• Advances in Energy Research
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• v.5 no.2
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• pp.91-105
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• 2017

Thermal hydraulic (TH) analysis of nuclear power reactors is utmost important. In this way, the numerical codes that preparing TH data in reactor core are essential. In this paper, a subchannel analysis of a Russian pressurized water reactor (WWER1000) core with enhanced numerical code is carried out. For this, in fluid domain, the mass, axial and lateral momentum and energy conservation equations for desired control volume are solved, numerically. In the solid domain, the cylindrical heat transfer equation for calculation of radial temperature profile in fuel, gap and clad with finite difference and finite element solvers are considered. The dependence of material properties to fuel burnup with Calza-Bini fuel-gap model is implemented. This model is coupled with Isotope Generation and Depletion Code (ORIGEN2.1). The possibility of central hole consideration in fuel pellet is another advantage of this work. In addition, subchannel to subchannel and subchannel to rod connection data in hexagonal fuel assembly geometry could be prepared, automatically. For a demonstration of code capability, the steady state TH analysis of a the WWER1000 core is compromised with Thermal-hydraulic analysis code (COBRA-EN). By thermal hydraulic parameters averaging Fuel Assembly-to-Fuel Assembly method, the one sixth (symmetry) of the Boushehr Nuclear Power Plant (BNPP) core with regular subchannels are modeled. Comparison between the results of the work and COBRA-EN demonstrates some advantages of the presented code. Using the code the thermal modeling of the fuel rods with considering the fission gas generation would be possible. In addition, this code is compatible with neutronic codes for coupling. This method is faster and more accurate for symmetrical simulation of the core with acceptable results.

• Nuclear Engineering and Technology
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• v.23 no.2
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• pp.165-173
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• 1991

The recycle of PWR spent fuels in a CANDU reactor, so called the tandem fuel cycle is Investigated in this study. This scheme of utilizing Pm spent fuels will ease the shortage of spent fuel storage capacity as well as will improve the use of uranium resources. The minimum modification to the design of present CANDU reactor is seeked in the recycle. Nine different fuel types are considered in this work and are classified into two categories： refabrication and reconfiguration For refabrication, PWR spent fuels are processed and refabricated into the present 37 rod lattice structure of fuel bundle, and for reconfiguration, meanwhile, spent fuels are simply disassembled and rods are cut to fit into the present grid configuration of fuel bundle without refabrication. For each fuel option, the neutronics calculation of lattice was conducted to evaluate the allowable burnup and power distribution. The fuel cycle cost of each option was also computed to assess the economic justification. The result show that most tandem fuel cycle options considered in this study are technically feasible as well as economically viable.

• Nuclear Engineering and Technology
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• v.25 no.1
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• pp.166-177
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• 1993

A simplified thermal analysis method to evaluate the maximum temperature of the CANDU 37-element fuel bundle within a fuel basket in a given spent fuel dry storage canister has been presented along with the results of sample analyses performed to examine the parametric effects of the ambient conditions on the maximum fuel temperature within a canister. To solve the multi-dimensional heat transfer problem of the complex geometry of rod bundles within a canister where three modes of heat transfer are superimposed, the CANDU spent fuel bundles stored in the dry storage canister are first replaced by equivalent concentric fuel cylinders. The simplified axi-symmetric two-dimensional multi-mode heat transfer problem of the equivalent fuel cylinders is then analyzed with an existing computer code, HEATING5, using additional input data and heat transfer correlations. A comparison between the predicted temperature profile and the mock-up test results shows that the agreement is quite satisfactory.

• Nuclear Engineering and Technology
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• v.56 no.3
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• pp.973-979
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• 2024

In the framework of the Generation IV research and development project, in which the French Commission of Alternative and Atomic Energies (CEA) is involved, a main objective for the design of Sodium-cooled Fast Reactor (SFR) is to meet the safety goals for severe accidents. Among the severe ones, the Unprotected Transient OverPower (UTOP) accidents can lead very quickly to a global melting of the core. UTOP accidents can be considered either as slow during a Control Rod Withdrawal (CRW) or as fast. The paper focuses on fast UTOP accidents, which occur in a few milliseconds, and three different scenarios are considered: rupture of the core support plate, uncontrolled passage of a gas bubble inside the core and core mechanical distortion such as a core flowering/compaction during an earthquake. Several levels and rates of reactivity insertions are also considered and the thermal-mechanical behavior of an ASTRID fuel pin from the ASTRID CFV core is simulated with the GERMINAL code. Two types of fuel pins are simulated, inner and outer core pins, and three different burn-up are considered. Moreover, the feedback from the CABRI programs on these type of transients is used in order to evaluate the failure mechanism in terms of kinetics of energy injection and fuel melting. The CABRI experiments complete the analysis made with GERMINAL calculations and have shown that three dominant mechanisms can be considered as responsible for pin failure or onset of pin degradation during ULOF/UTOP accident: molten cavity pressure loading, fuel-cladding mechanical interaction (FCMI) and fuel break-up. The study is one of the first step in fast UTOP accidents modelling with GERMINAL and it has shown that the code can already succeed in modelling these type of scenarios up to the sodium boiling point. The modeling of the radial propagation of the melting front, validated by comparison with CABRI tests, is already very efficient.

• Nuclear Engineering and Technology
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• v.24 no.3
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• pp.263-273
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• 1992

The study on the velocity distribution and the pressure drop characteristic of the nuclear fuel assembly is of importance for the thermal hydraulic design and safety analysis. The purpose of this experimental study is to investigate the hydraulic mixing behind the different kinds of spacer grids in the now or rod bundles. In this study, the detailed hydraulic characteristics in subchannels of 5$\times$5 PWR(Pressurized Water Reactor) rod bundles were measured using one-component He-Ne LDV(Laser Doppler Velocimeter). Measurements of the axial velocity, turbulent intensities and pressure drops were peformed Lateral velocity, turbulent intensities and Reynolds shear stress were also measured by adjust-ing LDV alignment. Friction factors in rod bundles and loss coefficients for spacer grids were evaluated from the measured pressure drops. Hydraulic mixing performance for different kinds of spacer grids could be investigated by estimating the turbulent cross-flow mixing rates between neighboring subchannels.

• Nuclear Engineering and Technology
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• v.25 no.1
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• pp.8-19
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• 1993

The natural convection model of the consolidated system has been developed to make sure the removal of decay heat generated in the spent fuel for the loss of forced cooling accident. The numerical technique employed was based on the ADI scheme. The calculation of heat generation rate in the spent fuel was peformed by the ANS-79 decay heat model, and the nonuniform surface heat flux is assumed with a chopped sine curve for the conservative decay heat generation input. The sensitivity study was performed to examine the possibility of the pool bulk boiling by varying the various parameters, i.e. inter-fuel spacing ratio, heat generation power, and radius of the fuel rod. The application results of this model show that the natural circulation flow through compacted spent fuel bundles enables the pool temperature to control in a safe and effective manner, after the required cooling time. The corresponding acceptance criteria of the cooling time for rearranging the spent fuel rods were also found.

• Nuclear Engineering and Technology
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• v.15 no.4
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• pp.248-255
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• 1983

The thermal analysis on the spent fuel shipping cask for a PWR fuel assembly is performed. Under the normal and fire-accident conditions the temperature distribution through a multilayer cask calculated in compliance with 10 CFR Part 71. A KNU 5&6 spent fuel assembly is assumed to be the decay heat source, which has the maximum discharge turnup of 45, 000MWD/MTU and has been stored in the spent fuel storage pool for 300 days. As a result of thermal analysis, the maximum cladding temperature in case of dry cavity under fire-accident conditions is calculated to be 455$^{\circ}C$. This value is much less than the limiting value specified in 10 CFR Part 50.46. It indicates that no fuel rod cladding rupture could occur under fire-accident conditions. It was also found that no melting of lead would take place in the major shield region.

• Nuclear Engineering and Technology
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• v.22 no.1
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• pp.66-74
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• 1990

The thermal-hydraulic characteristics of the 17$\times$17 OFA (Optimized Fuel Assembly) used in the KNU 7&8 are analyzed and compared with that of the 17$\times$17 SFA (Standard Fuel Assembly) loaded in the KNU 5&6. The thermal-hydraulic characteristics analyzed are minimum DNBR, fuel centerline temperature and exit void fraction at normal operation and design over power transient. Additionally, local linear rod power, which will cause fuel centerline melting, is calculated. The DNBR sensitivity calculations are performed with respect to the reactor operating parameters. COBRA-IV-I code is used for these calculations. The modified W-3 correltion and the drift-flux model are applied for the critical heat flux calculation and the void fraction calculation, respectively. From the calculated results, it has been found that the possibility of DNB occurrence is higher in the OFA than in the SFA. The other hand, the local linear power resulting in fuel centerline moiling of the OFA is nearly equal to that of the SFA.

• Proceedings of the Korean Society of Precision Engineering Conference
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• 2007.11a
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• pp.439-440
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• 2007

• Transactions of the Korean Society of Mechanical Engineers
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• v.8 no.1
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• pp.48-56
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• 1984

In order to estimate the fuel rod temperature during the reflooding phase of the PWR LOCA, it is essential to obtain a better understanding of the rewetting mechanism. This paper presents the results of analytical and experimental investigations aimed at analyzing the rewetting phenomena in a heated tube. A two-dimensional solution of the rewetting for a tube is described and used to yield the correlation of the rewetting heat transfer coefficient as the function of flooding rate and inlet subcooling. This correlation prediction is in good agreement with the experimental data.