Root Cause Analysis of Pipeline Rupture in the Condensate Extraction System of Nuclear Power Plant

2017 ◽  
Author(s):  
Chen Qiang ◽  
Han Ning ◽  
Chen Weihan ◽  
Che Yinhui
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Root Cause Analysis ◽  
Extraction System ◽  
Stable Operation ◽  
Pressurized Water ◽  
High Oxygen Content ◽  
Cause Analysis ◽  
Root Cause

Pipe rupture in the Condensate Extraction system (CEX) of nuclear power plants will lead to high oxygen content in the secondary circuit and therefore exacerbates equipment and pipeline corrosion. At the same time, there is a risk of loss of vacuum in the system, which has a direct impact on the safe and stable operation of nuclear power plants and will affect the economic benefits. In this paper, Equipment Failure Root Cause Analysis (ERCA) methodology is employed combined with metallography analysis (SEM analysis, XRD) and finite element simulation analysis, to investigate the root cause for drainage pipeline rupture in CEX System of Pressurized Water Reactor (PWR). Detailed analysis process of ERCA was introduced including RCA project establishment, data collection, failure modes analysis and so on. The most probable failure mode is pointed out through the investigation and evidence analyzed. It suggests that the improper design and the installation of limiting orifice plate should be the root cause. And corresponding corrective actions are put forward in details to prevent the recurrence.

Corrosion Cause Analysis of Electric Feedwater Pump Motor Shaft in Nuclear Power Plants and Relevant Corrective Actions

2017 ◽  
Author(s):  
Che Yinhui ◽  
Guan Jianjun ◽  
Zu Shuai ◽  
Chen Qiang
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Economic Benefits ◽  
Motor Shaft ◽  
Cause Analysis ◽  
Root Cause ◽  
Feedwater Pump ◽  
Corrective Actions

Electric feedwater pump is an important feedwater equipment of nuclear power plants, and its reliability is directly related to the safe and steady operation of nuclear power plants and also economic benefits. In fact, corrosion of electric feedwater pump motor shaft occurs repeatedly, and even bearing shell in the motor can be burned out happen sometimes. This text sets out to analyze the cause of corrosion of electric feedwater pump motor shaft, identify the root cause, and further work out pertinent corrective actions based on the structure of the feedwater pump.

Root Cause Analysis of the Unexpected Behaviour of a Flaked Material Under Irradiation and Transferability to the Doel 3/Tihange 2 Reactor Pressure Vessels

2016 ◽  
Author(s):  
Evy De Bruycker ◽  
Séverine De Vroey ◽  
Xavier Hallet ◽  
Jacqueline Stubbe ◽  
Steve Nardone
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Root Cause Analysis ◽  
Pressure Vessels ◽  
Operating Cycle ◽  
Cause Analysis ◽  
Root Cause ◽  
History Of ◽  
Manufacturing History ◽  
Reactor Pressure

During the 2012 outage at Doel 3 (D3) and Tihange 2 (T2) Nuclear Power Plants (NPP), a large number of nearly-laminar indications were detected mainly in the lower and upper core shells. The D3/T2 shells are made from solid casts that were pierced and forged. Restart authorization in 2013 was accompanied by a number of “mid-term” requirements, to be completed during the first operating cycle after the restart. One of these requirements was the mechanical testing of irradiated specimens containing hydrogen flakes. These tests showed unexpected results regarding the shift in the Reference Temperature for Nil Ductility Transition (RTNDT) of the flaked material VB395 (Steam Generator shell rejected because of flakes) after irradiation. This paper presents the root cause analysis of this unexpected behaviour and its transferability (or not) to the D3/T2 Reactor Pressure Vessels (RPVs). A mechanistic and a manufacturing based approach were used, aiming at identifying the microstructural mechanisms responsible for the atypical embrittlement of VB395 and evaluating the plausibility of these mechanisms in the D3/T2 RPVs. This work was based on expert’s opinions, literature data and test results. Both flaked and unflaked samples have been investigated in irradiated and non-irradiated condition. All hydrogen-related mechanisms were excluded as root cause of the unexpected behaviour of VB395. Two possible mechanisms at the basis of the atypical embrittlement of VB395 were identified, but are still open to discussion. These mechanisms could be linked to the specific manufacturing history of the rejected VB395 shell. Since the larger than predicted shift in transition temperature after irradiation of VB395 is not linked with the hydrogen flaking and since none of the specific manufacturing history features that are possible root causes are reported for the D3/T2 RPVs, the D3/T2 shells should not show the unexpected behaviour observed in VB395.

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.

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