scholarly journals Determination of 18-month fuel cycle parameters for the purpose of fuel costs minimization at the basis of use constructions of fuel assemblies in VVER-1200 reactors

2019 ◽  
Vol 5 (1) ◽  
pp. 9-15
Author(s):  
Taha M. Hashlamoun ◽  
Sergey B. Vygovsky ◽  
Sergey T. Leskin ◽  
A. Safa Duman
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Fuel Cycle ◽  
Life Time ◽  
Reactor Vessel ◽  
Reactor Fuel ◽  
The Core ◽  
Fuel Assemblies ◽  
Fuel Costs

This article presents the results of research, that were focused on determining the optimal parameters of the extension of (reactor life-time) reactor fuel cycle in order to reduce the total operating costs of nuclear power plants during the transition from 12-month reactor fuel cycle to 18-month fuel cycle. The relevance of the research is related to the fact that, in recent years, there is a transition at all operating nuclear power plants VVER-1000 (1200) from 12-month reactor fuel cycle to extended 18-month fuel cycle. At the same time, represent the interests to solve the problem of conservation the extension of reactor life-time while reducing the number of loaded fuel assemblies with fresh fuel assemblies, which would reduce the total operating, and fuel costs. Search for solutions of this problem is associated with mandatory implementation of all requirements for the safe operation of the reactor and the reduction of the maximum fast neutron fluence on the reactor vessel in comparison with its value at the operating nuclear power plants. In the present work, with using the program PROSTOR software complex researched the neutron-physical characteristics of the core at the nominal parameters of the VVER-1200 reactor through the implementation of various fuel cycle strategies. The article developed various schemes of fuel-reloading for an 18-month fuel cycle with a different number of fuel assemblies. The article carries out a comparative analysis of the main parameters in the core for fuel-reloading schemes options of an 18- and 12-month fuel cycle with each other. Determine the minimum amount of fuel assemblies and provide the necessary duration of the reactor life-time for 18-month fuel cycle with using the extension of reactor life-time by reducing the power at the end of the reactor cycle to 70% of the nominal power. In the article, the arrangements of fuel assemblies were developed to provide limitations of local power by volume of the core, which reduce the fluence of fast neutrons on the reactor vessel in comparison with the projected value of the fluence. This article shows that the 18-month fuel cycle for the VVER-1200 reactor is more economical than the 12-month fuel cycle. These studies were carried out for the VVER-1200 reactor at the power of 100% of the nominal.

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 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.

scholarly journals Temperature conditions in the RBMK spent fuel pool in the event of disturbances in its cooling mode

2021 ◽  
Vol 7 (1) ◽  
pp. 9-13
Author(s):  
David A. Hakobyan ◽  
Victor I. Slobodchuk
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Spent Nuclear Fuel ◽  
Maximum Temperature ◽  
Spent Fuel ◽  
Cooling Mode ◽  
Temperature Conditions ◽  
Fuel Assemblies ◽  
Spent Fuel Pool

The problems of reprocessing and long-term storage of spent nuclear fuel (SNF) at nuclear power plants with RBMK reactors have not been fully resolved so far. For this reason, nuclear power plants are forced to search for new options for the disposal of spent fuel, which can provide at least temporary SNF storage. One of the possible solutions to this problem is to switch to compacted SNF storage in reactor spent fuel pools (SFPs). As the number of spent fuel assemblies (SFAs) in SFPs increases, a greater amount of heat is released. In addition, no less important is the fact that a place for emergency FA discharging should be provided in SFPs. The paper presents the results of a numerical simulation of the temperature conditions in SFPs both for compacted SNF storage and for emergency FA discharging. Several types of disturbances in normal SFP cooling mode are considered, including partial loss of cooling water and exposure of SFAs. The simulation was performed using the ANSYS CFX software tool. Estimates were made of the time for heating water to the boiling point, as well as the time for heating the cladding of the fuel elements to a temperature of 650 °С. The most critical conditions are observed in the emergency FA discharging compartment. The results obtained make it possible to estimate the time that the personnel have to restore normal cooling mode of the spent fuel pool until the maximum temperature for water and spent fuel assemblies is reached.

scholarly journals Small nuclear power plants for power supply in arctic regions: assessment of spent nuclear fuel radioactivity

2018 ◽  
Vol 4 (2) ◽  
pp. 119-125
Author(s):  
Vadim Naumov ◽  
Sergey Gusak ◽  
Andrey Naumov
Keyword(s):  
Nuclear Fuel ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Fuel Cycle ◽  
Mass Composition ◽  
Published Data ◽  
Pressurized Water ◽  
Fuel Cycles ◽  
Reactor Cores

The purpose of the present study is the investigation of mass composition of long-lived radionuclides accumulated in the fuel cycle of small nuclear power plants (SNPP) as well as long-lived radioactivity of spent fuel of such reactors. Analysis was performed of the published data on the projects of SNPP with pressurized water-cooled reactors (LWR) and reactors cooled with Pb-Bi eutectics (SVBR). Information was obtained on the parameters of fuel cycle, design and materials of reactor cores, thermodynamic characteristics of coolants of the primary cooling circuit for reactor facilities of different types. Mathematical models of fuel cycles of the cores of reactors of ABV, KLT-40S, RITM-200M, UNITERM, SVBR-10 and SVBR-100 types were developed. The KRATER software was applied for mathematical modeling of the fuel cycles where spatial-energy distribution of neutron flux density is determined within multi-group diffusion approximation and heterogeneity of reactor cores is taken into account using albedo method within the reactor cell model. Calculation studies of kinetics of burnup of isotopes in the initial fuel load (235U, 238U) and accumulation of long-lived fission products (85Kr, 90Sr, 137Cs, 151Sm) and actinoids (238,239,240,241,242Pu, 236U, 237Np, 241Am, 244Cm) in the cores of the examined SNPP reactor facilities were performed. The obtained information allowed estimating radiation characteristics of irradiated nuclear fuel and implementing comparison of long-lived radioactivity of spent reactor fuel of the SNPPs under study and of their prototypes (nuclear propulsion reactors). The comparison performed allowed formulating the conclusion on the possibility in principle (from the viewpoint of radiation safety) of application of SNF handling technology used in prototype reactors in the transportation and technological process layouts of handling SNF of SNPP reactors.

The Application of an Integrated Head Assembly for Advanced Power Reactor 1400

2016 ◽  
Author(s):  
Tae Kyo Kang ◽  
Won Ho Jo ◽  
Yeon Ho Cho ◽  
Sang Gyoon Chang ◽  
Dae Hee Lee
Keyword(s):  
United Arab Emirates ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Power Reactor ◽  
Reactor Vessel ◽  
Head Region ◽  
The Usa ◽  
Head Assembly ◽  
Storage Area

The reactor vessel head region consists of a number of components and systems including reactor vessel head, CEDMs with their cables, cooling air system with ducts and fans, missile shield, seismic supports, head lift rig and cable supports. Prior to refueling operation, those components must be dismantled separately, and moved to the designated storage area. It was a very complicated and time consuming process. As a result, the integrated head assembly (IHA) was introduced to simplify those disassembling procedures, reduce refueling outage period, and improve safety in the containment building as those components are combined into a single system. To reduce refueling outage duration and radiation exposures to the workers by integrating the complicated reactor head region structures, KEPCO E&C has developed the IHA concept in the Korean Next Generation Reactor (KNGR) project [1]. The first application was implemented for the Optimized Power Reactor 1000 (OPR1000) at Shin-Kori units 1&2 and Shin-Wolsong units 1&2. With the past experience, the IHA was upgraded to be applied to the Advanced Power Reactor 1400 (APR1400). The design was patented in Korea [2], China, EU and the USA as modular reactor head area assembly. The IHA was applied for APR1400 nuclear power plants at Shin-Kori and Shin-Hanul, Korea. The design was also supplied to Barakah Nuclear Power Plants in the United Arab Emirates. This paper presents the design features and a variety of analysis which have been used for the APR1400 IHA.

Life Time Management: A Practical Approach of Nuclear Power Plants

2002 ◽  
Author(s):  
Jaroslav Bartonicek ◽  
Klaus-Juergen Metzner ◽  
Friedrich Schoeckle
Keyword(s):  
Time Management ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Mechanical Systems ◽  
Life Time ◽  
Practical Approach ◽  
Degradation Mechanisms ◽  
High Level ◽  
Quality Status

A comprehensive life time management has to take care of all safety and availability relevant components in nuclear power plants, with different intensity, of course. For instance, mechanical systems and components can be basically classified/ranked into three different groups: (1): The quality status of the components in this group has to be guaranteed on a pre-defined (high) level. (2): The quality status of the components in this group has to be maintained on its actual level. (3): Other components with no specific quality demands. Regarding the first group, integrity has to be guaranteed. Therefore it is necessary to monitor the possible root causes of degradation mechanisms during plant operation; thus the degradation effects can be assessed and — more important — controlled to maintain the safety standard on the demanded high level without any compromise. The monitoring of consequences of degradation mechanisms is being performed as an additional redundant measure. The requirements to maintain the quality status of the second group of components can be fulfilled by monitoring of the consequences of operational degradation mechanisms to be performed by preventive maintenance activities, in terms of tests, inspections and repairs, using either time dependant procedures or component condition orientated methods. For the third group of components, no preventive action is necessary. However, failures and malfunctions have to be assessed statistically to avoid a reduction of the required basic component quality. In the first two groups all safety relevant components and systems are included. Generally, aging management programs cover these two groups of components; life time management covers all of above groups. This paper concentrates on mechanical systems and components; it summarizes the practical approach to life time management as it is realized in German nuclear power plants. The application is discussed using dedicated examples.

scholarly journals HEALTH IN THE SYSTEM OF VALUES OF THE POPULATION OF THE SURVEILLANCE ZONES NUCLEAR POWER PLANTSS

2021 ◽  
Vol 26 ◽  
pp. 219-234
Author(s):  
V. Prylypko ◽  
◽  
Yu. Ozerova ◽  
I. Bondarenko ◽  
M. Morozova ◽  
...  
Keyword(s):  
Family Life ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Sampling Error ◽  
Close Correlation ◽  
The State ◽  
The Core ◽  
Terminal Values ◽  
Happy Family

Objective: to determine the place of health in the system of values of the population of the surveillance zone (SZ) of nuclear power plants (NPPs) and its importance in the perception of emergency risks (ER). Materials and methods. To determine the place of health in the value system, a survey of the able-bodied population of satellite cities of Rivne (RNPP) and South Ukrainian (SUNPP) nuclear power plants was conducted using nonrepetitive sampling, where the sampling error does not exceed 7,0 %. The motivational and behavioral component that determined health in the individual hierarchy of values of the subject according to the questionnaire Berezovskaya R. A. was studied. Statistical and mathematical methods were used in the research process. Results. The array of respondents was conditionally divided into 4 groups according to their attitude to human health. And the group where a person’s life position is focused exclusively on health is the most common – 77,0 %. Group IV, which wants to live without limiting itself, is 8,1 %. The component integrity of values-goals and valuesmeans among the urban population of the SZ of both nuclear power plants is the same: the main goal in life is health, happy family life, and as a means – perseverance, diligence and health. Goal values in groups I and IV have some differences: in the first group of respondents the main goal in life is health, and in the fourth, where a person’s life guidelines exclude any restrictions – a happy family life. Values for these populations have some differences, but in both groups health appears to be the main means to an end. There is a close correlation between the core of terminal values and the average indicators of the state of concern about the risk of emergencies. Conclusions. Identified hierarchy of values: a group of stable dominant values; average status values; group of least significant values. The values of the highest status among the values-goals are – health, happy family life and interesting work. Most respondents plan to achieve them through values such as «health», «perseverance and hard work». There is a close correlation between the core of terminal values and the average indicators of the state of concern about the risk of emergencies. Key words: health, values, population, NPP surveillance zone, perception of emergency risks.

Sign in / Sign up

Export Citation Format

Share Document