Development of Approaches to Simulate Fuel Rod Destruction With Different Fuel Type

2020 ◽  
Author(s):  
Eduard Usov ◽  
Nikolay Pribaturin ◽  
Vladimir Chukhno ◽  
Ilya Klimonov ◽  
Anton Butov ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Liquid Metals ◽  
Numerical Algorithms ◽  
Severe Accident ◽  
Theoretical Research ◽  
Detailed Calculation ◽  
Fuel Rods ◽  
Fuel Rod ◽  
The Core

Abstract Due to the revival of interest to the development of fast reactors cooled by liquid metals, the problem of carrying out theoretical research in support of their safety is actual. A detailed calculation of all stages of the accident from the beginning to the end requires knowledge of the laws for modeling physical processes occurring in the reactor in an emergency. The most serious are accidents with the destruction of the core. Simulation of severe accident in nuclear reactor is the key element in safety analysis of nuclear power plants. Destruction of fuel rods is one of the most important processes that should be calculated during core degradation. For different type of fuels the mechanism of the degradation are different too. For example, oxide and metallic fuels usually melt congruently at high temperature, but nitride fuel dissociates. The main objective of the proposed research is developing of models and numerical algorithms for calculation fuel rods destruction with oxide, metallic and nitride fuels. The models of the destruction processes and some calculation results are presented in the paper. The processes are investigated for the first phase of severe accidents covering the period from the onset of fuel-rod melting to the melt escape from the core center.

scholarly journals Effect of Corrosion of the Fuel Rod Construction Materials on the Radiation Safety of Nuclear Power Plants with PWR

2020 ◽  
Vol 63 (1) ◽  
pp. 89-98
Author(s):  
V. V. Kravchenko ◽  
S. D. Tsygankova
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Construction Materials ◽  
Zirconium Alloys ◽  
General Concept ◽  
Corrosion Process ◽  
Operating Conditions ◽  
Metal Corrosion ◽  
Fuel Rods ◽  
The Core

The article considers the general concept of corrosion in accordance with GOST 5272–68 “Metal Corrosion”, the classification of the corrosion process, the stages of corrosion as energy function of the flow path of the corrosion process, the main indicators of the corrosion process. According to the forecasts of the International Monetary Fund and Focus Economics, the amount of funds that will be spent on counteracting corrosion and its consequences in selected industrialized countries has been estimated. The growth of funds invested in the counteracting the effects of metal corrosion in the Russian Federation for 2016–2019 is presented in the form of a diagram. The substantiation of the use of zirconium as a structural material for the shell of fuel rods has been fulfilled. The values of the thermal neutron absorption cross sections for various elements serving as structural elements for the core of a nuclear reactor are presented. Factors influencing the choice of alloying elements and their percentage in various alloys (Zr-2, Zr-4, ZIRLO™, M5®), which are the special development that reduce the corrosion rate, are also considered. The composition and mechanical properties of E110 and E635 alloys, which were used as materials for the fuel rods shell in the core of WWER-1200 reactors at the Belarusian NPP, are considered as well. The behavior of zirconium alloys E110 and E635 in the core is analyzed. The main factors that make a significant contribution to the corrosion process in actual operating conditions of zirconium alloys as fuel rods shell have been identified. The existing methods of preliminary special processing of fuel rods shells stored in the air for a long time before their receipt for assembly are presented. The structure of the oxide on the shells of alloys E110 and E635 oxidized in an autoclave is demonstrated.

Research and Engineering Application of Repaired Fuel Assembly Reusing Technique

2017 ◽  
Author(s):  
Lihua Wang ◽  
Qingxiang Yang ◽  
Ping Yang ◽  
Jiazheng Liu ◽  
Libing Zhu ◽  
...  
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Engineering Application ◽  
Evaluation Technique ◽  
Fuel Rods ◽  
The Core ◽  
Fuel Assemblies

Due to debris in the coolant against clad, fuel clad wear, fuel handling fault and so on, fuel rods maybe be damaged during the operation of nuclear power plants, in order that the fuel assemblies with damaged fuel rods are discharged before scheduled. If the damaged fuel assemblies are not reloaded into the core of the nuclear power plant, the fuel utilization decreases and the economy of the nuclear power plant is partly lost. For retrieving the loss of the economy, the damaged fuel assemblies can be repaired by replacing damaged fuel rods with dummy rods which don’t include fissile nuclides. Then, the repaired fuel assemblies can be reloaded into the core. As the repaired fuel assemblies are different with the normal fuel assemblies, especially the number of the damaged fuel rods is considerable, a whole quantitative analysis is very necessary to evaluate the effects from the reuse of the repaired fuel assemblies. In this paper, a full scope evaluation of reload design are performed including nuclear design, fuel design, thermal hydraulic design and safety evaluation, and some necessary improvements are done for the software system, design methods and progress which have been used in the normal reload design. As results, an integrated evaluation technique is developed to evaluate the feasibility and safety of reusing the repaired fuel assemblies, and the key effects due to the reuse of the repaired fuel assemblies are extracted, and the different effects are studied for the different materials of the dummy rods which can be used to conduct how to choose the proper material of dummy rods. In addition, this technique has been successfully applied in the engineering and the loss of economy due to the damage of fuel assemblies was retrieved partly. Therefore, the integrated evaluation technique has also important directive to other nuclear power plants if the repaired fuel assemblies are planned to reuse.

scholarly journals Analysis of criticality of melt during severe accidents in reactor vessel

2018 ◽  
pp. 3-10
Author(s):  
Yu. Kovbasenko ◽  
Yevgen Bilodid
Keyword(s):  
Boric Acid ◽  
Nuclear Power ◽  
Power Plants ◽  
Reactor Core ◽  
Cooling Water ◽  
Nuclear Fission ◽  
Reactor Vessel ◽  
Severe Accident ◽  
Emergency Situations ◽  
The Core

The article investigates the possibility of a self-sustaining chain nuclear fission reaction during the development of a severe accident in the core at nuclear power plants with reactors WWER-1000 of Ukraine. Some models for calculating a criticality at different stages of the severe accident in the reactor VVER-1000 vessel were developed and calculations of multiplication properties of fuel containing masses were performed. The severe accident in the VVER-1000 core approximately divided into seven major stages: the intact reactor core, beginning of cladding damage (swelling), cladding melting and flowing down to the support grid, melting of constructional materials, homogenization of the materials at the bottom of the reactor vessel, stratification of corium at the bottom of the reactor vessel, the exit of the corium from the reactor shaft. It was shown that at the beginning of an accident, if fuel rods geometry is maintained, criticality might appear even if the emergency protection rods is triggered. With further development of the accident, the melt of fuel and structural materials will be deeply subcritical if water cannot penetrate into the pores or voids of the melt. In the case of the formation of pores or voids in the melt and the ingress of water into them, a recriticality may arise. A compensating measure is the addition of a boric acid solution to a cooling water with a certain concentration. According to the results of the computation analysis, a reactor core loaded with TVSA fuel (Russian production) requires a higher concentration of boric acid in water to compensate the multiplication properties of the fuel system in emergency situations compared to the core loaded with TVS-WR fuel (manufactured by Westinghouse), i.e. TVS-WR fuel is safer from the criticality point of view.

scholarly journals Development of a criterion for assessment of fuel washout during operation of WWER power units

2020 ◽  
Vol 6 (4) ◽  
pp. 307-312
Author(s):  
Igor A. Evdokimov ◽  
Andrey G. Khromov ◽  
Petr M. Kalinichev ◽  
Vladimir V. Likhanskii ◽  
Aleksey A. Kovalishin ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Fission Products ◽  
Reactor Operation ◽  
Fuel Rod ◽  
The Core ◽  
Fuel Assemblies ◽  
Reliable Detection ◽  
Coolant Activity

Fuel failures may occur during operation of nuclear power plants. One of the possible and most severe consequences of a fuel failure is that fuel may be washed out from the leaking fuel rod into the coolant. Reliable detection of fuel washout is important for handling of leaking fuel assemblies after irradiation is over. Detection of fuel washout is achievable in the framework of coolant activity evaluation during reactor operation. For this purpose, 134I activity is historically used in WWER power units. However, observed 134I activity may increase during operation even if leaking fuel in the core is absent, and fuel deposits are the only source of the fission products release. The paper describes a criterion which enables to reveal the cases when the increase in 134I activity results from the fuel washout from the leaking fuel rods during operation of the WWER-type reactor. Some examples of applications at nuclear power plants are discussed.

Effect of gap design pressure on the LWR fuel rods lifetime

Kerntechnik ◽  
2021 ◽  
Vol 86 (3) ◽  
pp. 202-209
Author(s):  
M. Ghasabian ◽  
F. Mofidnakhaei ◽  
S. Talebi
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Mechanical Performance ◽  
Fission Products ◽  
Fuel Temperature ◽  
Fuel Rods ◽  
Fuel Rod ◽  
Fuel Burn ◽  
Water Reactors ◽  
Centerline Temperature

Abstract The fuel burn-up rate has been raised in recent years to improve the efficiency of nuclear LWRs (light water reactors). Therefore, surveying and estimating changes in fuel properties and structural materials during radiation exposure is of paramount importance. In the present study, the researchers focused on analyzing the role of LWR fuel rod initial gap pressure (initial gas pressure when a fuel rod is fabricated) on the rod’s thermal and mechanical performance. FRAPCON-4.0 steady-state fuel performance code was used to simulate the effect of initial gap pressure on the behavior of a specific BWR-type fuel rod that was irradiated under the HALDEN research program. This fuel rod is similar to commercial BWR fuel rods in all respects, except that the research reactors have a height limit. The important fuel design criteria, such as the centerline temperature, effective stresses, total released fission gas to the fuel rod’s void volumes, and the cladding strains, were included in the analysis. According to the present study, a potential initial gap pressure range could be suggested to increase fuel rods’ lifetime by improving the safety criteria margins, especially fuel centerline temperature and the released amount of gaseous fission products. As we know, lower fuel temperature leads to having a reactor with a higher power density and, consequently, a maximum fuel burn-up rate, which can affect the economy and safety of nuclear power plants.

scholarly journals A decision-support approach to severe accident management in nuclear power plants

2020 ◽  
pp. 1-12
Author(s):  
Marko Bohanec ◽  
Ivan Vrbanić ◽  
Ivica Bašić ◽  
Klemen Debelak ◽  
Luka Štrubelj
Keyword(s):  
Decision Support ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Severe Accident ◽  
Accident Management

scholarly journals A Conceptual Approach to Eliminate Bypass Release of Fission Products by In-Containment Relief Valve under SGTR Accident

2018 ◽  
Vol 2018 ◽  
pp. 1-12
Author(s):  
Taeseok Kim ◽  
Wonjun Choi ◽  
Joongoo Jeon ◽  
Nam Kyung Kim ◽  
Hoichul Jung ◽  
...  
Keyword(s):  
Steam Generator ◽  
Nuclear Power ◽  
Power Plants ◽  
Fission Products ◽  
Mitigation Strategies ◽  
Severe Accident ◽  
Base Case ◽  
Conceptual Approach ◽  
Relief Valve ◽  
Radioactive Nuclides

During a hypothesized severe accident, a containment building is designed to act as a final barrier to prevent release of fission products to the environment in nuclear power plants. However, in a bypass scenario of steam generator tube rupture (SGTR), radioactive nuclides can be released to environment even if the containment is not ruptured. Thus, thorough mitigation strategies are needed to prevent such unfiltered release of the radioactive nuclides during SGTR accidents. To mitigate the consequence of the SGTR accident, this study was conducted to devise a conceptual approach of installing In-Containment Relief Valve (ICRV) from steam generator (SG) to the free space in the containment building and it was simulated by MELCOR code for numerical analysis. Simulation results show that the radioactive nuclides were not released to the environment in the ICRV case. However, the containment pressure increased more than the base case, which is a disadvantage of the ICRV. To minimize the negative effects of the ICRV, the ICRV linked to Reactor Drain Tank (RDT) and cavity flooding was performed. Because the overpressurization of containment is due to heat of ex-vessel corium, only cavity flooding was effective for depressurization. The conceptual design of the ICRV is effective in mitigating the SGTR accident.

Analyzing Droplet Size Distributions Inside a Self-Priming Venturi Scrubber for Filtered Containment Venting Systems

2018 ◽  
Author(s):  
P. Papadopoulos ◽  
T. Lind ◽  
H.-M. Prasser
Keyword(s):  
Droplet Size ◽  
High Speed ◽  
Nuclear Power ◽  
Power Plants ◽  
Gas Flow ◽  
Tube Bundle ◽  
Droplet Size Distribution ◽  
Severe Accident ◽  
Design Element ◽  
Venturi Scrubber

After the accident in the Fukushima Daiichi nuclear power plant, the interest of adding Filtered Containment Venting Systems (FCVS) on existing nuclear power plants to prevent radioactive releases to the environment during a severe accident has increased. Wet scrubbers are one possible design element which can be part of an FCVS system. The efficiency of this scrubber type is thereby depending, among others, on the thermal-hydraulic characteristics inside the scrubber. The flow structure is mainly established by the design of the gas inlet nozzle. The venturi geometry is one of the nozzle types that can be found in nowadays FCVS. It acts in two different steps on the removal process of the contaminants in the gas stream. Downstream the suction opening in the throat of the venturi, droplets are formed by atomization of the liquid film. The droplets are contributing to the capture of aerosols and volatile gases from the mixture coming from the containment. Studies state that the majority of the contaminants is scrubbed within this misty flow regime. At the top of the venturi, the gas stream is injected into the pool. The pressure drop at the nozzle exit leads to the formation of smaller bubbles, thus increasing the interfacial area concentration in the pool. In this work, the flow inside a full-scale venturi scrubber has been optically analyzed using shadowgraphy with a high-speed camera. The venturi nozzle was installed in the TRISTAN facility at PSI which was originally designed to investigate the flow dynamics of a tube rupture inside a full-length scale steam generator tube bundle. The data analysis was focused on evaluating the droplet size distribution and the Sauter mean diameter under different gas flow rates and operation modes. The scrubber was operated in two different ways, submerged and unsubmerged. The aim was to include the effect on the droplet sizes of using the nozzle in a submerged operation mode.

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