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.

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 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.

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.

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.

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 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.