Materials and Stress Analysis of Fuel-Channel of a Generic 1200-MWel Pressure-Channel Reactor With Nuclear Steam Reheat

2014 ◽  
Author(s):  
Rand Abdullah ◽  
Matthew Baldock ◽  
Andrei Vincze ◽  
Khalil Sidawi ◽  
Igor Pioro
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Pressure Tube ◽  
Outlet Temperature ◽  
Pressurized Water ◽  
Fuel Channel ◽  
Suitable Material ◽  
Inconel 600 ◽  
Fuel Bundle ◽  
Ss 304

The objective of this paper is to modify the current existing CANFLEX® fuel-bundle design to examine its ability to withstand high-temperature conditions of a proposed generic reactor with nuclear steam reheat. One of this reactor’s characteristics is having Super-Heated Steam (SHS) channels in addition to Pressurized-Water (PW) channels in order to increase the thermal efficiency of the plant by about 7–12%. This increase may be attained by raising the outlet temperature of the SHS-channels coolant to about 550°C. Operating at the higher temperatures will definitely have an effect on the mechanical and neutronic properties of fuel-channel materials, specifically on fuel-sheath and pressure-tube materials. This paper compares Inconel-600 and SS-304 in order to determine the most suitable material for SHS-channel’s sheath and pressure tube. This is achieved by comparing strength of materials by performing stress- and displacement-analysis simulation using NX8.5 software (NX8.5, 2009). The analysis in this paper can also be applied to other Nuclear-Power Plants (NPPs) that require operating at higher temperatures such as Super-Critical Water-cooled Reactors (SCWRs).

Failure of Resistance Thermometer Devices Due to Flow-Induced Vibrations

2005 ◽  
Author(s):  
S. Mahmood Husaini ◽  
Riyad K. Qashu ◽  
Robert D. Blevins
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Plant ◽  
Primary System ◽  
Pressurized Water Reactor ◽  
Resistance Thermometer ◽  
Pressurized Water ◽  
Inconel 600 ◽  
Safety Function ◽  
Vortex Induced Vibrations

Resistance Thermometer Devices (RTDs) are used to monitor the temperatures in sensitive locations of process piping, such as the hot and cold legs of the primary system of pressurized water reactor nuclear power plants. The RTDs are housed in thermowells that protrude in the cold and hot leg pipes. Eight of the twelve cold leg RTDs in the San Onofre Nuclear Plant have a direct safety function in the operation of the reactor. They provide the Core Protection Calculators with two temperature measurements from each of the cold legs. During the period 1997–2004, four RTD failures occurred in the cold legs of the San Onofre Nuclear Plant. Extensive investigations showed that all of the RTD failures were caused by cracking and fracturing of the platinum wires in the sensors. There was no damage to the nozzles or thermowells of the cold leg RTDs. Also, none of the RTDs in the hot leg were damaged. These failures occurred after the Cycle 9 refueling outage, when the Reactor Coolant System thermowells were replaced due to potential cracking of the Inconel 600 material. A root cause analysis was performed to identify the reasons for the failure of the RTDs in the reactor cold leg piping. The thermowells used for both the hot and cold leg RTDs are identical. The thermowells are 11.375” (289 mm) in length. Since the thickness of the cold leg pipe is 0.75” (19.1 mm) less than the thickness of the hot leg pipe, the cold leg thermowells protrude 0.75” (19.1 mm) more into the flow stream as compared to the hot leg thermowells. The protrusion lengths of the hot and cold leg thermowells in the pipe are 2.5” (63.5 mm) and 3.25” (82.6 mm), respectively. The increased protrusion length of the thermowells in the cold leg significantly lowers their natural frequency (as compared to the hot leg thermowells), which results in vortex-induced vibrations at lower flow velocities that cause failure. Also, the increased protrusion length results in increased amplitude of vortex-induced vibration of their tip that exceeds the capability of the installed RTDs. Thus, a recommendation was made to decrease the protrusion length into the cold leg pipe from 3.25” (82.6 mm) to 2.5” (63.5 mm). Since this recommendation was implemented more than eight months ago, there have been no RTD failures in the cold legs.

The Effects of Ion Implantation on the Surface Features of Inconel 600

1985 ◽  
Vol 43 ◽  
pp. 288-289
Author(s):  
John D. Rubio
Keyword(s):  
Grain Boundary ◽  
Ion Implantation ◽  
Nuclear Power ◽  
Power Plants ◽  
Surface Region ◽  
Selective Dissolution ◽  
Boundary Region ◽  
Near Surface ◽  
Inconel 600 ◽  
Steam Generator Tubing

The degradation of steam generator tubing at nuclear power plants has become an important problem for the electric utilities generating nuclear power. The material used for the tubing, Inconel 600, has been found to be succeptible to intergranular attack (IGA). IGA is the selective dissolution of material along its grain boundaries. The author believes that the sensitivity of Inconel 600 to IGA can be minimized by homogenizing the near-surface region using ion implantation. The collisions between the implanted ions and the atoms in the grain boundary region would displace the atoms and thus effectively smear the grain boundary.To determine the validity of this hypothesis, an Inconel 600 sample was implanted with 100kV N2+ ions to a dose of 1x1016 ions/cm2 and electrolytically etched in a 5% Nital solution at 5V for 20 seconds. The etched sample was then examined using a JEOL JSM25S scanning electron microscope.

The Failure Analysis and Processing of Digital Reactor Protection System

2013 ◽  
Author(s):  
Jun Zhao ◽  
Xing Zhou ◽  
Jin Hu ◽  
Yanling Yu
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Phase 1 ◽  
Protection System ◽  
Pressurized Water Reactor ◽  
Communication Failure ◽  
Pressurized Water ◽  
Power Failure ◽  
Self Test ◽  
Reactor Protection System

The Qinshan Nuclear Power Plant phase 1 unit (QNPP-1) has a power rating of 320 MWe generated by a pressurized water reactor that was designed and constructed by China National Nuclear Corporation (CNNC). The TELEPERM XS I&C system (TXS) is to be implemented to transform analog reactor protection system (RPS) in QNPP-1. The paper mainly describes the function, structure and characteristic of RPS in QNPP-1. It focuses on the outstanding features of digital I&C, such as strong online self-test capability, the degradation of the voting logic processing, interface improvements and CPU security. There are some typical failures during the operation of reactor protection system in QNPP-1. The way to analyze and process the failures is different from analog I&C. The paper summarizes typical failures of the digital RPS in the following types: CPU failure, communication failure, power failure, Input and output (IO) failure. It discusses the cause, risk and mainly processing points of typical failure, especially CPU and communication failures of the digital RPS. It is helpful for the maintenance of the system. The paper covers measures to improve the reliability of related components which has been put forward effective in Digital reactor protection system in QNPP-1. It will be valuable in nuclear community to improve the reliability of important components of nuclear power plants.

Development of Long Rotating Blade Applied in the New Generation Nuclear Steam Turbine

2012 ◽  
Author(s):  
Kai Cheng ◽  
Zeying Peng ◽  
Gongyi Wang ◽  
Xiaoming Wu ◽  
Deqi Yu
Keyword(s):  
Steam Turbine ◽  
Nuclear Power ◽  
Power Plants ◽  
Structure Design ◽  
Good Reliability ◽  
Pressurized Water ◽  
Height Range ◽  
Water Reactor ◽  
Rotating Blade ◽  
Last Stage

In order to meet the high economic requirement of the 3rd generation Pressurized Water Reactor (PWR) or Boiling Water Reactor (BWR) applied in currently developing nuclear power plants, a series of half-speed extra-long last stage rotating blades with 26 ∼ 30 m2 nominal exhaust annular area is proposed, which covers a blade-height range from 1600 mm to 1900 mm. It is well known that developing an extra long blade is a tough job involving some special coordinated sub-process. This paper is dedicated to describe the progress of creating a long rotating blade for a large scaled steam turbine involved in the 3rd generation nuclear power project. At first the strategy of how to determine the appropriate height for the last-stage-rotating-blade for the steam turbine is provided. Then the quasi-3D flow field quick design method for the last three stages in LP casing is discussed as well as the airfoil optimization method. Furthermore a sophisticated blade structure design and analyzing system for a long blade is introduced to obtain the detail dimension of the blade focusing on the good reliability during the service period. Thus, except for CAD and experiment process, the whole pre-design phase of the extra-long turbine blade is presented which is regarded as an assurance of the operation efficiency and reliability.

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.

scholarly journals Reactor Vessel Internals Segmentation Experience using Mechanical Cutting Tools

2013 ◽  
Vol 10 (2) ◽  
pp. 6-10 ◽  
Author(s):  
Petr Pospíšil
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Cutting Tools ◽  
Reactor Vessel ◽  
Boiling Water Reactors ◽  
Abrasive Waterjet ◽  
Pressurized Water ◽  
Water Reactors ◽  
Mechanical Cutting ◽  
Reactor Internals

Abstract Some commercial nuclear power plants have been permanently shut down to date and decommissioned using dismantling methods. Other operating plants have decided to undergo an upgrade process that includes replacement of reactor internals. In both cases, there is a need to perform a segmentation of the reactor vessel internals with proven methods for long term waste disposal. Westinghouse has developed several concepts to dismantle reactor internals based on safe and reliable techniques, including plasma arc cutting (PAC), abrasive waterjet cutting (AWJC), metal disintegration machining (MDM), or mechanical cutting. Mechanical cutting has been used by Westinghouse since 1999 for both Pressurized Water Reactors (PWR’s) and Boiling Water Reactors (BWR’s) and its process has been continuously improved over the years. The complexity of the work requires well designed and reliable tools. Different band saws, disc saws, tube cutters and shearing tools have been developed to cut the reactor internals. All of those equipments are hydraulically driven which is very suitable for submerged applications. Westinghouse experience in mechanical cutting has demonstrated that it is an excellent technique for segmentation of internals. In summary, the purpose of this paper will be to provide an overview of the Westinghouse mechanical segmentation process, based on actual experience from the work that has been completed to date.

scholarly journals Plant Performance History of an Innovative Gate Valve in Critical Service Applications

2017 ◽  
Author(s):  
M. S. Kalsi ◽  
Patricio Alvarez ◽  
Thomas White ◽  
Micheal Green
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Gate Valve ◽  
Plant Performance ◽  
Flow Loop ◽  
Performance History ◽  
Pressurized Water ◽  
Maintenance Costs ◽  
History Of ◽  
Water Reactors

A previous paper [1] describes the key features of an innovative gate valve design that was developed to overcome seat leakage problems, high maintenance costs as well as issues identified in the Nuclear Regulatory Commission (NRC) Generic Letters 89-10, 95-07 and 96-05 with conventional gate valves [2,3,4]. The earlier paper was published within a year after the new design valves were installed at the Pilgrim Nuclear Plant — the plant that took the initiative to form a teaming arrangement as described in [1] which facilitated this innovative development. The current paper documents the successful performance history of 22 years at the Pilgrim plant, as well as performance history at several other nuclear power plants where these valves have been installed for many years in containment isolation service that requires operation under pipe rupture conditions and require tight shut-off in both Pressurized Water Reactors (PWRs) and Boiling Water Reactors (BWRs). The performance history of the new valve has shown to provide significant performance advantage by eliminating the chronic leakage problems and high maintenance costs in these critical service applications. This paper includes a summary of the design, analysis and separate effects testing described in detail in the earlier paper. Flow loop testing was performed on these valves under normal plant operation, various thermal binding and pressure locking scenarios, and accident/pipe rupture conditions. The valve was designed, analyzed and tested to satisfy the requirements of ANSI B16.41 [9]; it also satisfies the requirements of ASME QME 1-2012 [10]. The results of the long-term performance history including any degradation observed and its root cause are summarized in the paper. Paper published with permission.

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