Drop Impact Analysis of Plate-Type Fuel Assembly in Research Reactor (original) (raw)
Related papers
Nuclear fuel rod fragmentation under accidental conditions
h i g h l i g h t s Fuel rods can be fragmented in case of reflooding during a core meltdown accident. The core coolability strongly depends on the extent of fuel rod fragmentation. Evaluation of the size and of the surface area of the fuel rod fragments. Fuel cracking in normal conditions leads to mean particle diameter higher than 2 mm. Lower diameter in case of additional fragmentation of highly irradiated fuel rods. a b s t r a c t This paper deals with fuel rod fragmentation during a core meltdown accident in a Nuclear Power Plant. If water is injected on the degraded core to stop the degradation, embrittled fuel rods may crumble to form a reactor debris bed. The size and the morphology of the debris are two key parameters which determine in particular heat transfer and flow friction in the debris bed and as a consequence its coolability. To address this question, a bibliographic survey is performed with the aim of evaluating the size and the surface area of the fragments resulting from fuel rod fragmentation. On this basis, a model to estimate the mean particle diameter obtained in a reflooded degraded core is proposed. Modelling results show that the particle size distribution is very narrow if we only take into account fuel cracking resulting from normal operating conditions. It leads to minimum mean diameters of 2.5 mm (for fuel particles), 1.35 mm (for cladding particles) and 2 mm (for the mixing of cladding and fuel fragments). These results are obtained with fuel rods of 9.5 mm outer diameter and cladding thickness of 570 m. The particle size distribution is larger if fine fragmentation of the highly irradiated fuel rods during temperature rise is accounted for. This is illustrated with the computation by the severe accident code ASTEC, codeveloped by IRSN abd GRS, of the size of the debris expected to form in case of reflooding of a French 900 MW reactor core during a core meltdown accident.
Shielding Considerations for Advanced Fuel Irradiation Experiments
Journal of Nuclear Science and Technology, 2008
An in-pile test program for the development of a high burn-up fuel is planned for the HANARO reactor. The source term originates from a leakage of fission products from the anticipated failed fuels into the gas flow tubes and around the instrumentation and control system. In order to quantify the fuel composition in the event of a fuel failure, the isotope generation and depletion code ORIGEN 2.0 was used. The computer program Microshield 6.2 was used to calculate the doses from specific locations, where a high radioactivity is expected during an irradiation. The results indicate that the equivalent dose in the investigated working areas is less than the permitted dose rate of 6.25 Sv/hr. However, access to the area of a decay vessel may need to be limited, and the installation of a Pb wall with a 20.5 cm thickness is recommended. From the analysis of a radioactive decay with time, most of the concerned gaseous nuclides with short half-lives after 3 months, were decayed, with one exception which was Kr-85, thus it should be released in accordance with applicable government laws after measuring its activity in individual holding vessels.
Radiological effect of cask‑drop accident in open‑pool spent fuel storage
Radiation Protection and Environment, 2019
This study investigates the radiological risk that may occur during transport of cobalt device after irradiation in open‑pool‑type reactor from the reactor core to the cobalt cell. The cobalt transport process depends on using heavy shielded cask of 3500 kg weight under the water surface of the spent fuel storage pool which may cause a load drop accident. The load drop accident in the spent fuel storage pool would result in damage of 48 spent fuel elements (FE) maximally. Conservative evaluation for the amount of fission products release from the damage of the spent FE was considered in this study by assuming that all spent FE was recently stored at the time of accident. Consequently, the resulting radiation dose distribution was calculated around the reactor building depending on the meteorological data of the reactor site. GENII‑2 code was used to estimate the individual effective dose distribution around the reactor. The result shows that the receptor who located at 1500 m from the reactor building in the south‑east direction will receive the maximum individual effective dose of 0.11 Sv.
2016
In this paper we present a brief description of the last version of the nuclear fuel performance code DIONISIO. Designed to describe most of the main phenomena occurring in a fuel rod during operation of the power reactor, the code is under continued development in the Code and Models Section of the Nuclear Fuel Cycle Management of CNEA. The new version of the code was improved with the incorporation of calculation tools designed to extend the application range of the code to high burnup and accident condition across the participation of two international projects sponsored by IAEA.
Safety Analysis of Irradiated Nuclear Fuel Transportation Container
11th International Conference on Environmental Remediation and Radioactive Waste Management, Parts A and B, 2007
At present the design basis accidents for RBMK-1500 are rather thoroughly investigated. The performed analyses helped to develop and implement a number of safety modifications. Further plant safety enhancement requires developing emergency procedures that would enable beyond design basis accidents management by preventing core damage or mitigating consequences of severe accidents.
Nuclear Engineering and Design, 2020
Fuel of fast reactors is designed in the form of cylindrical pellets of solid or annular geometry. The annular configuration provides advantages in terms of higher achievable energy extraction and lower fuel-clad mechanical interaction. Another advantage of this configuration is the hydrodynamic flow of molten fuel inside the annular pellets during accidental meltdown, known as in-pin fuel motion. This motion can provide safety benefits during an unprotected transient overpower accident (UTOPA), if the molten fuel is dispersed significantly away from the core mid-plane. The physics behind the flow is complex. Intricate theoretical modelling and detailed experimental validation are prerequisites for the reliable estimation of safety benefits. To meet this requirement, a Multi-phase In-pin Thermal hydraulic Relocation Algorithm (MITRA) is developed under the purview of the pre-disassembly analysis code, PREDIS. This paper presents a theoretical foundation of the algorithm followed by experimental verifications, burnup and top blanket design sensitivity analyses and whole core UTOPA simulations. Results show that regardless of the fuel burnup level, the melt tends to agglomerate into a column, slightly below the core mid-plane. Upon further melting, this column grows slowly, devoid of the rapid dispersion associated with fuel squirting. Minor deviations in the melt position arise due to variations in thermal parameters with burnup. UTOPA analysis shows a reduction in the safety feedback and increased melting. Modifying the conventionally solid top axial blanket to annular geometry enhances the safety feedback only after 34.2% melting.
A model for release of fission products from a breached fuel plate under wet storage
Progress in Nuclear Energy, 2008
MTR fuel elements burned-up inside the core of nuclear research reactors are stored worldwide mainly under the water of storage pools. When cladding breach is present in one or more fuel plates of such elements, radioactive fission products are released into the storage pool water. This work proposes a model to describe the release mechanism considering the diffusion of nuclides
Comparative Radiation Dose Study of a Hypothetical Accident in a Research Reactor
Kemija u industriji
This study is a contribution for radiation dose calculations of a hypothetical accident of a 1 MW research reactor Triga Mark II using HotSpot code. A postulated accidental release of noble gases and halogens were considered. The total effective dose (TED) was estimated for 1 day and 50 years after release. The total damage of fuel element cladding with a maximum radioactivity was considered. The obtained results show minimal TED values at the beginning of the release and at a shorter distance from the source. The maximum calculation results are acceptable and below the recommended public dose limit.
Numerical and experimental analysis of the impact of a nuclear spent fuel cask
Nuclear Engineering and Design, 2010
This paper deals with the numerical and experimental analyses of a shell type shock absorber for a nuclear spent fuel cask. Nine-meter free drop tests performed on reduced scale models are described. The results are compared with numerical simulations performed with FEM computer codes, considering reduced scale models as well as the prototype. The paper shows the results of a similitude analysis, with which the data obtained by means of the reduced scale models can be extrapolated to the prototype. Small discrepancies were obtained using large-scale models (1:2 and 1:6), while small-scale models (1:12) did not give reliable results. A 1:9 scale model provided useful information with a less than 20% error.
Progress in Nuclear Energy, 2018
In CANDU type reactors there are two situations for irradiated fuel bundles to remain accidentally in the air. One is represented by End Fitting Failure accident when one or more fuel bundles are ejected from the channels during the fuel handling process. The second one is determined by the uncovering of fuel bundles in the Spent Fuel Pool (SFP) due to a partial or total loss of cooling followed by a progressive evaporation of the water. In both cases there is a potential for the release of radio-nuclides that is dependent on the irradiation history and the duration between unloading of the bundle from the core and the entering into the atmosphere. The paper is devoted to investigate the release in these cases. The inventory at the moment of the accidentally entering into the atmosphere is calculated by ORIGEN code. The source term into the containment, and respectively in the SFP building, is obtained by treating the fuel degradation and the release from the fuel elements by ICARE module of ASTEC code. The most important influence factors on the source term formation (such as the velocity of the ventilation, the residence time) are identified and discussed. In a realistic CANDU SFP loss of cooling accident the potential of release is reduced due to relative fuel low burnup, long residence time, and specific configuration of the storage. The uncovering of the first plane of fuel bundles needs around 15 days for the heat-up and evaporation of water. In such case, a small convection (v ≥ 1.E−05 m/s) is enough to keep the temperature of the claddings under the failure thresholds. The paper identifies and investigates some other cases (with violation of the operational procedures) with a relevant release of fission products. The results are presented for the Caesium element.