[en] Highlights: • The melt pool natural convection experiments were summarized chronologically. • The main remaining issues regarding melt pool natural convection behavior in the RPV lower plenum were pointed out. • The suggestions for the future melt pool natural convection experiments were provided. - Abstract: In a hypothetical severe accident that happens in a light water reactor (LWR), the melt pool could be formed in the lower plenum of reactor pressure vessel (RPV). Remaining these molten materials inside RPV and preventing further leakage of these melt materials from RPV are treated as one of the severe accident management methods. In this aspect, one of the key issues is that the heat transfer from melt pool to RPV wall should be less than the corresponding critical heat flux (CHF) outside the vessel. The natural convection behavior in melt pools plays an important role in determining the heat flux from the melt pool to the RPV wall. Up to now, various experiments have been conducted to study the internally heated natural convection behavior in melt pools. In 1970s, several experiments were conducted, and mainly investigated the mechanism of internally heated natural convection phenomena. After TMI-2 and Chernobyl-4 severe accidents, such kinds of experiments were conducted worldwide, mainly aimed to further investigate the melt pool natural convection behavior in the lower plenum and conduct the experiment closer to the prototypical condition that may happen in a hypothetical severe accident. Some of these experiments were used to provide the data for severe accident management method assessment of their corresponding reactor types, and to solve the problems that were found or not solved by previous experiments. Especially the COPRA (COrium Pool Research Apparatus) experiment, of which the test section was the full scale and nitrate salt of which the physical properties are close to the prototypical material was used as the simulant material, improved the theoretical knowledge of the transient and steady-state phases of melt pool natural convection behavior. However, there are still uncertain and unclear phenomena related to the natural convection behavior in internally heated melt pools. This paper mainly summaries the melt pool natural convection experiments chronologically, and then points out the main remaining issues regarding the natural convection behavior in the melt pools in the RPV lower plenum.

[en] In a hypothetical severe accident, the fuel assemblies and structure in the core region could melt. The corium from the core region may fall into the reactor lower plenum. Jet breakup phenomenon would happen while corium falling into water pool in the lower plenum. In a boiling water reactor lower plenum, the control rod guide tubes (CRGTs) will create flow channels for the falling jet. Therefore corium jet breakup behavior could be affected by these CRGTs. Thus, it is important to investigate CRGTs effect on the jet breakup behavior. In order to investigate CRGTs effect on the jet breakup behavior, a molten material (U-alloy) breakup experiment was conducted under isothermal condition. The experiment results indict that CRGTs restrain the jet breakup process. For the case with CRGTs pitch/diameter ratio of 1.37, the jet breakup fraction was approximately 20% of that for the case without CRGTs. The test was also conducted for the case with a coarser pitch/diameter ratio of 2.47, but the amount of jet breakup was only slightly reduced in this configuration. The experiments also indict that the departure droplets diameter was almost not affected by CRGTs. Furthermore, the particle image velocimetry (PIV) method was used to measure water velocity distribution around the jet. The water velocities surrounding the jet for the case without CRGTs were smaller than those in the case with CRGTs. (author)

[en] Highlights: • MCNP and ORIGEN are coupled to perform nuclides depletion and decay calculation. • Coupled system MCORE uses “modified predictor corrector” approach. • MCORE can use different depletion schemes and simulate fuel shuffling. • MCORE is assessed by a “VVER-1000 LEU Assembly Computational Benchmark”. • MCORE is also assessed by a fast reactor benchmark problem. - Abstract: An MCNP–ORIGEN burn-up calculation code system, named MCORE (MCNP and ORIGEN burn-up Evaluation code), is developed in this work. MCORE makes use of the Monte Carlo neutron and photon transport code MCNP4C and nuclides depletion and decay calculation code ORIGEN2.1. MCNP and ORIGEN are coupled by data processing and linking subroutines. In MCORE, a so called “modified predictor corrector” approach is used. MCORE provides the capability of using different depletion calculation schemes and simulating fuel shuffling. Total nuclide density changes in active cells are considered in MCORE. The validity and applicability of the developed code are tested by investigating and predicting the neutronic and isotopic behavior of a “VVER-1000 LEU Assembly Computational Benchmark” at lattice level and a “Physics of Plutonium Recycling” fast reactor at core level (OECD-NEA). The comparison results show that the MCORE code predicts the nuclide composition within 5% accuracy and k_∞ within 800 pcm at the end of the burn-up for LEU assembly (40 MWD/kg HM). For a fast reactor, the results obtained by MCORE are in the range of reported results except for ²⁴³Am. In general, MCORE results show a good agreement with the benchmark values

[en] In 2011 at Fukushima Daiichi Nuclear Power Plant, which was a boiling water reactor (BWR), core meltdown occurred. There are control rod guide tubes (CRGTs) in lower plenum of pressure vessel of BWR and it is considered that CRGTs influence behavior of jet breakup of corium. In this research, the experiment about difference of behavior of jet breakup with and without pillar, which simulate CRGTs, and mechanism of difference of behavior of jet breakup in was simulated and analyzed by particle method. (author)

[en] As a new conceptual reactor, traveling wave reactor (TWR) is under fundamental research phase. The physical and mathematical models of the main parts of the primary circuit of TP-1 sodium cooled TWR designed by TerrraPower Co. were established, and the transient and safety analysis code TAST for sodium cooled TWR was preliminarily developed with Fortran program. Steady state analysis proves that the stability and dependability of TAST code for TWR are preserved. The variations of main parameters for loss of flow and reactivity insertion accidents were calculates, the results show that TWR is safe under such two transients. (authors)

[en] Highlights: • The internal heated water pool natural convection behavior is simulated. • The transient behavior of internal heated water pool is investigated. • The effect of various turbulence models on natural convection behavior is evaluated. - Abstract: In vessel retention (IVR) is treated as one of the strategies in the severe accident management aspect for the safety of nuclear reactor. The investigation of internal heated melt pool natural convection behavior is important for providing the reference for the IVR strategy. Various researchers conducted internal heated melt pool simulation with focusing on the different aspects and conditions. Based on the previous research, the large scale internal heated melt pool flow behavior located in the turbulence flow region and turbulence models used could affect the simulation result of the internal heated melt pool. Therefore, it is important to conduct the simulation of comprehensive assessment of different turbulence models on the melt pool simulation. In this work, the simulation of large scale internal heated water pool, which is one of the COPRA (COrium Pool Research Apparatus) water tests, is conducted by using ANSYS FLUENT 18.2. The large scale internal heated water pool transient behavior is investigated. The effect of various turbulence models on the internal heated water pool simulation is investigated. This work could provide better understanding of temperature distribution and velocity distribution of large scale internal heated water pool natural convection behavior from the numerical aspect. Based on the current simulation, Realizable k-epsilon turbulence model with the consideration of full buoyancy effects, RNG k-epsilon turbulence model with the consideration of full buoyancy effects, DES (Spalart-Allmaras) turbulence model and DES (SST k-omega) turbulence model are recommended for the simulation of large scale internal heated water pool. This work could bring the benefits for the further study on the numerical investigation of large scale internal heated water pool natural convection behavior.

[en] Highlights: • A theoretical model for effective thermal conductivity of porous carbon materials in TRISO particles is developed and verified. • The temperature-pressure dependence of thermal conductivity for the released gas mixture is considered. • The effects of initial porosity of buffer layer, appearance of a buffer-IPyC gap, different kernel materials and irradiation temperature are investigated. • The analysis reveals that a strong coupling relation exists among the pore pressure, the external hydrostatic pressure, the gas-induced expansion and porosity of the buffer layer. • With the developed model applied to the FE simulation, the kernel-buffer-IPyC thermo-mechanical interactions can be bridged. - Abstract: The application of fully ceramic microencapsulated (FCM) fuels to light water reactors and small modular reactors are attracting great interest. The porous carbon buffer layer is the main release volume for gas products, and it is the important part for transferring the fission heat. In this study, a theoretical model for effective thermal conductivity of porous carbon materials is developed and verified, which depends on the current porosity, the thermal conductivity of gas mixture and PyC skeleton. In this model, the temperature-pressure dependence of thermal conductivity for the released gas mixture is considered, with the fitting relation based on the existing experimental data. The porosity, the amounts of released gas mixture in the pores and the pore pressure are considered to evolve with burnup. Appearance of a buffer-IPyC gap is also involved. The effects of initial porosity of buffer layer, appearance of a buffer-IPyC gap, different kernel materials and irradiation temperature on the effective thermal conductivity are investigated. The effect of uncertainties in the thermal conductivity of dense PyC skeleton material is also analyzed. This work is expected to provide a fundamental basis for further precise simulation of the in-pile thermo-mechanical coupling behavior in FCM fuel pellets.

[en] Highlights: • A model of gas-induced effective expansion strain rate for porous carbon materials in TRISO particles is developed. • The gas-induced effective expansion strain is mainly attributed to the irradiation creep contribution of frame materials. • The effects of irradiation temperature, irradiation conditions, initial porosity and kernel materials are investigated. • TRISO particles with UN kernels turn out to be superior to those with UO₂ kernels. • The materials easy to experience creep are good candidates for frame materials of the buffer layer. - Abstract: The application of fully ceramic microencapsulated (FCM) fuels to light water reactors has attracted increasing attention. As the main storage space for fission gases and some other gases in FCM fuels, the microstructure and thickness of porous carbon buffer layer should be optimally designed. In this study, a model of gas-induced effective expansion strain rate for porous carbon materials is developed based on meso-mechanical method and homogenization theory. By combining theoretical analysis with finite element (FE) simulation, the resulting model is verified, which considers the irradiation-induced creep deformation of dense PyC skeleton, the gas contained in the pores of the buffer layer, the accumulated effective volumetric strain and the external pressure. The effects of temperature, fission rate, kernel materials, initial porosity, external hydrostatic pressure and creep coefficient of dense PyC skeleton are investigated. This work is expected to provide a theoretical reference for optimization design of a porous carbon buffer layer, and the developed model can be used for further simulation of irradiation-induced thermo-mechanical behavior in FCM fuels.

[en] Highlights: • The large scale internal heated water pool natural convection simulation is validated by using ANSYS FLUENT 18.2. • The numerical study of geometric size effect on the internal heated water pool natural convection behavior is conducted. • The Nu-Ra' correlations for the internal heated water pool are proposed based on the current simulation results. -- Abstract: The internal heated melt pool natural convection behavior is one of the important phenomena in the nuclear reactor severe accident investigation aspect. Based on the previous research, the Nusselt number is the function of the internal Rayleigh number, and the geometric size could affect the internal Rayleigh number in the order of 5. According to the previous experimental and simulation results, the relationship of Nusselt number and internal Rayleigh number varies with different tests and conditions, and the local heat transfer coefficient is also varied with the vessel polar angle. Therefore, it is important to investigate how geometric size could affect the internal heated melt pool natural convection behavior. In the present work, the validation of numerical simulation method is conducted through comparison with the COPRA water test experimental data. Then the geometric size effect on the internal heated water pool natural convection behavior is investigated by using ANSYS FLUENT 18.2 and the wall modeled large eddy simulation (WMLES) turbulence model is applied in the current simulation. The water pool temperature distribution, water pool velocity distribution, and local heat flux along the curved wall are obtained in this study. The mechanism of geometric size effect on the internal heated water pool natural convection behavior is investigated. Besides, both the upward and downward Nusselt number with internal Rayleigh number correlations are proposed based on the current simulation results. Based on the current simulation, the cases with 0.1 m, 0.3 m and 0.5 m water pool radius show the similar natural circulation behavior in the temperature distribution and heat flux distribution aspects, and the cases with 1.0 m, 1.5 m, 2.0 m and 2.5 m water pool radius show another type of natural circulation behavior. However, the cases with 0.1 m, 0.3 m, 0.5 m and 1.0 m water pool radius show the similar Nusselt number with internal Rayleigh number correlations, and the cases with 1.5 m, 2.0 m and 2.5 m water pool radius show another type of Nusselt number with internal Rayleigh number correlations. The present study could provide useful references for the severe accident management, and further improve the understanding of how geometric size could affect the internal heated water pool natural convection behavior.

[en] The results of effective irradiation swelling in a wide range of burnup levels are numerically obtained for an inert matrix fuel, which are verified with DART model. The fission gas swelling of fuel particles is calculated with a mechanistic model, which depends on the external hydrostatic pressure. Additionally, irradiation and thermal creep effects are included in the inert matrix. The effects of matrix creep strains, external hydrostatic pressure and temperature on the effective irradiation swelling are investigated. The research results indicate that (1) the above effects are coupled with each other; (2) the matrix creep effects at high temperatures should be involved; and (3) ranged from 0 to 300 MPa, a remarkable dependence of external hydrostatic pressure can be found. Furthermore, an explicit multi-variable mathematic model is established for the effective irradiation swelling, as a function of particle volume fraction, temperature, external hydrostatic pressure and fuel particle fission density, which can well reproduce the finite element results. The mathematic model for the current volume fraction of fuel particles can help establish other effective performance models