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 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.

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.

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.

Structural Integrity for the RPV of French NPP during the LRA

2016 ◽  
Vol 853 ◽  
pp. 453-457
Author(s):  
Ming Ya Chen ◽  
Wei Wei Yu ◽  
Jin Hua Shi ◽  
Rong Shan Wang ◽  
Lv Feng ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Structural Integrity ◽  
The Other ◽  
Light Water Reactor ◽  
Original Design ◽  
Light Water ◽  
Pressurized Thermal Shock ◽  
The Usa ◽  
Reactor Pressure

Most of the French Nuclear Power Plants (NPPs) are currently embarking upon efforts to renew their operating license, while the pressurized thermal shock (PTS) events and environmentally assisted fatigue (EAF) pose potentially significant challenges to the structural integrity of the reactor pressure vessel (RPV) which has the potential to be NPP life-limiting conditions. In the assessment of the PTS events, the deterministic fracture mechanics (DFM) is still used as the basic mechanics in most countries except for the USA. While the maximum nil-ductility-transition temperature (RTNDT) is about 80°C for 54 French RPVs after 40 years operation, the maximum allowable RTNDT is only about 70 oC and 80 oC for the typical PTS events in the IAEA and NEA reports, respectively. On the other hand, the effects of light water reactor (LWR) environmental (other than moderate environment in the code) were not considered in the original design, while the effects of LWR environmental are needed to be considered in the LRA according to the USA regulations. In this paper, the challenges of the PTS and EAF are discussed, and some suggestions are also given for the LRA

Decommissioning of light-water reactor nuclear power plants

1978 ◽  
Vol 45 (1) ◽  
pp. 1-51 ◽  
Author(s):  
R. Bardtenschlager ◽  
D. Bottger ◽  
A. Gasch ◽  
N. Majohr
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Light Water Reactor ◽  
Light Water ◽  
Water Reactor

Lessons Learned From the Operation of Large-Scale Reprocessing-Recycling Facilities in France: What Is in It for China?

2017 ◽  
Author(s):  
Jean-Pierre Gros
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Large Scale ◽  
Central Government ◽  
Fuel Cycle ◽  
Public Information ◽  
Lessons Learned ◽  
Mitigation Measures ◽  
Spent Fuel ◽  
The Impact

AREVA has been running since decades nuclear reprocessing and recycling installations in France. Several industrial facilities have been built and used to this aim across the time. Following those decades and with the more and more precise monitoring of the impact of those installations, precise data and lessons-learned have been collected that can be used for the stakeholders of potential new facilities. China has expressed strong interest in building such facilities. As a matter of fact, the issue of accumulation of spent fuel is becoming serious in China and jeopardizes the operation of several nuclear power plants, through the running out of space of storage pools. Tomorrow, with the extremely high pace of nuclear development of China, accumulation of spent fuel will be unbearable. Building reprocessing and recycling installations takes time. A decision has to be taken so as to enable the responsible development of nuclear in China. Without a solution for the back end of its nuclear fuel cycle, the development of nuclear energy will face a wall. This is what the Chinese central government, through the action of its industrial CNNC, has well understood. Several years of negotiations have been held with AREVA. Everybody in the sector seems now convinced. However, now that the negotiation is coming to an end, an effort should be done towards all the stakeholders, sharing actual information from France’s reference facilities on: safety, security, mitigation measures for health protection (of the workers, of the public), mitigation measures for the protection of the environment. Most of this information is public, as France has since years promulgated a law on Nuclear transparency. China is also in need for more transparency, yet lacks means to access this public information, often in French language, so let’s open our books!

scholarly journals Investigation on Aerosol Pool Scrubbing Model During Severe Accidents

2021 ◽  
Vol 9 ◽  
Author(s):  
L. W. He ◽  
Y. X. Li ◽  
Y. Zhou ◽  
S. Chen ◽  
L. L. Tong ◽  
...  
Keyword(s):  
Bubble Size ◽  
Nuclear Power ◽  
Predictive Ability ◽  
Decontamination Factor ◽  
Light Water Reactor ◽  
Severe Accident ◽  
Spent Fuel ◽  
Analysis Model ◽  
Light Water ◽  
Water Reactor

During a nuclear power plant severe accident, discharging gas mixture into the spent-fuel pool is an alternative containment depressurization measurement through which radioactive aerosols can be scrubbed. However, it is necessary to develop a code for analyzing the decontamination factor of aerosol pool scrubbing. This article has established the analysis model considering key aerosol pool scrubbing mechanisms and introduced the Akita bubble size relationship. In addition, a code for evaluating the decontamination factor of aerosol pool scrubbing was established. The Advanced Containment Experiment and Light Water Reactor Advanced Containment Experiment were simulated with the code considering different bubble sizes of the Akita model and MELCOR default value to verify the suitability of the Akita bubble size model for simulating aerosol pool scrubbing. Furthermore, the simulation results were compared with the results analyzed by MELCOR code and COCOSYS code from literature, and equivalent predictive ability was observed. In addition, a sensitivity analysis on bubble size was conducted, and the contribution of different behaviors and mechanisms has been discussed. Finally, the bubble breakup equation was revised and verified with the conditions of the multi-hole bubbler in the Advanced Containment Experiment and Light Water Reactor Advanced Containment Experiment.

scholarly journals Nuclear Fuel Transmutation

2021 ◽  
Author(s):  
Akbar Abbasi
Keyword(s):  
Nuclear Fuel ◽  
Nuclear Power ◽  
Power Plants ◽  
Fuel Cycle ◽  
Electric Energy ◽  
Computer Code ◽  
Natural Uranium ◽  
Spent Fuel ◽  
Used Nuclear Fuel ◽  
Front End

Nuclear power plants to generates electric energy used nuclear fuel such as Uranium Oxide (UOX). A typical VVER−1000 reactor uses about 20–25 tons of spent fuel per year. The fuel transmutation of UOX fuel was evaluated by VISTA computer code. In this estimation the front end and back end components of fuel cycle was calculated. The front end of the cycle parameter are FF requirements, enrichment value requirements, depleted uranium amount, conversion requirements and natural uranium requirements. The back-end component is Spent Fuel (SF), Actinide Inventory (AI) and Fission Product (FP) radioisotopes.

Regulatory Perspectives on ASME OM Code, Mandatory Appendix IV, Revision 0, “Inservice Testing of Air-Operated Valves in Light-Water Reactor Nuclear Power Plants”

2014 ◽  
Author(s):  
Jason Carneal
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Diagnostic Testing ◽  
Light Water Reactor ◽  
Federal Regulations ◽  
Light Water ◽  
Water Reactor ◽  
Asme Code ◽  
American Society

The American Society of Mechanical Engineers (ASME) Code for Operation and Maintenance of Nuclear Power Plants (OM Code) establishes the requirements for preservice and inservice testing and examination of certain components to assess their operational readiness in light-water reactor nuclear power plants. The Code of Federal Regulations (CFR) endorses and mandates the use of the ASME OM Code for testing air-operated valves in 10 CFR 50.55a(b)(3)(ii) and 10 CFR 50.55a(f)(4), respectively. ASME has recently approved Mandatory Appendix IV, Revision 0. NRC currently anticipates that Mandatory Appendix IV will first appear in the 2014 Edition of the ASME OM Code. Publication of the 2014 Edition of the ASME OM Code begins the NRC rulemaking process to modify 10 CFR 50.55a to incorporate the 2014 Edition of the ASME OM Code by reference. NRC staff has actively participated in the development of Mandatory Appendix IV, Revision 0, through participation in the ASME OM Code Subgroup on Air-Operated Valves (SG-AOV). The purpose of this paper is to provide NRC staff perspectives on the contents and implementation of Mandatory Appendix IV, Revision 0. This paper specifically discusses Mandatory Appendix IV, Sections IV-3100, “Design Review,” IV-3300, “Preservice Test,” IV-3400, “Inservice Test,” IV-3600, “Grouping of AOVs for Inservice Diagnostic Testing,” and IV-3800, “Risk Informed AOV Inservice Testing.” These topics were selected based on input received during NRC staff participation in the SG-AOV and other industry meetings. The goal of this paper is to provide NRC staff perspectives on the topics of most interest to NRC staff and members of the SG-AOV. Paper published with permission.

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