The Investigation of Classification Management for SSCS Susceptible to Ageing in PWR

2010 ◽  
Author(s):  
Fei Xue ◽  
Peng Liu ◽  
Xin-ming Meng ◽  
Wen-xin Ti ◽  
Lei Lin ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Evaluation Method ◽  
Residual Life ◽  
Classification Method ◽  
Energy Agency ◽  
Pressurized Water ◽  
Management Method ◽  
Classification Management

A set of screening and classification method is proposed in this paper to manage the Systems, Structures and Components (SSCs) of the Pressurized Water Reactor (PWR) for the aging management. The method proposed and carried out in China Nuclear Power Plants (NPP) is based on the systematic aging evaluation method and the management method for the license renewal from International Atomic Energy Agency (IAEA) and U.S. Nuclear Regulatory Commission (NRC), whose successful management methods are discussed in this paper. Considering the importance of the classification management of the SSCs, the classification method is investigated from the aspects of the safety, reliability and replacement of the SSCs with the Nuclear Power Plant operation experience feed-back. With the classification management method, the SSCs are managed economically and effectively for the Nuclear Power Plants in China. The method also makes great foundation for the aging management and residual life evaluation of the PWR in China.

scholarly journals Calculation of Potential Radiation Doses Associated with Predisposal Management of Dismantled Steam Generators from Nuclear Power Plants

2020 ◽  
Vol 12 (12) ◽  
pp. 5149
Author(s):  
Ga Hyun Chun ◽  
Jin-ho Park ◽  
Jae Hak Cheong
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Radiation Doses ◽  
Steam Generators ◽  
Pressurized Water ◽  
Water Reactor ◽  
General Pathway ◽  
Heavy Water Reactor ◽  
Past Experiences

Although the generation of large components from nuclear power plants is expected to gradually increase in the future, comprehensive studies on the radiological risks of the predisposal management of large components have been rarely reported in open literature. With a view to generalizing the assessment framework for the radiological risks of the processing and transport of a representative large component—a steam generator—12 scenarios were modeled in this study based on past experiences and practices. In addition, the general pathway dose factors normalized to the unit activity concentration of radionuclides for processing and transportation were derived. Using the general pathway dose factors, as derived using the approach established in this study, a specific assessment was conducted for steam generators from a pressurized water reactor (PWR) or a pressurized heavy water reactor (PHWR) in Korea. In order to demonstrate the applicability of the developed approach, radiation doses reported from actual experiences and studies are compared to the calculated values in this study. The applicability of special arrangement transportation of steam generators assumed in this study is evaluated in accordance with international guidance. The generalized approach to assessing the radiation doses can be used to support optimizing the predisposal management of large components in terms of radiological risk.

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.

INTERCOMPARISON ON INTERNAL DOSE ASSESSMENT FOR NUCLEAR POWER PLANTS IN KOREA

2019 ◽  
Vol 186 (4) ◽  
pp. 524-529
Author(s):  
Si Young Kim
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Evaluation Method ◽  
Dose Assessment ◽  
Assessment Procedure ◽  
Internal Dose ◽  
Nuclear Facilities ◽  
International Intercomparison ◽  
Selection Of

Abstract The intercomparison test is a quality assurance activity performed for internal dose assessment. In Korea, the intercomparison test on internal dose assessment was carried out for nuclear facilities in May 2018. The test involved four nuclear facilities in Korea, and seven exposure scenarios were applied. These scenarios cover the intake of 131I, a uranium mixture, 60Co and tritium under various conditions. This paper only reviews the participant results of three scenarios pertinent to the operation of nuclear power plants and adopts the statistical evaluation method, used in international intercomparison tests, to determine the significance values of the results. Although no outliers were established in the test, improvements in the internal dose assessment procedure were derived. These included the selection of intake time, selection of lung absorption type according to the chemical form and consideration of the contribution of previous intake.

AFCEN RCC-F: A New Standard for the Fire Protection Design of New Built Light Water Nuclear Power Plants

2018 ◽  
Author(s):  
Bernard Gautier ◽  
Mickael Cesbron ◽  
Richard Tulinski
Keyword(s):  
Fire Protection ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Fire Hazard ◽  
Pressurized Water ◽  
Light Water ◽  
Protection Scheme ◽  
Safety Case ◽  
Under Construction

Fire hazard is an important issue for the safety of nuclear power plants: the main internal hazard in terms of frequency, and probably one the most significant with regards to the design costs. AFCEN is publishing in 2018 a new code for fire protection of new built PWR nuclear plants, so-called RCC-F. This code is an evolution of the former ETC-F code which has been applied to different EPR plants under construction (Flamanville 3 (FA3, France), Hinkley Point C (HPC, United Kingdom), Taïshan (TSN, China)). The RCC-F code presents significant enhancement and evolutions resulting from eight years of work by the AFCEN dedicated sub-committee, involving a panel of contributors from the nuclear field. It is now opened to any type of PWR (Pressurized Water Reactor) type of nuclear power plants and not any longer limited to EPR (European Pressurized Reactor) plants. It can potentially be adapted to other light water concepts. Its objective is to help engineers design the fire prevention and protection scheme, systems and equipment with regards to the safety case and the defense in depth taking into account the French and European experience in the field. It deals also with the national regulations, with two appendices dedicated to French and British regulations respectively. The presentation gives an overview of the code specifications and focuses on the significant improvements.

Background on Low Temperature Overpressure Protection System Setpoints for Pressure-Temperature Curves

2020 ◽  
Author(s):  
Jeffrey C. Poehler ◽  
Gary L. Stevens ◽  
Anees A. Udyawar ◽  
Amy Freed
Keyword(s):  
Low Temperature ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Pressure Vessels ◽  
Pressurized Water Reactors ◽  
Documentation System ◽  
Pressurized Water ◽  
System Pressure ◽  
Water Reactors

Abstract ASME Code, Section XI, Nonmandatory Appendix G (ASME-G) provides a methodology for determining pressure and temperature (P-T) limits to prevent non-ductile failure of nuclear reactor pressure vessels (RPVs). Low-Temperature Overpressure Protection (LTOP) refers to systems in nuclear power plants that are designed to prevent inadvertent challenges to the established P-T limits due to operational events such as unexpected mass or temperature additions to the reactor coolant system (RCS). These systems were generally added to commercial nuclear power plants in the 1970s and 1980s to address regulatory concerns related to LTOP events. LTOP systems typically limit the allowable system pressure to below a certain value during plant operation below the LTOP system enabling temperature. Major overpressurization of the RCS, if combined with a critical size crack, could result in a brittle failure of the RPV. Failure of the RPV could make it impossible to provide adequate coolant to the reactor core and result in a major core damage or core melt accident. This issue affected the design and operation of all pressurized water reactors (PWRs). This paper provides a description of an investigation and technical evaluation regarding LTOP setpoints that was performed to review the basis of ASME-G, Paragraph G-2215, “Allowable Pressure,” which includes provisions to address pressure and temperature limitations in the development of P-T curves that incorporate LTOP limits. First, high-level summaries of the LTOP issue and its resolution are provided. LTOP was a significant issue for pressurized water reactors (PWRs) starting in the 1970s, and there are many reports available within the U.S. Nuclear Regulatory Commission’s (NRC’s) documentation system for this topic, including Information Notices, Generic Letters, and NUREGs. Second, a particular aspect of LTOP as related to ASME-G requirements for LTOP is discussed. Lastly, a basis is provided to update Appendix G-2215 to state that LTOP setpoints are based on isothermal (steady-state) conditions. This paper was developed as part of a larger effort to document the technical bases behind ASME-G.

Probabilistic Fracture Mechanics Benchmarking Study Involving the xLPR and PASCAL-SP Codes: Analysis by PASCAL-SP

2020 ◽  
Author(s):  
Akihiro Mano ◽  
Jinya Katsuyama ◽  
Yinsheng Li
Keyword(s):  
Fracture Mechanics ◽  
Stress Corrosion ◽  
Stress Corrosion Cracking ◽  
Nuclear Power ◽  
Power Plants ◽  
Corrosion Cracking ◽  
Energy Agency ◽  
Pressurized Water ◽  
Water Reactor ◽  
Probabilistic Fracture Mechanics

Abstract A probabilistic fracture mechanics (PFM) analysis code, PASCAL-SP, has been developed by Japan Atomic Energy Agency (JAEA) to evaluate the failure probability of piping within nuclear power plants considering aged-related degradations such as stress corrosion cracking and fatigue for both pressurized water reactor and boiling water reactor environments. To strengthen the applicability of PASCAL-SP, a benchmarking study is being performed with a PFM analysis code, xLPR, which has been developed by U.S.NRC in collaboration with EPRI. In this benchmarking study, deterministic and probabilistic analyses are undertaken on primary water stress corrosion cracking using the common analysis conditions. A deterministic analysis on the weld residual stress distributions is also considered. These analyses are carried out by U.S.NRC and JAEA independently using their own codes. Currently, the deterministic analyses by both xLPR and PASCAL-SP codes have been finished and probabilistic analyses are underway. This paper presents the details of conditions and comparisons of the results between the two aforementioned codes for the deterministic analyses. Both codes were found to provide almost the same results including the values of stress intensity factor. The conditions and results of the probabilistic analysis obtained from PASCAL-SP are also discussed.

Belarusian Ostrovets nuclear power plant: the challenge of cross-border negotiations to balance economic development and environmental protection†

2020 ◽  
Vol 13 (2) ◽  
pp. 157-168
Author(s):  
Aslan Khuseinovich Abashidze ◽  
Vladimir Mikhailovich Filippov ◽  
Alexander Mikhailovich Solntsev
Keyword(s):  
Economic Development ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Environmental Protection ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
International Environmental Agreements ◽  
Environmental Agreements ◽  
Energy Agency

Abstract States have sovereign rights that allow them to construct nuclear power plants. Moreover, engaging with nuclear power generation makes possible the achievement of the Sustainable Development Goals (2016–30) in combatting climate change, paramount to the Paris Agreement’s initiatives. In the same vein, however, constructing and operating power plants pose strict dangers to both general safety of the public and to national security. Thus, plant operations should strictly abide by the International Atomic Energy Agency (IAEA) standards and international law. As a result, it is important to consider the potential transboundary impacts of nuclear power plants and to conduct an appropriate transboundary environmental impact assessment (EIA). The article examines the construction of the Ostrovets Nuclear Power Plant by Belarus, close to the border of the Republic of Lithuania. The question in focus, however, is as follows: what international procedure can be used to coordinate issues of potentially negative transboundary impacts? Lithuania, in order to avoid the operation of the nuclear power plant, thus sought peaceful settlement of the dispute making use of the dispute resolution mechanisms based on international environmental agreements. The authors of this study show that the treaty bodies, established on the basis of international environmental agreements, provide important assistance in this matter in coordination with the IAEA. The use of these quasi-judicial means of resolving interstate disputes proves effective in pursuing a compromise between economic development and environmental protection. In the absence of such mechanisms at a universal level, one should consider utilizing such mechanisms in other regions of the world.

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