material is not greatly depleted. The temperature distribution of the core generally agrees with the power distribution, except the radial temperature peak of each rod is at the center because heat is conducted outwards at steady state. Detailed coolant temperature distribution is obtained thanks to the utilization of CTF. Temperatures at the assembly boundary and near the central water chamber is noticeably lower than the other part, which is not shown in the parallel multi-channel models.

Heat pipe cooled reactors passively transfer heat from the core to the energy conversion system through alkali metal heat pipes. Further studies of the heat transfer characteristics of alkali metal heat pipes are needed to develop heat pipe cooled reactors. This work used a two-dimensional heat and mass transfer model of an alkali metal heat pipe. The flow field was validated against CFD predictions which showed that the model has less than 10 % errors in the velocity distribution and less than 5 % errors in the pressure distribution. A comparison of the predictions with experimental data in the literature indicated that the predicted wall temperature error was within 30 °C. This model was then used to analyze four factors that affect heat transfer in sodium heat pipes. The simulations show that higher heat fluxes shorten the startup time by approximately 70 % but significantly increase the operating temperature. The heat flux distribution significantly affects the evaporator temperature distribution. The heat transfer boundary condition on the condenser’s outer surface mainly affects the operating temperature and startup time. Increasing the heat pipe length-to-diameter ratio with a fixed heat flux increases the operating temperature with the increase proportional to the heat flux. This model provides accurate simulations of sodium heat pipes for various heat transfer boundary conditions. The simulations provide a reference for the selection of working conditions for heat pipe operation instability experiments.

One of the potential design concepts for small modular reactor(SMR) is coupling an integral pressurized water reactor(IPWR) with a supercritical CO₂(sCO₂) power cycle to reduce the plant footprint. However, the passive safety systems of this SMR concept have not been well established compared to steam power cycle based SMRs. In this study, a passive turbocharger utilizing the unique characteristics of sCO₂ is proposed as a way to improve the passive safety of sCO₂ power cycle. This is the concept of installing a turbine-compressor set at the connection of passive residual heat removal system(PRHRS) to effectively remove the high level of residual heat in the early stages of an accident rather than relying only on natural circulation cooling. Based on the findings, it can be concluded that proposed sCO₂ turbocharger concept has the potential to enhance performance of PRHRS and reduce the height of natural circulation loop leading to smaller volume.

To develop the expected isotopic signatures of online nuclear power plants, a CANDU-6 quarter-core reactor was modeled using Serpent, a Monte Carlo and Burnup code. The reactor model was simulated using nominal operating parameters, steady power levels and standard refueling procedures, to set the baseline for online operations. The model was burned for 500 refuelings totaling 1400 effective full power days. The core was divided into 1140 spatially-discrete fuel bundles with each tracking the density of 237 isotopes. Instantaneous core inventory snapshots were recorded at the time of each refueling to create a continuous inventory database. These snapshots provide the expected isotopic densities and ratios for virtually any fuel bundle position or burnup under nominal operating parameters. These values are useful in the event of accidents, short-cycles, or nuclear proliferation. The time-dependent and spatially dependent results for xenon effluent are used to develop an analytical method for calculating the expected International Monitoring System xenon ratio measurements based on fuel bundle leak rates. A possible false-positive nuclear proliferation scenario for a CANDU-6 operating under nominal parameters is also identified.

The heat exchanger (HE) is widely used in energy and power industries, which has a vital impact on the stability and efficient operation of relevant industrial systems. Different shaped heat exchange tubes are often designed specially to improve the thermal–hydraulic efficiency of HEs. In the study, two new geometries named the cruciform twisted tube and cruciform straight tube are introduced, and the heat & mass transfer properties of liquid lead (Pb) and supercritical carbon dioxide (s-CO₂) in the HEs are numerically studied to assess the thermal–hydraulic efficiency and application prospect of two kinds of cruciform tube HEs. The results demonstrate that the unique cross-section geometry of the cruciform tube improves the uniformity of the flow field, and the spiral structure of the cruciform twisted tube effectively improves the blending and turbulence of working fluid. The cruciform tube can improve the comprehensive performance of the HE by more than 12%.