scholarly journals Operating Experience of Desalination Unit Coupled to Primary Coolant System of Cirus

10.5772/21313 ◽  
2011 ◽  
Author(s):  
R.C. Sharma ◽  
Rakesh Ranj
Keyword(s):  
Operating Experience ◽  
Primary Coolant ◽  
Coolant System

Thermal Fatigue of Pwr Coolant System Loop Drain Lines Due to Outflow Activities

2021 ◽  
Author(s):  
Yue Zou ◽  
Brian Derreberry
Keyword(s):  
Thermal Fatigue ◽  
Operating Experience ◽  
Fatigue Cracking ◽  
Screening Methods ◽  
Primary Coolant ◽  
Pressurized Water ◽  
Thermal Transients ◽  
Piping Systems ◽  
Class 1 ◽  
Coolant System

Abstract Thermal cycling induced fatigue is widely recognized as one of the major contributors to the damage of nuclear plant piping systems, especially at locations where turbulent mixing of flows with different temperature occurs. Thermal fatigue caused by swirl penetration interaction with normally stagnant water layers has been identified as a mechanism that can lead to cracking in dead-ended branch lines attached to pressurized water reactor (PWR) primary coolant system. EPRI has developed screening methods, derived from extensive testing and analysis, to determine which lines are potentially affected as well as evaluation methods to perform evaluations of this thermal fatigue mechanism for the U.S. PWR plants. However, recent industry operating experience (OE) indicate that the model used to predict thermal fatigue due to swirl penetration is not fully understood. In addition, cumulative effects from other thermal transients, such as outflow activities, may also contribute to the failure of the RCS branch lines. In this paper, we report direct OE from one of our PWR units where thermal fatigue cracking is observed at the RCS loop drain line close to the welded region of the elbow. A conservative analytical approach that takes into account the influence of thermal stratification, in accordance with ASME Class 1 piping stress method, is also proposed to evaluate the severity of fatigue damage to the RCS drain line, as a result of transients from outflow activities. Finally, recommendations are made for future operation and inspection based on results of the evaluation.

scholarly journals Operating experience of desalination unit coupled to primary coolant system of CIRUS

2010 ◽  
Vol 4 (2) ◽  
pp. 134 ◽  
Author(s):  
Rakesh Ranjan ◽  
Surinder Kumar ◽  
D. Mathur ◽  
N. Ramesh ◽  
R.C. Sharma
Keyword(s):  
Operating Experience ◽  
Primary Coolant ◽  
Coolant System

Stand-Alone Containment Analysis of the PHÉBUS FPT-1 Test With the ASTEC and the MELCOR Codes

2016 ◽  
Author(s):  
Bruno Gonfiotti ◽  
Sandro Paci
Keyword(s):  
State Of The Art ◽  
External Environment ◽  
Fission Products ◽  
Severe Accident ◽  
Sensitivity Analyses ◽  
Complex Task ◽  
Primary Coolant ◽  
Control Volumes ◽  
Coolant System

The estimation of Fission Products (FPs) release from the containment system of a nuclear plant to the external environment during a Severe Accident (SA) is a quite complex task. In the last 30–40 years several efforts were made to understand and to investigate the different phenomena occurring in such a kind of accidents in the primary coolant system and in the containment. These researches moved along two tracks: understanding of involved phenomenologies through the execution of different experiments, and creation of numerical codes capable to simulate such phenomena. These codes are continuously developed to reflect the actual SA state-of-the-art, but it is necessary to continuously check that modifications and improvements are able to increase the quality of the obtained results. For this purpose, a continuous verification and validation work should be carried out. Therefore, the aim of the present work is to re-analyze the Phébus FPT-1 test employing the ASTEC (F) and MELCOR (USA) codes. The analysis focuses on the stand-alone containment aspects of the test, and three different modellisations of the containment vessel have been developed showing that at least 15/20 Control Volumes (CVs) are necessary for the spatial schematization to correctly predict thermal-hydraulics and the aerosol behavior. Furthermore, the paper summarizes the main thermal-hydraulic results, and presents different sensitivity analyses carried out on the aerosols and FPs behavior.

Modeling of Porous Crud Layer Based on Effects of Diffusion and Water Radiolysis on Anomalous Porous Crud Deposition on Fuel Pin Surfaces in PWRs

2002 ◽  
Author(s):  
Hyun-Jong Joe ◽  
Barclay G. Jones
Keyword(s):  
Negative Impact ◽  
Nucleate Boiling ◽  
Water Radiolysis ◽  
Trace Impurities ◽  
Primary Coolant ◽  
Pressurized Water ◽  
Fuel Pin ◽  
Negative Effect ◽  
Water Reactors ◽  
Coolant System

Many studies have been undertaken to understand crud formation on the upper spans of fuel pin clad surfaces, which is called axial offset anomaly (AOA), is observed in pressurized water reactors (PWR) as a result of sub-cooled nucleate boiling. Separately, researchers have considered the effect of water radiolysis in the primary coolant of PWR. This study examines the effects of radiolysis of liquid water, which aggressively participate in general cladding corrosion and solutes within the primary coolant system, in the terms of pH, temperature, and Linear Energy Transfer (LET). It also discusses the effect of mass transfer, especially diffusion, on the concentration distribution of the radiolytic products, H2 and O2, in the porous crud layer. Finally it covers the effects of chemical reactions of boric acid (H3BO3), which has a negative impact on the mechanisms of water recombination with hydrogen, lithium hydroxide (LiOH), which has a negative effect on water decomposition, dissolved hydrogen (DH), and some trace impurities.

scholarly journals Generic Aging Management Programs for License Renewal of BWR Reactor Coolant System Components

2002 ◽  
Author(s):  
V. N. Shah ◽  
Y. Y. Liu
Keyword(s):  
Operating Experience ◽  
Extended Period ◽  
Aging Effect ◽  
Lessons Learned ◽  
Monitoring And Control ◽  
Aging Effects ◽  
Management Activity ◽  
Reactor Coolant ◽  
License Renewal ◽  
Coolant System

The paper reviews the existing generic aging management programs (AMPs) for the reactor coolant system (RCS) components in boiling water reactors (BWRs), including the reactor pressure vessel and internals, the reactor recirculation system, and the connected piping. These programs have been evaluated in the U.S. Nuclear Regulatory Commission (NRC) report, Generic Aging Lessons Learned (GALL), NUREG-1801, for their use in the license renewal process to manage several aging effects, including loss of material, crack initiation and growth, loss of fracture toughness, loss of preload, wall thinning, and cumulative fatigue damage. The program evaluation includes a review of ten attributes (scope of program, preventive actions, parameters monitored/inspected, detection of aging effects, monitoring and trending, acceptance criteria, corrective actions, confirmative process, administrative control, and operating experience) for their effectiveness in managing a specific aging effect in a given component(s). The generic programs are based on the ASME Section XI inservice inspection requirements; industry guidelines for inspection and evaluation of aging effects in BWR reactor vessel, internals, and recirculation piping; monitoring and control of BWR water chemistry; and operating experience as reported in the USNRC generic communications and industry reports. The review concludes that all generic AMPs are acceptable for managing aging effects in BWR RCS components during an extended period of operation and do not need further evaluation. However, the plant-specific programs for managing aging in certain RCS components during an extended period of operation do require further evaluation. For some plant-specific AMPs, the GALL report recommends an aging management activity to verify their effectiveness. An example of such an activity is a one-time inspection of Class 1 small-bore piping to ensure that service-induced weld cracking is not occurring in the piping. Several of the AMPs, which do not require further evaluation, do need enhancements to allow for an extended period of reactor operation. The paper concludes that the GALL report systematically evaluates the generic AMPs for their effectiveness during an extended period of operation and provides a technical basis for their acceptance. The use of the GALL report should facilitate both preparation of a license renewal application and timely and uniform review by the NRC staff.

scholarly journals Fouling Of Nuclear Steam Generators: Fundamental Studies, Operating Experience and Remedial Measures Using Chemical Additives

2013 ◽  
Vol 2 (1) ◽  
pp. 61-88 ◽  
Author(s):  
C.W. Turner
Keyword(s):  
Economic Performance ◽  
Nuclear Power ◽  
Power Plants ◽  
Tube Bundle ◽  
Operating Experience ◽  
Feed Water ◽  
Chemical Additives ◽  
Steam Generators ◽  
Primary Coolant ◽  
And Performance

Fouling remains a potentially serious issue that if left unchecked can lead to degradation of the safety and performance of nuclear steam generators (SGs). It has been demonstrated that the majority of the corrosion product transported with the feed water to the SGs accumulates in the SG on the tube-bundle. By increasing the risk of tube failure and acting as a barrier to heat transfer, deposit on the tube bundle has the potential to impair the ability of the SG to perform its two safety-critical roles: provision of a barrier to the release of radioactivity from the reactor coolant and removal of heat from the primary coolant during power operation and under certain post accident scenarios. Thus, it is imperative to develop improved ways to mitigate SG fouling for the long-term safe, reliable and economic performance of nuclear power plants (NPPs). This paper provides an overview of our current understanding of the mechanisms by which deposit accumulates on the secondary side of the SG, how this accumulation affects SG performance and how accumulation of deposit can be mitigated using chemical additives to the secondary heat-transport system. The paper concludes with some key questions that remain to be addressed to further advance our knowledge of deposit accumulation and how it can be controlled to maintain safe, economic performance of nuclear SGs.

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