Feedback From Westinghouse Experience on Segmentation of Reactor Vessel Internals

2011 ◽  
Author(s):  
Paul J. Kreitman ◽  
Joseph Boucau ◽  
Per Segerud ◽  
Stefan Fallstro¨m
Keyword(s):  
Reactor Vessel ◽  
Lessons Learned ◽  
Abrasive Waterjet ◽  
Limiting Factor ◽  
Project Schedule ◽  
Pressurized Water ◽  
Successful Removal ◽  
Water Reactors ◽  
Mechanical Cutting ◽  
Reactor Internals

With more than 25 years of experience in the development of reactor vessel internals segmentation and packaging technology, Westinghouse has accumulated significant know-how in the reactor dismantling market. Building on tooling concepts and cutting methodologies developed decades ago for the successful removal of nuclear fuel from the damaged Three Mile Island Unit 2 reactor (TMI-2), Westinghouse has continuously improved its approach to internals segmentation and packaging by incorporating lessons learned and best practices into each successive project. 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. Westinghouse has applied its technology to all types of reactors covering Pressurized Water Reactors (PWR’s), Boiling Water Reactors (BWR’s), Gas Cooled Reactors (GCR’s) and sodium reactors. The primary challenges of a segmentation and packaging project are to separate the highly activated materials from the less-activated materials and package them into appropriate containers for disposal. Since space is almost always a limiting factor it is therefore important to plan and optimize the available room in the segmentation areas. The choice of the optimum cutting technology is important for a successful project implementation and depends on some specific constraints like disposal costs, project schedule, available areas or safety. Detailed 3-D modeling is the basis for tooling design and provides invaluable support in determining the optimum strategy for component cutting and disposal in waste containers, taking account of the radiological and packaging constraints. Westinghouse has also developed a variety of special handling tools, support fixtures, service bridges, water filtration systems, video-monitoring systems and customized rigging, all of which are required for a successful reactor vessel internals segmentation and packaging project. The purpose of this paper is to provide an overview of the Westinghouse reactor internals segmentation experience by illustrating projects related to various types of reactors and providing feedback from project execution.

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.

Westinghouse PWR and BWR Reactor Vessel Segmentation Experience in Using Mechanical Cutting Process

2010 ◽  
Author(s):  
Per Segerud ◽  
Joseph Boucau ◽  
Stefan Fallstro¨m ◽  
Paul J. Kreitman
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Reactor Vessel ◽  
Boiling Water Reactors ◽  
Abrasive Waterjet ◽  
Pressurized Water ◽  
Water Reactors ◽  
Mechanical Cutting ◽  
Reactor Internals

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.

Application of Aging Management Strategies for Reactor Vessel Internals and Core Support Structures

2010 ◽  
Author(s):  
Tama´s R. Liszkai ◽  
Matthew Snyder ◽  
Steve Fyfitch ◽  
Hongqing Xu ◽  
Hasan Charkas
Keyword(s):  
Degradation Mechanism ◽  
Management Strategies ◽  
Operating Experience ◽  
Management Program ◽  
Reactor Vessel ◽  
Lessons Learned ◽  
Support Structures ◽  
Acceptance Criteria ◽  
Pressurized Water ◽  
Reactor Internals

The Materials Reliability Program (MRP) Reactor Internals Focus Group (RI-FG) developed Pressurized Water Reactor (PWR) Internals Inspection and Evaluation (I&E) Guidelines under the sponsorship of the Electric Power Research Institute (EPRI). The I&E guidelines summarized in MRP-227 [1], provide a generic basis for U.S. utilities to develop their Aging Management Program (AMP) for managing the long-term aging degradation of PWR reactor internals including the existing and extended license periods. A number of internals structural bolts in the Babcock & Wilcox (B&W) design PWRs are fabricated from high-strength alloys such as Alloy A-286 or Alloy X-750. The materials in general, and bolts in particular, are known to be susceptible to stress corrosion cracking (SCC) based on past operating experience. The Upper and Lower Core Barrel (UCB and LCB) bolts have a core support function and have been generically categorized as Primary components for inspection in the I&E Guidelines. The remaining Alloy A-286 and Alloy X-750 structural bolts are in the Expansion category. Per 10CFR54, all U.S. PWRs are required to establish a unit-specific AMP for the extended license period in accordance with the ten elements of an effective AMP outlined in the Generic Aging Lessons Learned (GALL, NUREG-1801 Rev. 01, [2]) report published by the U.S. Nuclear Regulatory Commission (NRC). The goal of this paper is to provide an overview of the work performed by AREVA NP Inc. to support the development of the MRP I&E guidelines and unit-specific AMP for UCB and LCB bolts. A review of Alloy A-286 and Alloy X-750 bolts in the B&W design PWR is provided including the degradation mechanism, operating and inspection experience, replacement, and autoclave and in-reactor test results. The latest UT inspection technique used to characterize the extent of flaws is also discussed. Acceptance criteria for evaluating degraded conditions in UCB and LCB bolts were developed in accordance with the requirements of the ASME Section III, Subsection-NG core support structures requirements. In addition to Code compliance, special limits were established to limit the change in the core support structure stiffness. The acceptance criteria enable utilities to rapidly disposition UT inspection findings during an outage within 48 hours. In order to support the objectives of an efficient AMP for the UCB and LCB bolts, three-dimensional finite element models were prepared capable of evaluating all potential failure scenarios. These models enable accurate representation of flange flexibility and redistribution of loads due to deficient bolts. Prior to an outage, hypothetical patterns of bolt failures could be evaluated to support pre-outage planning and contingency preparation. During an outage, these models are used to disposition inspection results and help operability assessment of continued operation, and re-inspection requirement to ensure continued safety and integrity of the reactor vessel internals. Based on the existing work performed, future improvement and expansion of analytical capability is outlined in the last section of this paper. In conclusion, AREVA NP Inc. has demonstrated an effective use of a multi-disciplined approach using structural analyses, operating experience, material evaluations, and non-destructive examination (NDE) to fulfill both the development and implementation of unit-specific aging management commitments as required by MRP-227 for the current and extended license periods.

Neutronics Studies on a Generic Pressure-Channel Reactor With Nuclear Steam Superheat

2014 ◽  
Author(s):  
Matthew Baldock ◽  
Wargha Peiman ◽  
Andrei Vincze ◽  
Rand Abdullah ◽  
Khalil Sidawi ◽  
...  
Keyword(s):  
Power Distribution ◽  
Power Plants ◽  
Saturation Pressure ◽  
Pressure Vessels ◽  
Limiting Factor ◽  
Outlet Temperature ◽  
Pressurized Water ◽  
Water Reactors ◽  
1960S And 1970S ◽  
Pressure Channel

In order to increase the thermal efficiency of steam-cycle power plants it is necessary to achieve steam temperatures as high as possible. Current limiting factor for Nuclear Power Plants (NPPs) in achieving higher operating temperatures and, therefore, thermal efficiencies is pressures at which they can operate. From basic thermodynamics it is known that to increase further an outlet temperature in water-cooled reactors a pressure must also be increased. Current level of pressures in Pressurized Water Reactors (PWRs) is about 15–16 MPa. Therefore, next stage should be supercritical pressures, at least 23.5–25 MPa. However, such supercritical-water reactors with pressure vessels of 45–50 cm thickness don’t exist yet. One way around larger pressure vessels as well as the limit of temperature of the coolant on the saturation pressure is to employ a Pressure Channel (PCh) design with Superheated Steam channels (SHS). PCh reactors allow for different coolants and bundle configurations in one reactor core, in this case, steam would be a secondary coolant. In the 1960s and 1970s the USA and Soviet Union tested reactors using pressure channels to super-heat steam in-core to achieve outlet temperatures greater than what is currently possible with convention reactors. Nuclear materials are carefully chosen based on their neutron interaction properties in addition to their strength and resistance to corrosion. Introducing steam channels will not only change the neutronics behavior of the coolant, but require different fuel cladding and pressure-channel materials, specifically, stainless steels or Inconels, to withstand high-temperature steam. This paper will investigate the affect that steam, SS-304 and Inconel will have on neutron economy when introduced into a reactor design as well as required changes to fuel enrichment. It will also be necessary to investigate the effects of these material changes on power distribution inside a reactor. Pressure-channel design requires methods of fine control to maintain a balanced core-power distribution, the introduction of non-uniform coolant and reactor materials will further complicate maintaining uniform reactor power. The degree to which SHS channels will affect the power distribution is investigated in this paper.

Validation and Benchmarking of Comprehensive Vibration Assessment Program for Prototype Reactor Internals

2012 ◽  
Author(s):  
Jianfeng Yang ◽  
Lixin Yu ◽  
Byounghoan Choi
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Boiling Water ◽  
Pressurized Water ◽  
Water Reactor ◽  
Assessment Program ◽  
Water Reactors ◽  
The U.S ◽  
Reactor Internals

Reactor internals important to nuclear power plant safety shall be designed to accommodate steady-state and transient vibratory loads throughout the service life of the reactor. Operating experience has revealed failures of reactor internals in both pressurized water reactors (PWRs) and boiling water reactors (BWRs) due to flow-induced vibrations (FIVs). U.S. Nuclear Regulatory Commission (NRC) Regulatory Guide 1.20 presents a Comprehensive Vibration Assessment Program (CVAP) that the NRC staff considers acceptable for use in verifying the structural integrity of reactor internals for FIV prior to commercial operation. A CVAP supports the NRC reviews of applications for new nuclear reactor construction permits or operating licenses under 10 CFR Part 50, as well as design certifications and combined licenses that do not reference a standard design under 10 CFR Part 52. The overall CVAP should be implemented in conjunction with preoperational and initial startup testing. For prototype reactor internals, the comprehensive program should consist of a vibration and fatigue analysis, a vibration measurement program, an inspection program, and a correlation of their results. Validation and benchmarking processes should be integrated into the CVAP throughout each individual program. Based on the authors’ experiences in Advanced Boiling Water Reactor and AP1000® CVAPs and based on detailed reviews of the U.S. Evolutionary Power Reactor and the U.S. Advanced Pressurized Water Reactor CVAPs, this article summarizes the essential CVAP validation and benchmarking processes with proper consideration of bias errors and random uncertainties. This article provides guidance to a successful CVAP that satisfies the NRC requirements and ensures the reliability of the evaluation of potential adverse flow effects on nuclear power plant components.

Fast Neutron Reduction Techniques for Plant Lifetime Extension

2010 ◽  
Author(s):  
Choon Sung Yoo ◽  
Byoung Chul Kim ◽  
Tae Je Kwon
Keyword(s):  
Fast Neutron ◽  
Thermal Shock ◽  
Neutron Irradiation ◽  
Neutron Transport ◽  
Reactor Vessel ◽  
Pressurized Water ◽  
Reduction Techniques ◽  
Pressurized Thermal Shock ◽  
Water Reactors ◽  
Inner Vessel

A Pressurized Thermal Shock (PTS) Event is an event or transient in pressurized water reactors (PWRs) causing severe overcooling (thermal shock) concurrent with or followed by significant pressure in the reactor vessel. A PTS concern arises if one of these transients acts in beltline region of a reactor vessel where a reduced fracture resistance exists due to neutron irradiation. Such an event may produce a flaw or cause the propagation of a flaw postulated to exist near the inner vessel wall surface, thereby potentially affecting the integrity of the vessel. In this paper fast neutron flux reduction techniques were implemented to reduce the potential risk of PTS due to the neutron irradiation on the pressure vessel beltline region. And the RTPTS value for the end of life of the plant was projected using the fast neutron fluence obtained by neutron transport calculations according to the various core loading pattern and reduction program possible for the future cycles.

Risk Managed Technical Specification Implementation at South Texas Project Units 1 and 2

2008 ◽  
Author(s):  
Ernie Kee ◽  
Drew Richards ◽  
C. Rick Grantom ◽  
James K. Liming
Keyword(s):  
Technical Specification ◽  
Lessons Learned ◽  
South Texas ◽  
Pressurized Water Reactors ◽  
The South ◽  
Pressurized Water ◽  
Software Application ◽  
Guidance Documents ◽  
Water Reactors ◽  
Completion Times

Implementation of Risk Managed Technical Specification (RMTS) Initiative 4b, Risk-Informed Completion Times, for the two 3800 MW pressurized water reactors operated at the South Texas Project is described. Implementation issues encountered are presented and their resolutions described. The STP approach using a software application RICTCal to calculate risk informed completion times (RICTs) on the existing enterprise database to calculate risk informed completion times (RICTs) is described. Advantages, disadvantages, and lessons learned of the RMTS implementation are discussed. The methodology used to calculate completion times and maintain the database is described and a brief description of the developed documentation (procedures and guidance documents) for RMTS implementation at STP is given.

Aging Management Strategies for Pressurized Water Reactor Vessel Internals

2004 ◽  
Author(s):  
Timothy J. Griesbach ◽  
Robert E. Nickell ◽  
H. T. Tang ◽  
Jeff D. Gilreath
Keyword(s):  
Reactor Vessel ◽  
Performance Criteria ◽  
Pressurized Water Reactor ◽  
Aging Effects ◽  
Pressurized Water ◽  
Water Reactor ◽  
Flaw Tolerance ◽  
License Renewal ◽  
Technical Basis ◽  
Reactor Internals

Management of materials aging effects, such as loss of material, reduction in fracture toughness, or cracking, depends upon the demonstrated capability to detect, evaluate, and potentially correct conditions that could affect function of the internals during the license renewal term. License renewal applicants in their submittals to NRC have identified the general elements of aging management programs for Pressurized Water Reactor (PWR) internals, including the use of inservice inspection and monitoring with the possibility of enhancement or augmentation if a relevant condition is discovered. As plants near the license renewal term, plant-specific aging management programs will be implemented focusing on those regions most susceptible to aging degradation. A framework for the implementation of an aging management program is proposed in this paper. This proposed framework is based on current available research results and state of knowledge and utilizes inspections and flaw tolerance evaluations to manage the degradation issues. The important elements of this framework include: • The screening of components for susceptibility to the aging mechanisms, • Performing functionality analyses of the components with representative material toughness properties under PWR conditions, • Evaluating flaw tolerance of lead components or regions of greatest susceptibility to cracking, loss of toughness, or swelling, and • Using focused inspections to demonstrate no loss of integrity in the lead components or regions of the vessel internals. The EPRI Material Reliability Program (MRP) Reactor Internals Issue Task Group (RI-ITG) is actively working to develop the data and methods to quantify an understanding of aging and potential degradation of reactor vessel internals, to develop materials/components performance criteria, and to provide utilities tools for extending plant operations. Under this MRP Program, the technical basis for the framework will be documented. Then, based on that technical basis, PWR internals inspection and flaw evaluation guidelines will be developed for plants to manage reactor internals aging and associated potential degradation.

Evaluation of Laser Peening for Mitigation of Primary Water Stress Corrosion Cracking in Pressurized Water Reactors

2019 ◽  
Author(s):  
Stephen Marlette ◽  
Stan Bovid
Keyword(s):  
Surface Stress ◽  
Reactor Vessel ◽  
Pressurized Water Reactors ◽  
Laser Peening ◽  
Alloy 600 ◽  
Test Results ◽  
Pressurized Water ◽  
Performance Requirements ◽  
Primary Water ◽  
Water Reactors

Abstract For several decades pressurized water reactors have experienced Primary Water Stress Corrosion Cracking (PWSCC) within Alloy 600 components and welds. The nuclear industry has developed several methods for mitigation of PWSCC to prevent costly repairs to pressurized water reactor (PWR) components including surface stress improvement by peening. Laser shock peening (LSP) is one method to effectively place the surface of a PWSCC susceptible component into compression and significantly reduce the potential for crack initiation during future operation. The Material Reliability Program (MRP) has issued MRP-335, which provides guidelines for effective mitigation of reactor vessel heads and nozzles constructed of Alloy 600 material. In addition, ASME Code Case N-729-6 provides performance requirements for peening processes applied to reactor vessel head penetrations in order to prevent degradation and take advantage of inspection relief, which will reduce operating costs for nuclear plants. LSP Technologies, Inc. (LSPT) has developed and utilized a proprietary LSP system called the Procudo® 200 Laser Peening System. System specifications are laser energy of 10 J, pulse width of 20 ns, and repetition rate of 20 Hz. Scalable processing intensity is provided through automated focusing optics control. For the presented work, power densities of 4 to 9.5 GW/cm2 and spot sizes of nominally 2 mm were selected. This system has been used effectively in many non-nuclear industries including aerospace, power generation, automotive, and oil and gas. The Procudo® 200 Laser Peening System will be used to process reactor vessel heads in the United States for mitigation of PWSCC. The Procudo® 200 Laser Peening System is a versatile and portable system that can be deployed in many variations. This paper presents test results used to evaluate the effectiveness of the Procudo® 200 Laser Peening System on Alloy 600 material and welds. As a part of the qualification process, testing was performed to demonstrate compliance with industry requirements. The test results include surface stress measurements on laser peened Alloy 600, and Alloy 182 coupons using x-ray diffraction (XRD) and crack compliance (slitting) stress measurement techniques. The test results are compared to stress criteria developed based on the performance requirements documented in MRP-335 and Code Case N-729-6. Other test results include surface roughness measurements and percent of cold work induced by the peening process. The test results demonstrate the ability of the LSP process to induce the level and depth of compression required for mitigation of PWSCC and that the process does not result in adverse conditions within the material.

Emergency Response Guidance for Reactor Vessel Pressurized Thermal Shock Events

1986 ◽  
Vol 108 (3) ◽  
pp. 346-351
Author(s):  
W. T. Kaiser ◽  
B. S. Monty
Keyword(s):  
Thermal Shock ◽  
Temperature Limit ◽  
Reactor Vessel ◽  
Pressurized Water ◽  
Plant Safety ◽  
Pressurized Thermal Shock ◽  
Water Reactors ◽  
Accident Conditions ◽  
Core Cooling ◽  
Emergency Operating

The operational concern of pressurized thermal shock (PTS) can be minimized by proper operator guidance. This paper presents a method for calculating a pressure temperature limit curve for reactor vessel integrity which can be used to identify an ongoing potential PTS event. This method has been developed for use and is applicable to all pressurized water reactors. The curve is used in emergency operating procedures developed to prioritize various plant safety concerns including PTS and core cooling to ensure proper operator action during accident conditions. This paper emphasizes the development of the pressure-temperature limit and how it is used within the emergency operating procedures.

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