scholarly journals RECENT VALIDATION STUDY OF THE TECHNIQUE FOR EXPRESS EVALUATION OF BURNUP IN LEAKING FUEL ASSEMBLIES OF WWER POWER UNITS

2021 ◽  
Vol 247 ◽  
pp. 10013
Author(s):  
O.V. Vilkhivskaya ◽  
I.A. Evdokimov ◽  
V.V. Likhanskii ◽  
E.Yu. Afanasieva
Keyword(s):  
Validation Study ◽  
Nuclear Power ◽  
Power Plants ◽  
Fuel Cycle ◽  
Conventional Technique ◽  
Fuel Burnup ◽  
Fuel Cycles ◽  
Activity Measurements ◽  
Fuel Assemblies ◽  
Activity Ratios

The present work continues the series of papers on the revision of the conventional technique for evaluation of leaking fuel burnup during reactor operation at nuclear power plants (NPPs). The focus was made on reduction of uncertainties in evaluation of leaking fuel burnup in modern fuel cycles at WWER-1000 power units. A set of models was proposed for express calculation of the build-up of caesium isotopes in fuel and to relate 134Cs/137Cs activity ratio with fuel burnup for each rod in the core. These models are based on routine neutronic calculations of pin-by-pin linear heat generation rates which are performed at NPPs for each particular fuel cycle with particular core loading pattern (however, these calculations do not provide data on caesium inventory in fuel). Previously, the proposed models have been validated against several practical cases. This latest validation study relied on the analysis of the most recent fuel cycles at two NPPs that reported spike-events and identified the leaking fuel assemblies (LFAs) after the reactor shutdown. The calculated 134Cs/137Cs activity ratios in the fuel of the LFAs were compared to the NPPs data on the activity measurements, and to the post-irradiation examination (PIE) data provided for one FA. A reasonable agreement between the model predictions and the experimental data on 134Cs/137Cs activity ratios in the fuel as a function of its burnup is shown for the advanced FA designs in modern fuel cycles.

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.

Development Trends in Nuclear Technology and Related Safety Aspects

2002 ◽  
Author(s):  
B. Kuczera ◽  
P. E. Juhn ◽  
K. Fukuda
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Fuel Cycle ◽  
Nuclear Reactors ◽  
Nuclear Technology ◽  
User Requirements ◽  
Safety Standards ◽  
Safety Requirements ◽  
Fuel Cycles ◽  
High Level

The IAEA Safety Standards Series include, in a hierarchical manner, the categories of Safety Fundamentals, Safety Requirements and Safety Guides, which define the elements necessary to ensure the safety of nuclear installations. In the same way as nuclear technology and scientific knowledge advance continuously, also safety requirements may change with these advances. Therefore, in the framework of the International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO) one important aspect among others refers to user requirements on the safety of innovative nuclear installations, which may come into operation within the next fifty years. In this respect, the major objectives of the INPRO subtask “User Requirements and Nuclear Energy Development Criteria in the Area of Safety” have been: a. to overview existing national and international requirements in the safety area, b. to define high level user requirements in the area of safety of innovative nuclear technologies, c. to compile and to analyze existing innovative reactor and fuel cycle technology enhancement concepts and approaches intended to achieve a high degree of safety, and d. to identify the general areas of safety R&D needs for the establishment of these technologies. During the discussions it became evident that the application of the defence in depth strategy will continue to be the overriding approach for achieving the general safety objective in nuclear power plants and fuel cycle facilities, where the emphasis will be shifted from mitigation of accident consequences more towards prevention of accidents. In this context, four high level user requirements have been formulated for the safety of innovative nuclear reactors and fuel cycles. On this basis safety strategies for innovative reactor designs are highlighted in each of the five levels of defence in depth and specific requirements are discussed for the individual components of the fuel cycle.

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.

scholarly journals Study of Thorium Fuel Cycles for Light Water Reactor VBER-150

2013 ◽  
Vol 2013 ◽  
pp. 1-9
Author(s):  
Daniel Evelio Milian Lorenzo ◽  
Daniel Milian Pérez ◽  
Lorena Pilar Rodríguez García ◽  
Jesús Salomón Llanes ◽  
Carlos Alberto Brayner de Oliveira Lira ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Fuel Cycle ◽  
Light Water Reactor ◽  
Spent Fuel ◽  
Light Water ◽  
Fuel Cycles ◽  
Standard Design ◽  
Pu Isotopes ◽  
Open Cycle

The main objective of this paper is to examine the use of thorium-based fuel cycle for the transportable reactors or transportable nuclear power plants (TNPP) VBER-150 concept, in particular the neutronic behavior. The thorium-based fuel cycles included Th232+Pu239, Th232+U233, and Th232+U and the standard design fuel UOX. Parameters related to the neutronic behavior such as burnup, nuclear fuel breeding, MA stockpile, and Pu isotopes production (among others) were used to compare the fuel cycles. The Pu transmutation rate and accumulation of Pu with MA in the spent fuel were compared mutually and with an UOX open cycle. The Th232+U233 fuel cycle proved to be the best cycle for minimizing the production of Pu and MA. The neutronic calculations have been performed with the well-known MCNPX computational code, which was verified for this type of fuel performing calculation of the IAEA benchmark announced by IAEA-TECDOC-1349.

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.

Transmutation and the Global Nuclear Energy Partnership

2006 ◽  
Vol 985 ◽  
Author(s):  
James Bresee
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Energy ◽  
Energy Use ◽  
Fuel Cycle ◽  
Spent Fuel ◽  
Closed Fuel Cycle ◽  
State Of The Union ◽  
Foreign Countries ◽  
The U.S

AbstractIn the January 2006 State of the Union address, President Bush announced a new Advanced Energy Initiative, a significant part of which is the Global Nuclear Energy Initiative. Its details were described on February 6, 2006 by the U.S. Secretary of Energy. In summary, it has three parts: (1) a program to expand nuclear energy use domestically and in foreign countries to support economic growth while reducing the release of greenhouse gases such as carbon dioxide. (2) an expansion of the U.S. nuclear infrastructure that will lead to the recycling of spent fuel and a closed fuel cycle and, through transmutation, a reduction in the quantity and radiotoxicity of nuclear waste and its proliferation concerns, and (3) a partnership with other fuel cycle nations to support nuclear power in additional nations by providing small nuclear power plants and leased fuel with the provision that the resulting spent fuel would be returned by the lessee to the lessor. The final part would have the effect of stabilizing the number of fuel cycle countries with attendant non-proliferation value. Details will be given later in the paper.

Possibility of Different Fuel Cycles Usage in GT-MHR

2002 ◽  
Author(s):  
N. Kodochigov ◽  
Yu. Sukharev ◽  
E. Marova ◽  
N. Ponomarev-Stepnoy ◽  
E. Glushkov ◽  
...  
Keyword(s):  
Fuel Cycle ◽  
Block Design ◽  
Average Power ◽  
Reactor Core ◽  
Fuel Burnup ◽  
Fuel Load ◽  
Fuel Cycles ◽  
Partial Cycle ◽  
Control Rods ◽  
Neutronic Characteristics

The GT-MHR reactor core is characterized by flexibility of neutronic characteristics at the given average power density and fixed geometrical dimensions of reactor core. Such flexibility makes it possible to start the reactor operation with one fuel cycle, and then to turn to another type of core fuel load without changes of main reactor elements: fuel block design, core and reflector size, control rod number etc. Preliminary analysis reindicates the commercial viability of the GT-MHR, part of which is due to the ability to accommodate different fuel types and cycles. This paper presents the results of studies of the neutronic characteristics of reactor cores using different fuel (low- and high-enriched uranium, MOX fuel). Comparison of different fuel cycles is carried out for a three-batch refueling option with respect to following characteristics: discharged fuel burnup, reactivity change during one partial cycle of fuel burnup, consumption of fissile isotopes per unit of produced energy, power distribution, reactivity effects, control rods worth. It is shown, that the considered options of fuel loads provide the three-year fuel campaign (with accounting of capacity factor ∼ 0,8) without change of core design, number and design of control rods at transition from the one fuel type to another.

Scenario Analysis on the Benefits of Multi-National Cooperation for the Development of a Common Nuclear Energy System Based on PWR and LFR Fleets

2014 ◽  
Author(s):  
Marco Ciotti ◽  
Jorge L. Manzano ◽  
Vladimir Kuznetsov ◽  
Galina Fesenko ◽  
Luisa Ferroni ◽  
...  
Keyword(s):  
Public Opinion ◽  
Nuclear Fuel ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Energy ◽  
Fuel Cycle ◽  
Proper Time ◽  
Nuclear Fuel Cycle ◽  
Levelized Cost Of Electricity ◽  
Base Load

Financial aspects, environmental concerns and non-favorable public opinion are strongly conditioning the deployment of new Nuclear Energy Systems across Europe. Nevertheless, new possibilities are emerging to render competitive electricity from Nuclear Power Plants (NPPs) owing to two factors: the first one, which is the fast growth of High Voltage lines interconnecting the European countries’ national electrical grids, this process being triggered by huge increase of the installed intermittent renewable electricity sources (Wind and PV); and the second one, determined by the carbon-free constraints imposed on the base load electricity generation. The countries that due to public opinion pressure can’t build new NPPs on their territory may find it profitable to produce base load nuclear electricity abroad, even at long distances, in order to comply with the European dispositions on the limitation of the CO2 emissions. In this study the benefits from operating at multinational level with the deployment of a fleet of PWRs and subsequently, at a proper time, the one of Lead Fast Reactors (LFRs) are analyzed. The analysis performed involves Italy (a country with a current moratorium on nuclear power on spite that its biggest utility operates NPPs abroad), and the countries from South East and Central East Europe potentially looking for introduction or expansion of their nuclear power programmes. According to the predicted evolution of their Gross Domestic Product (GDP) a forecast of the electricity consumption evolution for the present century is derived with the assumption that a certain fraction of it will be covered by nuclear electricity. In this context, evaluated are material balances for the front and the back end of nuclear fuel cycle associated with the installed nuclear capacity. A key element of the analysis is the particular type of LFR assumed in the scenario, characterized by having a fuel cycle where only fission products and the reprocessing losses are sent for disposition and natural or depleted uranium is added to fuel in each reprocessing cycle. Such LFR could be referred to as “adiabatic reactor”. Owing to introduction of such reactors a substantive reduction in uranium consumption and final disposal requirements can be achieved. Finally, the impacts of the LFR and the economy of scale in nuclear fuel cycle on the Levelized Cost of Electricity (LCOE) are being evaluated, for scaling up from a national to a multinational dimension, illustrating the benefits potentially achievable through cooperation among countries.

An overview of research activities on cementitious materials for radioactive waste management

2012 ◽  
Vol 1475 ◽  
Author(s):  
Zoran Drace ◽  
Irena Mele ◽  
Michael I. Ojovan ◽  
R. O. Abdel Rahman
Keyword(s):  
Radioactive Waste ◽  
Waste Management ◽  
Nuclear Power ◽  
Power Plants ◽  
Management Practices ◽  
Fuel Cycle ◽  
Large Body ◽  
Cementitious Materials ◽  
Radioactive Waste Management ◽  
Research Activities

ABSTRACTAn overview is given on research activities on cementitious materials for radioactive waste management systems based on the IAEA Coordinated Research Project (CRP) held in 2007-2010. It has been joined by 26 research organizations from 22 countries which shared their research and practical activities on use of cementitious materials for various barrier purposes. The CRP has initially formulated the research topics considered within four specific streams: A) Conventional cementitious systems; B) Novel cementitious materials and technologies; C) Testing and waste acceptance criteria; and D) Modelling long term behaviour.The CRP has analysed both barrier functions and interactions envisaged between various components with focus on predisposal stage of waste management. Cementation processes have achieved a high degree of acceptance and many processes are now regarded as technically mature. A large body of information is currently available on proven waste conditioning technologies although novel approaches are continuing to be devised.Most of the existing technologies have been developed for conditioning of large amounts of operational radioactive waste from nuclear power plants and other nuclear fuel cycle facilities. However new waste streams including those resulting from legacy and decommissioning activities required improved material performance and technologies.The most important outcome of CRP was the exchange of information and research co-operation between different institutions and has contributed towards general enhancement of safety by improving waste management practices and their efficiency. The paper presents the most important results and trends revealed by CRP participants. The research contributions of participating organizations will be published as country contributions in a forthcoming IAEA technical publication.

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