Aging management is a technical process that provides reasonable assurance that the aging of important nuclear power plant systems, structures and components (SSC) is being managed so they will continue to perform their intended function(s). Outside the United States, the aging-management process is being utilized to confirm the aging-management elements of the International Atomic Energy Agency’s (IAEA) Periodic Safety Review (PSR) process. Most member states require a PSR review once every 10 years for continued plant operation. In the United States, the aging-management process provides compliance with the requirements for obtaining a renewed operating license. License renewal is a regulatory process that allows a nuclear plant in the United States to extend its operating license for an additional 20 years (beyond the 40 years of its original license). Whether applied to the PSR or license renewal, this process provides an extremely cost-effective way to assure available generation capability into the future.

Westinghouse provides balance of plant (BOP) engineering services that deliver solutions across the entire plant. From specialty consulting to integrated, comprehensive solutions for engineering projects, Westinghouse’s BOP engineering and project management teams understand customer needs and address them to optimize plant performance.

Westinghouse provides full-scope blast analysis and design capability to address needs for commercial and critical infrastructure clients.

Small leaks in pressurized water reactor (PWR) head penetrations can prevent a nuclear power plant from returning to power and cause expensive delays until a fix is devised. An increasing number of plants are reporting primary coolant leaks in the field-welded canopy seal area. To control these kinds of leaks, Westinghouse offers a full range of products and services including a unique mechanical clamp assembly named the Canopy Seal Clamp Assembly (CSCA™).

As the nuclear fleet ages, Westinghouse- and Combustion Engineering (CE)-designed plants are experiencing a greater number of Control Rod Drive Mechanism (CRDM)- and Control Element Drive Mechanism (CEDM)-related issues. These issues can range from improper polarity between coils; failed splices within coils; past operation of coils at excessive temperatures or current causing turn-to-turn shorts; cable and connector degradation; latch assembly wear; crud induced mis-stepping [2]; and failed latch assembly springs [1].

Combustion Engineering (CE) Control Element Drive Mechanism (CEDM) coils have been in use for more than 40 years with a respectable operating history. However, the upper gripper coil, which is normally energized continuously in order to hold the control rod at the full out position, has been found to be under particular stress. As a result, the CE CEDM upper gripper coil is prone to early failures, accelerated by excessively high voltage or exposure to temperatures above its design value.

Westinghouse has developed replacement Combustion Engineering (CE) Control Element Drive Mechanism (CEDM) High Temperature (HT) Upper and Lower Gripper Coils that have much higher temperature capability than the original CEDM gripper coil design. CE CEDM coils from the standard CE design have been in use for more than 40 years with a respectable operating history. However, the upper gripper coil, which is normally energized continuously in order to hold the control rod at the full out position, is under particular stress. As a result, the CE CEDM Upper Gripper Coil may fail after several years of service. Its failure is accelerated by excessively high voltage or exposure to temperatures above its design value.

Chemistry performance in a nuclear power plant strongly influences the efficiency of power operation, refueling outages, and routine maintenance. Utilities’ chemistry needs range from approval of consumable compatibility to a complete range of chemistry support for operations and outages.

Instances of flux thimble tube wear and leakage observed in operating Westinghouse designed reactors with bottom mounted instrumentation have been attributed to flow-induced vibrations in the lower internals support column area. As a result, the U.S. Nuclear Regulatory Commission (NRC) issued Bulletin 88-09, “Thimble Tube Thinning in Westinghouse Reactors,” which instructs affected utilities to establish and implement inspection programs to periodically confirm thimble tube integrity.

Westinghouse provides full-scope civil and structural engineering capabilities and offers a broad range of solutions for the nuclear power industry. Westinghouse has a long history of providing innovative solutions to address customer needs to meet code, regulatory, and other unique requirements.

The BOP and Design Engineering civil structural and geotechnical engineering services is a proven and consistent supplier of civil and geotechnical design and services for power generation, oil & gas, and industrial infrastructure projects.

In 1995, the U.S. Nuclear Regulatory Commission (NRC) provided an Option B to Title 10 Code of Federal Regulations (CFR) Part 50, Appendix J, which is a performance-based approach to leakage testing requirements that allows licensees with acceptable test performance history to change surveillance frequencies. Addressed with the performance-based approach are the surveillance frequencies for Type A, B and C tests. The Type A test assesses the overall leakage of containment. The Type B test assesses leakage for containment penetrations. The Type C test assesses leakage for containment isolation valves. At that time, provisions were made for extending the Type A test (integrated leak rate test [ILRT]) frequency from 3-in-10 years to 1-in-10 years.

The thermocouple nozzles on Westinghouse reactor vessel heads have two primary pressure boundary seals that have to be disassembled during each refueling outage: an upper and a lower Conoseal® joint. Each of these joints uses a Conoseal metal seal for the pressure boundary. If a seal fails during a refueling outage, the system has to be depressurized and drained below the seal elevation. After replacing the seal, the system must be refilled and vented, adding more than a day to critical path work and resulting in a significant increase in man-rem exposure.

The Westinghouse BOP and Design Engineering electrical engineering services is a proven and consistent supplier of electrical design and services for power generation, oil & gas, and industrial infrastructure projects. To simplify contracting activities, Westinghouse provides a wide range of electrical engineering services from simple engineering staff augmentation to integrated, comprehensive solutions for large, complex electrical designs to component replacements and plant-wide modification projects.

The Westinghouse BOP and Design Engineering Product Line has performed major engineering and design, plant modifications and upgrade work at many PWR and BWR nuclear stations. We have proven experience and detailed lessons learned developed from many projects that provide our clients with an extensive pool of resources best prepared to apply industry best practices honed through hands-on background.

All nuclear power plants must consider and evaluate external flooding risks such as flash flooding from rain, river flooding, dam failure, hurricane and tsunami. These events challenge off-site power, threaten many on-site plant mitigation components, challenge the integrity of plant structures and limit plant access. Plants must understand the impact of these events in order to fully comprehend and prepare for these plant risks. Existing plant mitigation procedures may not be adequate to deal with these types of events.

A Fire PRA is required to implement plant transition to NFPA 805, as well as to meet NRC Regulatory Guide (RG) 1.200 requirements (i.e., technical adequacy of PRA results for risk informed activities). NFPA 805 is a standard developed by the National Fire Protection Association that provides a risk-informed, performance-based alternative to a plant’s current fire protection program. NUREG/CR-6850 is the NRC-endorsed guidance for developing a Fire PRA that meets the Fire PRA Standard, ANS 58.23.

The Westinghouse Fire Risk Services Team is a one-stop shop for all aspects of nuclear power plant fire protection and fire risk assessment.

Regulatory Guide (RG) 1.200 endorses the American Society of Mechanical Engineers (ASME) consensus standard for internal events probabilistic risk assessment (PRA), which includes a set of minimal requirements for PRA modeling of large early release frequency (LERF). As the use of PRA in the nuclear industry matures, the capability of plant-specific LERF models, particularly those used in regulatory operations, must meet these standards.

The Westinghouse Solution Westinghouse is committed to bringing efficient, large-scale hydrogen production to nuclear facilities through operating plant integration and advanced reactor designs. Clean hydrogen supports societal decarbonization while yielding a significant, yet flexible revenue stream to utilities around the world. Westinghouse is positioned to be a full-scale hydrogen partner, maximizing power output, modernizing plants for long term operations and monetizing hydrogen production.

Westinghouse provides an independent review of probabilistic risk assessment (PRA) program studies to help customers show that their PRAs meet applicable technical-quality requirements. Per Revision 2 of Regulatory Guide 1.200, PRAs used for risk-informed regulatory applications must meet certain technical adequacy and quality requirements, as determined by the American Society of Mechanical Engineers (ASME) and the American Nuclear Society (ANS) PRA Standard (RA-S-2008).

The Westinghouse BOP and Design Engineering instrumentation and control systems (I&C) engineers have decades of experience supporting power generation, oil & gas, and industrial infrastructure projects. Westinghouse provides a wide range of services for integrated, comprehensive solutions for small minor modifications up to and including large, complex component replacements and plant-wide modification projects.

Background During a severe accident or a beyond-design-basis accident (BDBA), the reaction of water with zirconium alloy fuel cladding, radiolysis of water, molten coriumconcrete interaction (MCCI) and post-accident corrosion can generate hydrogen (H2). The total mass of H2 produced in-vessel depends on several factors. For most reactors, it is on the order of 1,000 kilograms. High peak rates for H2 release to the containment of up to several kg/s can result from discontinuous releases from the reactor pressure vessel. The detonation of H2 can result in damage to structures such as containment buildings or spent fuel buildings. In all reactor designs, H2 monitors can be utilized to monitor the risk of containment or spent fuel building damage due to H2 detonations.

An internal flooding (IF) risk assessment refers to the quantitative probabilistic risk assessment (PRA) treatment of flooding as a result of pipe and tank breaks inside the plants, as well as from other recognized flood sources. The industry consensus standard for Internal Events Probabilistic Risk Assessment (American Society of Mechanical Engineers/American Nuclear Society [ASME/ANS] RA-Sa-2009) includes high-level and supporting technical requirements for developing an internal flooding PRA (IF-PRA).

With nuclear power plants (NPPs) extending their licenses, there is a large backlog of work that must be performed prior to entering the period of extended operation. After the U.S. Nuclear Regulatory Commission (NRC) grants a plant a new license, the plant must fulfill a significant number of commitments before entering the period of extended operation. To satisfy all the commitments, many smaller tasks must be completed. Reductions in staff and a subsequent loss of experience are issues when completing the required commitments.

The increased focus on probabilistic risk assessments (PRAs) and risk-informed applications by the U.S. Nuclear Regulatory Commission (NRC) has led utilities to expand their PRA model update efforts. Their efforts have initially focused on updating the Level 1 and Level 2 PRA models for the internally initiated events.

At its nuclear parts operations shop in New Stanton, Pennsylvania (USA), Westinghouse provides full machine shop capabilities, offering both computer numerically controlled (CNC) and manual machine lathes as well as vertical milling machines (two-axis and three-axis CNC machining). The full-time tool and die makers who operate the machine shop have a combined 250 years of experience and can accommodate prototyping, one-piece runs and both low- and high-quantity production runs. The Westinghouse machine shop is certified for American Society of Mechanical Engineers (ASME) machining.

The Westinghouse BOP and Design Engineering mechanical engineers have decades of experiences providing support to a wide range of engineering activities at power generation, oil & gas, and industrial infrastructure projects. Westinghouse provides a wide range of services for integrated, comprehensive projects from minor modifications to complex component replacements and plant-wide modifications.

The nuclear industry continues to experience significant pressure to reduce costs, yet the safe and efficient operation of nuclear power plants requires well-trained, highly competent staff. Research evidence from Mind, Brain and Education (MBE) science can improve the efficiency of nuclear industry training practices; but, instructors and instructional designers must first understand these findings to align content and instruction with how people learn most effectively and efficiently.

Background Westinghouse maintains responsibility for the final safety analysis report (FSAR), and nonloss- of-coolant accident (LOCA) analyses for numerous Westinghouse-, Combustion Engineering- (CE-) and non Westinghousedesigned plants worldwide, including the System-80+ and the Westinghouse AP1000™ plant designs.

The annulus gap between the reactor vessel and the containment cavity floor must be sealed to permit the flood-up required for refueling and reactor internal maintenance activities. This sealing is accomplished by installing a temporary pool seal into the annulus gap. The installation of these “temporary” seals is a critical path process that lengthens outage duration as well as increases worker exposure. These seals have also been known to experience leakage, which negatively impacts both outage and normal operation processes.

The Westinghouse BOP and Design Engineering piping analysis group has decades of experience in piping design and modification analysis in the power generation and other industrial facilities. The piping group maintains full engineering capabilities for the design and analysis in both Safety-Related and Balance of Plant applications.

Westinghouse provides Plant Licensing (PL) support to aid individual safety analyses groups in licensing activities, from preparing engineering/licensing reports to supporting Request for Additional Information (RAI) from the U.S. Nuclear Regulatory Commission (NRC).

The Plant Process Computer system (PPC), or plant computer, is a plant-wide information system for new and retrofit plants. The PPC consists of data acquisition and presentation layer components, with configurable, reusable software programs for performing nuclear plant performance and monitoring applications. PPC uses a redundant network design with advanced connectivity features that provides high capacity data transmission and reliable external system communications via standard and custom protocols.

The plant protection system monitors plant temperatures, pressures, levels, flows and nuclear instrumentation system outputs. If these parameters exceed plant safety limits, the system issues “partial reactor trip” and “engineered safeguards” commands. The plant protection system sends isolated analog output signals to the plant control system, the data processing and monitoring system, and the main control board. It also provides alarm outputs to the plant annunciator system. The plant protection system can be supplied with an interface to the data processing and monitoring system, or implemented as a standalone upgrade to the existing plant system. When supplied with an interface, a phased approach permits small stand-alone upgrades to be eventually integrated into a total plant information network.

Many existing Computer-aided Fault Tree Analysis (CAFTA) and Fault Tree Reliability eXpert (FTREX)-based Probabilistic Risk Assessment (PRA) models have been developed, expanded and updated over several years without a strong focus on the ultimate use of the models.

As the probabilistic risk assessment (PRA) and risk applications quickly become part of the fabric of plant operation and licensing, the time demands on a utility’s current PRA staff are continually increasing to support workday scheduling, outage planning and emergent plant configurations.

Starting with the first commercial Westinghouse-design nuclear power plants, Westinghouse has been involved in the development of generic, as well as plant specific, guidance for response to plant events. Following the Three Mile Island (TMI) accident, the U.S. Nuclear Regulatory Commission (NRC) issued NUREG-0899, which provided requirements for utility preparation and implementation of emergency operating procedures (EOP), including development, writing and maintenance. This was followed by NUREG-1358, in which the NRC reinforced its expectations with respect to the plantspecific technical guidelines, EOP writers guide, EOP verification and validation (V&V) and EOP training. Together, these regulations comprise the plant-specific procedure generation package (PGP).

Operators of nuclear power plants frequently face problems that require accurate characterization and analysis of radiation. The wide-ranging difficult issues that can arise can be far beyond the scope of typical radiation analyses, such as nuclear fuel design and accident analysis services. The Westinghouse team of radiation experts can answer the tough questions on radiation analysis, and Westinghouse offers a suite of products to assist with radiation and thermal measurements.

The Westinghouse BOP and Design Engineering Radiological Engineering Analyses team is comprised of specialists with extensive technical and licensing backgrounds covering the disciplines of radiological engineering, nuclear engineering, mechanical engineering, chemistry, and physics.

Primary water stress corrosion cracking (PWSCC) of Alloy 600 materials and instrument nozzle Alloy 182/82 welds has become a top industry concern for PWR plants. PWSCC has produced significant losses in power generation and attracted considerable regulatory attention. There are many locations within the reactor coolant pressure boundary (RCPB) that can contain Alloy 600 base metal or weld metal that can be susceptible to PWSCC over time. Additionally, previously replaced instrument nozzles may also leak due to weld flaws.

Primary water stress corrosion cracking (PWSCC) of Alloy 600 materials and bottom mounted instrumentation (BMI) Alloy 182/82 welds has become a top industry concern for pressurized water reactor (PWR) plants. PWSCC has produced significant availability losses and attracted considerable regulatory attention. There are many locations within the reactor coolant pressure boundary (RCPB) that contain Alloy 600 base metal or weld metal that could be susceptible to PWSCC over time.

Nuclear power plants with reactor vessel closure heads (RVCHs) containing Alloy 600 base materials and Alloy 182 weld materials are susceptible to primary water stress corrosion cracking (PWSCC). In response to this concern, a number of PWR utilities have replaced their RVCHs. Replacements also provide an ideal opportunity to implement upgrades; this significantly reduces outage duration and dose, as well as addresses personnel safety issues that may exist during reactor disassembly and reassembly. To offer our customers a solution to this problem, Westinghouse has created a program to develop and implement RVCH upgrades integrated with the design and installation of a new RVCH that uses Alloy 690 and Alloy 152. Because these alloys aren’t prone to PWSCC, this is a risk- reducing option.

Background Westinghouse provides post-irradiation testing and evaluation of the reactor vessel material specimens, thermal monitors and dosimeters contained in the surveillance capsules to monitor the effects of neutron irradiation on the reactor vessel beltline materials under actual operating conditions.

Westinghouse drives the global risk industry in delivering well-documented Probabilistic Risk Analysis (PRA) models that effectively balance detail, execution time and the required engineering skill necessary to effectively interpret and communicate the risk insights derived from these models.

The key purpose of an in-service inspection (ISI) is to identify a flaw before it becomes a structural failure. In general, inspections have historically been performed based on such mandated requirements as those for nuclear power plant components in the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, insurance requirements or company policy.

For the past 10 years, the nuclear industry and the U.S. Nuclear Regulatory Commission (NRC) have been working together to develop Risk-informed Technical Specifications (RITS) to enhance plant safety and improve plant operations. Of the eight initiatives set forth by the partnership, many are currently available for plant implementation. The remaining initiatives will be available in the near future.

For the past 15 years, the nuclear industry and the U.S. Nuclear Regulatory Commission (NRC) have been working together to develop Risk-informed Technical Specifications (RITS) to enhance plant safety and improve plant operations. Following analysis and methodology development by the industry and subsequent approval by the U.S. NRC, the initiatives set forth by this partnership are now essentially complete and are available for plant implementation. The implementation of the initiatives is facilitated by the Technical Specification Task Force travelers (TSTFs).

For the past 15 years, the nuclear industry and the U.S. Nuclear Regulatory Commission (NRC) have been working together to develop Risk-informed Technical Specifications (RITS) to enhance plant safety and improve plant operations.

Reactor Vessel Closure Head (RVCH) disassembly and reassembly activities are major considerations when it comes to a refueling outage’s critical path schedule, personnel radiation exposure, critical containment resources, foreign material exclusion (FME) control, personnel safety and cost. Ductwork associated with the cooling of the control rod/element drive mechanisms has to be disassembled every outage and then reassembled again prior to start up. To help our customers improve this process, Westinghouse offers a solution that reduces outage duration, polar crane dependency, personnel risk, dose and overall manhours.

Designed as a diagnostic planning tool, Westinghouse’s Secondary Side Condition Monitoring and Operational Assessment (SS-CMOA) is a living document that evaluates the secondary side of the steam generator (SG) and interfacing systems.

For decades, Westinghouse has been supporting the nuclear industry as a full-scope seismic probabilistic risk assessment (PRA) provider, offering capabilities ranging from risk-analysis and risk-informed applications, new plant licensing and Post-Fukushima requirments.

Following the Three Mile Island Unit 2 accident, the U.S. Nuclear Regulatory Commission (NRC) developed a plan (NUREG-1050 – August 1985) to resolve the severe accident generic issue. This plan identified that utility commitment to excellence in risk management, including prevention and mitigation, is key to protection of public health and safety; it also identified the need for new severe accident research.

Reactor head disassembly and reassembly activities are major considerations when it comes to the refueling outage critical path schedule, personnel radiation exposure, critical containment resources, personnel safety and cost. The Westinghouse integrated head package (IHP) is an enhanced equipment design that offers a significant improvement in outage time. The IHP includes features specifically designed to reduce the efforts associated with disassembling and reassembling the reactor head in support of plant refueling.

Westinghouse’s has several hundred Specialists and Subject Matter Experts (SME) many of which are nationally recognized experts who support a broad client base to resolve emergent technical issues. Our SME’s provide expert advice, analyses and or engineering and design to implement plant modifications or resolve or improve plant performance issues.

Westinghouse provides full-scope structural capability, including seismic and dynamic analysis and design to address utility needs for both nuclear safety-related and conventional industrial buildings and structures.

Owners need augmented technical and managerial support from time to time to support both planned projects and emergent plant issue(s). Staff or management augmentation support to our clients can take various forms and can accommodate diverse options unique to the client’s organization and situational needs.

Westinghouse offers steam generator (SG) engineering services in the areas of component design and analysis, chemistry, diagnostics, and materials engineering, with the mission to: Provide engineering solutions that extend the life of the SGs, optimize plant performance and reduce the overall cost to maintain SGs, while meeting regulatory requirements. Integrate engineering with field services to provide coordinated inspection, repair and engineering services to optimize performance and extend the life of the SGs. Provide best-practice engineering analyses in support of plant performance improvement programs. Apply leading-edge technology to support utility asset management programs. Provide industry licensing leadership through development of low-risk licensing strategies.

Secondary side tube deposits can have an adverse effect on steam generator (SG) operation. Unless properly maintained, SGs can be subject to performance degradation as a result of tubing corrosion and steam pressure reduction.

Virtually all nuclear utilities are facing ever-increasing personnel and financial pressures. The aging work force and demand for talent from regulators and new nuclear plants are creating skill gaps. Financial pressures are increasing in today’s sluggish economy, forcing utilities to optimize the size and skill sets of their staffs.

The Westinghouse BOP and Design Engineering Thermal-Hydraulic Engineering and Safety Analyses team is comprised of specialists with extensive technical and licensing background covering the disciplines of heat transfer, fluid flow, nuclear engineering, and mechanical engineering. Most of the engineers have over 30 years of experience in the nuclear power industry and were part of the Balance-of-Plant (BOP) Architect Engineering (AE) teams that were responsible for putting numerous nuclear power plants online.

In response to the events at Fukushima Daiichi, the U.S. Nuclear Regulatory Commission (NRC) issued an interim staff guidance document (JLDISG-2012-01) that places increasing emphasis on analysis of external events including high winds. Additionally, the U.S. NRC issued Regulatory Issue Summary (RIS) 2015-06, “Tornado Missile Protection,” that addresses conformance to a plant’s current site-specific licensing basis for tornado-generated missiles and notes its acceptance of License Amendment Requests using probabilistic risk assessment (PRA) methodologies and computer tools to simulate tornado-generated missiles.

The Westinghouse Solution Nuclear power plant upratings are a timely and cost­ effective way to provide incremental electric generation. Westinghouse has successfully implemented more than 150 plant upratings, providing more than 5000 MWe of additional power generation.

Nuclear power plant uprating is a timely and cost-effective way to provide incremental electric generation. Westinghouse has successfully implemented more than 150 plant upratings, providing more than 5000 MWe of additional power generation worldwide.

The goal of Valve Program Management (VPM) is to maintain the safety-related and important-to-safety-related valves’ reliability so that they can perform the design basis requirements for the life of the plant. While this may be simply stated, the actual management and implementation of the program is a complex and comprehensive task that, if performed effectively, results in increases in both plant reliability and capacity factors.

The Westinghouse engineers, technicians and staff who specialize in evaluations through laboratory testing of irradiated and non-irradiated materials provide experimental evidence to support materials and processing solutions for its customers while supporting industry technical initiatives.