• Nuclear Engineering and Technology
• /
• v.39 no.6
• /
• pp.683-690
• /
• 2007

Because of increasing concerns regarding global warming and the longevity of oil and gas reserves, the importance of nuclear energy as a major source of sustainable energy is gaining recognition worldwide. To make nuclear energy truly sustainable, it is necessary to ensure not only the sustainability of the fuel supply but also the sustained availability of waste repositories, especially those for high-level radioactive waste (HLW). From this perspective, the effort to maximize the waste loading density in a given repository is important for easing repository capacity problems. In most cases, the loading of a repository is controlled by the decay heat of the emplaced waste. In this paper, a comparison of the decay heat characteristics of HLW is made among the various fuel cycle options. It is suggested that, for a future fast breeder reactor (FBR) cycle, the removal and burning of minor actinides (MA) would significantly reduce the heat load in waste and would allow for a reduction of repository size by half.

• Nuclear Engineering and Technology
• /
• v.55 no.5
• /
• pp.1821-1829
• /
• 2023

The temperature distribution of the reactor pool under natural circulation induced by the RVACS operation was experimentally studied. According to the Bo' based similarity law, which could reproduce the temperature distribution of the working fluid under natural circulation, SINCRO-2D facility was designed based on the PGSFR. It was reduced to 1 : 25 in length scale, having water as a simulant of the sodium, which is the original working fluid. In general, temperature was stratified, however, effect of the natural circulation flow could be observed by the entrainment of the stratified temperature. Relative cooling contribution of the upper plenum (narrow gap) and lower plenum was approximately 0.2 and 0.8, respectively. In the range of decay heat from 0.2% to 1.0%, only the magnitude of the temperature was changed, while the normalized temperature maintained. Boundary temperature distribution change made a global temperature offset of the pool, without a significant local change. Therefore, the decay heat and cooling boundary condition had no significant effect on temperature distribution characteristics of the pool within the given range of the decay heat and boundary temperature distribution.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.9 no.4
• /
• pp.237-245
• /
• 2011

As necessity of safety re-evaluation for spent fuel storage facility has emphasized after the Fukushima accident, accuracy improvement of heat load evaluation has become more important to acquire reliable thermal-hydraulic evaluation results. As groundwork, parametric and sensitivity analyses of various storage conditions for Kori Unit 4 spent fuel storage pool and spent fuel depletion parameters such as axial burnup effect, operation history, and specific heat are conducted using ORIGEN2 code. According to heat load evaluation and parametric sensitivity analyses, decay heat of last discharged fuel comprises maximum 80.42% of total heat load of storage facility and there is a negative correlation between effect of depletion parameters and cooling period. It is determined that specific heat is most influential parameter and operation history is secondly influential parameter. And decay heat of just discharged fuel is varied from 0.34 to 1.66 times of average value and decay heat of 1 year cooled fuel is varied from 0.55 to 1.37 times of average value in accordance with change of specific power. Namely depletion parameters can cause large variation in decay heat calculation of short-term cooled fuel. Therefore application of real operation data instead of user selection value is needed to improve evaluation accuracy. It is expected that these results could be used to improve accuracy of heat load assessment and evaluate uncertainty of calculated heat load.

• Nuclear Engineering and Technology
• /
• v.53 no.1
• /
• pp.44-60
• /
• 2021

In this paper, we validate the decay heat calculation capability via a two-step method to analyze spent nuclear fuel (SNF) discharged from pressurized water reactors (PWRs). The calculation method is implemented with a lattice code STREAM and a nodal diffusion code RAST-K. One of the features of this method is the direct consideration of three-dimensional (3D) core simulation conditions with the advantage of a short simulation time. Other features include the prediction of the isotope inventory by Lagrange non-linear interpolation and the use of power history correction factors. The validation is performed with 58 decay heat measurements of 48 fuel assemblies (FAs) discharged from five PWRs operated in Sweden and the United States. These realistic benchmarks cover the discharge burnup range up to 51 GWd/MTU, 23.2 years of cooling time, and spanning an initial uranium enrichment range of 2.100-4.005 wt percent. The SNF analysis capability of STREAM is also employed in the code-to-code comparison. Compared to the measurements, the validation results of the FA calculation with RAST-K are within ±4%, and the pin-wise results are within ±4.3%. This paper successfully demonstrates that the developed decay heat calculation method can perform SNF back-end cycle analyses.

• The KSFM Journal of Fluid Machinery
• /
• v.17 no.6
• /
• pp.109-114
• /
• 2014

After the Fukushima accident in 2011, it was revealed that nuclear power plant has the vulnerability to SBO accident and its extension situation without sufficient cooling of reactor core resulting core meltdown and radioactive material release even after reactor shutdown. Many safety systems had been developed like PAFS, hybrid SIT, and relocation of RPV and IRWST as a part of steps for the Fukushima accident, however, their applications have limitation in the situation that supply of feedwater into reactor is impossible due to high pressure inside reactor pressure vessel. The concept of hybrid heat pipe with control rod is introduced for breaking through the limitation. Hybrid heat pipe with control rod is the passive decay heat removal system in core, which has the abilities of reactor shutdown as control rod as well as decay heat removal as heat pipe. For evaluating the cooling performance hybrid heat pipe, a commercial CFD code, ANSYS-CFX was used. First, for validating CFD results, numerical results and experimental results with same geometry and fluid conditions were compared to a tube type heat pipe resulting in a resonable agreement between them. After that, wall temperature and thermal resistances of 2 design concepts of hybrid heat pipe were analyzed about various heat inputs. For unit length, hybrid heat pipe with a tube type of $B_4C$ pellet has a decreasing tendency of thermal resistance, on the other hand, hybrid heat pipe with an annular type $B_4C$ pellet has an increasing tendency as heat input increases.

• Nuclear Engineering and Technology
• /
• v.31 no.6
• /
• pp.66-82
• /
• 1999

The national long-term R&D program, updated in 1997, requires Korea Atomic Energy Research Institute(KAERI) to complete by the year 2006 the basic design of Korea Advanced Liquid Metal Reactor(KALIMER), along with supporting R&D work, with the capability of resolving the issue of spent fuel storage as well as with significantly enhanced safety. KALIMER is a 150 MWe pool-type sodium cooled prototype reactor that uses metallic fuel. The conceptual design is currently under way to establish a self-consistent design meeting a set of major safety design requirements for accident prevention. Some of the current emphasis includes those for inherent and passive means of negative reactivity insertion and decay heat removal, high shutdown reliability, prevention of and protection from sodium chemical reaction, and high seismic margin, among others. All of these requirements affect the reactor design significantly and involve extensive supporting R&D programs. This paper summarizes some of the results of conceptual engineering and design analyses performed for the safety of HAMMER in the area of inherent safety, passive decay heat removal, sodium water reaction, and seismic isolation.

• Nuclear Engineering and Technology
• /
• v.51 no.3
• /
• pp.654-664
• /
• 2019

The geometrical and electromagnetic variables of a rectangular-type magnetohydrodynamic (MHD) circulation system are optimized to solve MHD equations for the active decay heat removal system of a prototype Gen-IV sodium fast reactor. Decay heat must be actively removed from the reactor coolant to prevent the reactor system from exceeding its temperature limit. A rectangular-type MHD circulation system is adopted to remove this heat via an active system that produces developed pressure through the Lorentz force of the circulating sodium. Thus, the rectangular-type MHD circulation system for a circulating loop is modeled with the following specifications: a developed pressure of 2 kPa and flow rate of $0.02m^3/s$ at a temperature of 499 K. The MHD equations, which consist of momentum and Maxwell's equations, are solved to find the minimum input current satisfying the nominal developed pressure and flow rate according to the change of variables including the magnetic flux density and geometrical variables. The optimization shows that the rectangular-type MHD circulation system requires a current of 3976 A and a magnetic flux density of 0.037 T under the conditions of the active decay heat removal system.

• Nuclear Engineering and Technology
• /
• v.31 no.3
• /
• pp.276-286
• /
• 1999

A computer code PARS2 is developed for the analysis of PSDRS, which is the safety grade RHRS of HAMMER, and applied to the investigation of the relation between design parameters and performance of PSDRS. The concept of the heat transfer resistance network is applied in assessing the importance of the various heat transfer modes. From the analysis results, the qualitative relations between the PSDRS performance and design parameters are found and guidelines for the PSDRS design procedures are also proposed.

• Nuclear Engineering and Technology
• /
• v.39 no.2
• /
• pp.143-148
• /
• 2007

Spent nuclear fuel generated at nuclear power plants must be safely stored during interim storage periods. A dry storage cask to safely store the spent nuclear fuel should be able to adequately emit the decay heat from the spent nuclear fuel. Therefore, heat removal tests using a half scale dry storage cask have been performed to estimate the heat transfer characteristics of a dry storage cask under normal, off-normal, and accident conditions. In the normal condition, the heat transfer rate to an ambient atmosphere by convective air through a passive heat removal system reached 83%. Accordingly, the passive heat removal system is designed well and works adequately. In the off-normal condition, the influence of a half blockage in the inlet on the temperature appears minimal. In the accident condition, the temperature rose for 12 hours after the accident, but the temperature rise steadied after 36 hours.

• Ocean and Polar Research
• /
• v.25 no.1
• /
• pp.63-74
• /
• 2003

The build-up of the heat field in shallow coastal water due to a point source has been investigated using an analytical solution of a time-integral form derived by extending the solutions by Holley(1969) and also presented in Harleman (1971). The uniform water depth is assumed with non-isotropic turbulent dispersion. The alongshore-flow is assumed to be uni-directional, spatially uniform and oscillatory. Due to the presence of the oscillatory alongshore-flow, the heat build-up occurs in an oscillatory manner, and the excess temperature thereby fluctuates in that course and even in the quasi-steady state. A series of calculations reveal that proper choices of the decay coefficient as well as dispersion coefficients are critical to the reliable prediction of the excess temperature field. The dispersion coefficients determine the absolute values of the excess temperature and characterize the shoreline profile, particularly within the tidal excursion distance, while the decay coefficient determines the absolute value of the excess temperature and the convergence rate to that of the quasi-steady state. Within the e-folding time scale $1/k_d$ (where $k_d$ is the heat decay coefficient), heat build-up occurs more than 90% of the quasi-steady state values in a region within a tidal excursion distance (L), while occurs increasingly less the farther we go to the downstream direction (about 80% at 1.25L, and 70% at 1.5L). Calculations with onshore and offshore discharges indicate that thermal spreading in the direction of the shoreline is reduced as the shoreline constraint which controls the lateral mixing is reduced. The importance of collecting long-term records of in situ meteorological conditions and clarifying the definition of the heat loss coefficient is addressed. Interactive use of analytical and numerical modeling is recommended as a desirable way to obtain a reliable estimate of the far-field excess temperature along with extensive field measurements.