Westinghouse Nuclear

Westinghouse In-Core Information Surveillance & Engineering (WINCISE™)

admin — Tue, 08 May 2007 18:01:28 GMT

WINCISE™ is an Operational Support System that uses Westinghouse technology licensed by the NRC to obtain an accurate, continuous core power distribution measurement. Westinghouse developed the WINCISE System as a replacement for the Movable Detector (M/D) System. WINCISE uses an OPARSSEL™ (Optimized Proportional Axial Region Signal Separation Extended Life) detector assembly that contains five Vanadium detector elements and a core exit thermocouple (CET). The integral CET eliminates maintenance and resulting personnel dose and critical path time associated with the reactor vessel head mounted CET.

The power distribution monitoring software (PDMS) used to continuously process the detector signals is the Westinghouse BEACON™ System, licensed by the NRC for use with the OPARSSEL design to produce accurate core power distribution measurements. The OPARSSEL design nearly eliminates detector material depletion and mechanical failures, which are major costs associated with the use of other fixed in-core detector system designs.

Uprating Programs

admin — Tue, 08 May 2007 17:59:03 GMT

Nuclear power plant upratings provide incremental electric generation capacity in a timely and cost-effective manner. Westinghouse has successfully implemented over 70 plant upratings providing over 2900 MWe of additional power generation.

Three types of uprating programs are currently available:

• Measurement uncertainty recapture (MUR) uprates take advantage of improved power measurements to lower the power calorimetic uncertainty.

• Stretch power uprates (SPUs) raise power to within the design capacity of the plant.

• Extended power uprates (EPU require significant plant modification to the balance of plant (BOP) equipment). These uprates are typically 7 to 20%.

Upper Head Temperature Reduction (UHTR) Program

admin — Tue, 08 May 2007 17:57:06 GMT

Alloy 600 primary water stress corrosion cracking (PWSCC) is very sensitive to temperature. Lowering temperatures can be an effective way to increase the time for crack initiation and reduce the rate of crack propagation, thereby extending the useful life of a reactor vessel head (RVH). The Upper Head Temperature Reduction (UHTR) Program, also known as Tcold Conversion, is a method of lowering the RVH temperature of a Thot head, and is already in effect at a number of plants. UHTR can be evaluated as a cost-effective alternative, or as a method for delaying a planned RVH replacement.

Tube Integrity & Alternate Repair Criteria

admin — Tue, 08 May 2007 17:55:03 GMT

Since the early 1980s, Westinghouse has been recognized as the industry leader with regard to engineering analysis of operating steam generator degradation. We've gained experience and developed tools to accurately assess degradation and optimize operational efficiency when degradation mechanisms exist. Westinghouse has also led the industry in developing steam generator tube alternate repair criteria (ARC) for keeping degraded tubes in service. We continue to make advancements in flaw modeling and integrity projection analysis (operational assessment) methods, consistent with the current industry guidelines and in development of improved eddy current (EC) analysis techniques to support operational assessments.

Steam Generator Tube Stabilizer - Tube Damper

admin — Tue, 08 May 2007 17:53:09 GMT

The Westinghouse Steam Generator Tube Stabilizer System evolved from analytical and experimental efforts to effectively stabilize degraded steam generator tubes. Because tubes can be susceptible to vibration, the Tube Stabilizer System was tested to assess its effectiveness, characterize the vibration and damping of a stabilized tube, and characterize the in-service life of the system. The Steam Generator Tube Stabilizer System was verified and qualified.

Simplified Head Assembly (SHA)

admin — Tue, 08 May 2007 17:51:03 GMT

Reactor head assembly and re-assembly 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 developed to effect a significant improvement in this outage phase. The IHP includes features specifically designed to reduce the efforts associated with disassembling and re-assembling the reactor head in support of plant refueling.

Recently, Westinghouse has developed simplified head assembly (SHA) designs that costeffectively back-fit the outage optimizationfeatures of the IHP into an operating plant.

Replacement Steam Generator (RSG) Feedring Design

admin — Tue, 08 May 2007 17:48:55 GMT

Primary drivers for steam generator replacement include a utility's desire to reduce or eliminate maintenance costs due to Alloy 600 issues, uprate power, and realize operational benefits associated with advances in steam generator design. A number of plants have already made the decision to replace steam generators, while others are giving steam generator replacement serious consideration.

Replacement Steam Generator (RSG) Axial Flow Preheater (AXP) Design

admin — Tue, 08 May 2007 17:46:30 GMT

Replacement Pressurizer

admin — Tue, 08 May 2007 17:44:36 GMT

The primary driver for pressurizer replacement is a plant's desire to reduce or eliminate downtime due to pressurizer heater sleeve leaks or other Alloy 600 leaks that have plagued the industry for some time. A number of plants have already made the decision to replace their pressurizers; others are giving replacement serious consideration. Replacement pressurizers (RPZRs) eliminate Alloy 600 nozzles, thereby eliminating the downtime.

Reactor Vessel Closure Head Replacement

admin — Tue, 08 May 2007 17:42:57 GMT

In response to industry events involving Alloy 600 material in reactor vessel closure head (RVCH) penetrations, a number of pressurized water reactor (PWR) utilities have, or intend to, replace the RVCH. In addition to the potential for eliminating material issues associated with Alloy 600, the RVCH replacement provides an ideal opportunity to implement upgrades to significantly reduce outage time and dose, as well as address personnel safety issues that may exist during reactor disassembly and reassembly.

Westinghouse has developed a program to design and implement RVCH upgrades integrated with the design and installation of the new RVCH.

Permanent Cavity Seal Ring (PCSR)

admin — Tue, 08 May 2007 17:41:08 GMT

Conventional cavity seal rings must be installed and removed during each outage.They are difficult to handle and allow significant leakage past the seals. Westinghouse has designed the Permanent Cavity Seal Ring (PCSR), a permanently installed, stainless steel structure with flexures to compensate for thermal growth and seismic movement.

Mechanical Stress Improvement Process

admin — Tue, 08 May 2007 17:39:28 GMT

Pressurized water reactor (PWR) plants have experienced primary water stress corrosion cracking (PWSCC) since the 1980s. PWSCC occurs in Alloy 82/182 welds as well as Alloy 600 components. Left untreated, PWSCC can compromise the operations and economics of light water reactors.

Westinghouse has teamed with AEA Technology Engineering Services Inc. (AEA) to provide the Mechanical Stress Improvement Process (MSIP®) –a long-term solution to PWSCC in reactor vessel hot-leg nozzle piping, pressurizer nozzles, and other small bore pipes.

Mechanical Nozzle Seal Assembly (MNSA)

admin — Tue, 08 May 2007 17:37:10 GMT

Instrument nozzle penetrations are common to pressurizer and hot/cold leg reactor coolant system (RCS) piping. Inconel 600 partial penetration welded nozzles typical to these areas are susceptible to stress corrosion cracking (SCC). Small leaks in these nozzles can cause expensive delays in plant operation.

As a result, Westinghouse has developed the Mechanical Nozzle Seal Assembly (MNSA), a unique mechanical seal that has effectively eliminated leakage caused by primary water stress corrosion cracking (PWSCC) in Inconel 600 nozzles under 2 inches in diameter.

Lower Internals Upflow Conversion

admin — Tue, 08 May 2007 17:34:52 GMT

Flow leakage through the axial gap between the vertical baffle plates has caused fuel rod damage in some early generations of Westinghouse plants that have a downflow configuration. These vertical baffle plates are bolted (using baffle to former bolts) to horizontal plates called formers, which in turn are bolted (using former to barrel bolts) to the cylindrical core barrel. The core barrel contains flow holes located in the upper core region elevations, which permit a small amount of reactor coolant to flow downward through the region between the baffle plates and the core barrel.

In these downflow plants, the pressure drop across the baffle plates has been sufficient to force coolant through any existing gaps between the edges of adjacent baffle plates, sometimes causing flow-induced vibration (FIV) of fuel rods close to the baffle gaps. This FIV causes wear and, in some cases, fuel rod failures. Changing the configuration in these plants from downflow to upflow reduces the pressure drop across the baffle plates, thus eliminating the potential for fuel rod degradation due to baffle jetting.

Later generations of commercial nuclear power plants incorporated an upflow configuration in the original design and construction and are not susceptible to baffle jetting.

Guide Tube Support Pin Replacement

admin — Tue, 08 May 2007 17:31:13 GMT

In the early 1980s, Alloy X-750 guide tube support pins originally installed in the guide tubes in upper internals began exhibiting cracks. In some cases, the pin shank and the nut securing the pin broke off, migrated to the steam generator(s), and damaged the tube sheet.