Annals of Nuclear Energy最新文献

The recent revision of Emergency Core Cooling System (ECCS) regulations in Korea has necessitated the consideration of Fuel Fragmentation, Relocation, and Dispersal (FFRD) phenomena in nuclear reactor safety analyses. Consequently, the Korea Atomic Energy Research Institute (KAERI) has developed and integrated the FFRD model into the domestically licensed safety analysis code, SPACE. Globally, US NRC’s FRAPTRAN and IRSN’s DRACCAR are available for evaluating FFRD phenomena. The FRAPTRAN code incorporates the FFR model, developed by Quantum Technology, while the fuel dispersal model is currently not included. In contrast, the DRACCAR code functions as an integrated analysis platform capable of modeling multi-dimensional thermal, hydraulic, mechanical, and chemical phenomena during a Loss of Coolant Accident (LOCA). This study conducts a thorough examination of the FFRD model in the DRACCAR code and validates its applicability through analysis using the Halden IFA-650 tests. The results demonstrate satisfactory predictive capabilities. Furthermore, a parametric study of key FFRD model parameters enhances the understanding of the FFRD model in the DRACCAR code. The development of detailed physical models in the FFRD model could significantly enhance the performance of the DRACCAR code, warranting the establishment of a comprehensive framework for these advancements. In the future, code-to-code comparisons between the DRACCAR code and other domestically developed integrated analysis platforms will be conducted to investigate various phenomena in depth.

High-temperature heat pipes, utilizing alkali metals as working fluids, are essential heat transfer components in cooling systems of nuclear reactors. To assess the long-term isothermal behavior and heat transfer performance variation of high-temperature sodium heat pipes, this research presents the design and construction of a long-duration experimental test rig for high-temperature heat pipes. A 100-hour experimental investigation was conducted under operating conditions of 900 °C. The results demonstrate that the heat pipes can operate stably for extended periods after startup. The average temperature at the condenser section gradually increased, while the overall temperature difference fluctuation decreased. The magnitude of temperature difference reduction was measured to be 2.3 °C, and the effective thermal resistance of the heat pipe decreased to 0.0639 K/W. These results suggest an enhancement in both the isothermal performance and heat transfer capability characteristics of the sodium heat pipe after long-term testing. This study provides valuable insights for the design and assessment of high-temperature heat pipe systems in nuclear reactor cooling applications.

The transfer function is an important characteristic quantity of a nuclear reactor, since it contains the kinetic parameters. It expresses the response of a nuclear reactor to a disturbance of a certain frequency. If it is determined experimentally, it can be used to draw conclusions about the kinetic parameters. This article presents results of measurements of the zero power transfer function of the AKR-2 reactor at TU Dresden together with new data analysis methods. These measurements are compared to the theoretical zero power transfer function with kinetic parameters obtained via Monte Carlo simulations with MCNP and Serpent. To this end, advanced data analysis techniques based on a bootstrapping algorithms are employed. These techniques suppress the signal outside multiples of the fundamental frequency and additionally allow to obtain the full probability distribution of a peak in the frequency domain. This allowed for a reliable estimation of the mean value and uncertainty estimates of measured data of the zero power transfer function and the quantification of deviations between the experiments and the computations. It also made it possible to determine the phase of the zero power transfer function of AKR-2 for the first time. The experiments and computations are in agreement within the estimated uncertainties.

This study introduces the multigroup advanced semi-analytic nodal method (A-SANM) tailored for the high-temperature gas-cooled reactor (HTGR) analysis. The A-SANM has been crafted specifically for reactors with hexagonal geometries, such as the Vodo-Vodyanoi energetichesky reactor (VVER) and HTGR. A triangular node was constructed with a 12-term basis to delineate the flux by integrating both the polynomial and hyperbolic functions. The multigroup calculation kernel of this approach was embedded in the nodal diffusion code, RAST-V. To evaluate the computational efficiency of the A-SANM, we employed the MHTGR-350 benchmark. This benchmark, associated with a modular high-temperature gas-cooled reactor, was established by the OECD/NEA under the NGNP Project in 2021. In this study, we conducted the Phase I calculations to evaluate the performance of the neutronics code. Key parameters including the multiplication factor, rod worth, and axial and radial power distributions were meticulously assessed. When juxtaposed with the Monte Carlo code MCS, the A-SANM exhibited a deviation of –97 pcm. Differences in the axial and radial power were ± 4 and ± 3 %, respectively. Furthermore, the rod worth discrepancy was –6 pcm when set against the MCS. In summary, this study effectively elucidates the potential and precision of the multigroup A-SANM for the HTGR evaluations.

Since the compositions and properties of the fuel debris are uncertain, critical experiments are required to validate calculation codes and nuclear data used for the safety evaluation. For this purpose, the Japan Atomic Energy Agency (JAEA) has been modifying a critical assembly called “STACY.” The first criticality of the modified STACY is scheduled for spring 2024. This paper reports the consideration results of the core configurations of the modified STACY for the first criticality and inspections. The specifications of these core configurations were determined in advance, we tried to make them with a simplified computational model that considers the reactivity effect around the core. At the first criticality, two types of grid plates with different neutron moderation conditions (their hole spacings are 1.50  cm and 1.27  cm) were prepared. On the other hand, there is a limitation on the number of available UO₂ fuel rods. The core configurations for the first criticality satisfying these experimental constraints were designed by computational analysis. A cylindrical core configuration with a 1.50  cm pitch grid plate close to the optimum moderation condition needs 253 fuel rods to reach criticality. As to the 1.27 cm grid plate, we considered core configurations with 2.54  cm intervals by using doubled pitches of the grid plate. It will need 213 fuel rods for the criticality to be reached. In addition, the experimental core configuration was considered with steel/concrete simulant rods to simulate fuel debris conditions. This paper shows six core configurations with different conditions, and all of them satisfy the regulatory requests.

The passive containment cooling system is crucial important for marine nuclear power platform. This paper focuses on the transient characteristic of passive containment cooling system with equal height difference natural circulation under two starting strategy. Dynamic simulation model of passive containment cooling system is set up with APROS software. The simulation results indicate that the initial containment pressure, initial non-condensable gas content, and inlet water temperature have important effects on the pressure drop rate and natural circulation flow rate. And obvious differences exist between the empty tube starting mode and the full tube starting mode. Under single-phase natural circulation conditions, the pressure drop rate of the empty tube starting mode is more than 3.1 times that of the full tube starting mode, while under two-phase natural circulation conditions, the pressure drop rate of the empty tube starting mode is more than 1.5 times that of the full tube starting mode, since large fluctuation amplitude of two-phase flow instability caused by condensation induced water hammer in the natural circulation occurs under full tube starting mode.

Reactor coolant pump (RCP), the only rotating equipment in the nuclear island, directly determines the safety of the nuclear power system, has been likened to the “heart” of a nuclear power plant. The RCP needs to have very high reliability and stability. Its independent design and construction have always been the focus and difficulty of promoting the independent construction of nuclear power. To study the effect of the wear ring clearance on the energy loss characteristics of the RCP, the numerical simulation of the internal flow field of computation models with different wear ring clearance dimensions were conducted. The results indicate: As the wear ring clearance increases, the trend of performance reduction becomes more significant, especially under the large flow conditions; The increase of the wear ring clearance results in an increase of the pressure loss in the impeller and diffuser, and the variation in pressure distribution in different flow passages under different wear ring clearance dimensions is influenced by the position of the blade; The distribution of turbulent dissipation entropy production rate corresponds significantly to the distribution of direct dissipation entropy production rate, and the dissipation entropy production being more significantly affected by the wear ring clearance dimension closer to the shroud; The change of wear ring clearance dimension has limited impact on the wall shear stress entropy production of the impeller blade, while its effect on the diffuser blade is more pronounced; With the increase of the wear ring clearance, the scale of vortices near the impeller front shroud significantly increases, and the entropy production on the surface of vortex core at the impeller inlet gradually increases. This study is of great significance to ensure the secure and stable operation of the RCP, and provides an important reference for the hydraulic optimization design of the RCP.

Innovative reactors are typically designed with complex geometric structures, which put forward higher requirements for deterministic reactor core calculation. In the complex geometric reactor core calculation, the LAVENDER code in the SARAX code system was used, which is based on the three-dimensional neutron transport nodal method. A modeling method based on constructive solid geometry was applied to generate arbitrary triangular-z meshes. To meet the requirements of coupling core physics calculation with thermal and other physics fields, a flux reconstruction method based on the statistics of biharmonic spline interpolation of Green’s function has been proposed, which was used to reconstruct the flux of specified region. Three reactors with different types were calculated to verify the accuracy of the flux reconstruction method. Numerical results show that the fluxes of specified region calculated by the proposed flux reconstruction method in the SARAX code were in good agreement with the direct full-core Monte-Carlo results.

The kernel broadening approach to Doppler broadening has been shown to be equivalent to the convolution of a function related to the unheated cross section and Gaussian distribution function. This convolution may be expressed as the inverse Fourier transform of the product of the Fourier transforms of the convolution arguments. By recognizing that the unheated cross section function is an odd function, the resulting equations are reduced to a Hankel Transform of order $\alpha =\frac{1}{2}$. A discretized version of the Hankel Transform is applied to produce a set of equations that are independent of the interpolation scheme used in the pointwise representation of the cross sections. The heated cross section at a given energy is consequently calculated as the sum of the contributions from each of these pointwise intervals and each contribution is represented as a separate Hankel Transform.