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Westinghouse Advanced Loop Tester (WALT) Update

[+] Author Affiliations
G. Wang, W. A. Byers

Westinghouse Electric Company LLC, Pittsburgh, PA

M. Y. Young, Z. E. Karoutas

Westinghouse Electric Company LLC, Columbia, SC

Paper No. ICONE16-48480, pp. 469-475; 7 pages
doi:10.1115/ICONE16-48480
From:
  • 16th International Conference on Nuclear Engineering
  • Volume 3: Thermal Hydraulics; Instrumentation and Controls
  • Orlando, Florida, USA, May 11–15, 2008
  • Conference Sponsors: Nuclear Engineering Division
  • ISBN: 0-7918-4816-7 | eISBN: 0-7918-3820-X
  • Copyright © 2008 by ASME

abstract

In order to understand crud formation on the fuel rod cladding surfaces of pressurized water reactors (PWRs), a crud Thermal-Hydraulic test facility referred to as the Westinghouse Advanced Loop Tester (WALT) was built at the Westinghouse Science and Technology Department Laboratories in October 2005. Since then, a number of updates have been made and are described here. These updates include heater rod improvements, system pressure stabilization, and more effective protection systems. After these updates were made, the WALT system has been operated with higher stability and fewer failures. In this test loop, crud can be deposited on the heater rod surface and the character of the crud is similar to what has been observed in the PWRs. In addition, chemistry in the WALT loop can be varied to study its impact on crud morphology and associated parameters. The WALT loop has been successful in generating crud and measuring its thermal impact as a function of crud thickness. Currently, this test facility is supporting an Electric Power Research Institute (EPRI) program to assess the impact of zinc addition to PWR reactor coolant. Meanwhile, the WALT system is also being utilized by Westinghouse to perform dry-out and hot spot tests. These tests support the industry goal of 0 fuel failures by 2010 set by Institute of Nuclear Power Operations (INPO). Another major goal of the Westinghouse tests is to gain a better understanding of unexpected changes in core power distributions in operating reactors known as crud induced power shifts (CIPS) or axial offset anomalies (AOA).

Copyright © 2008 by ASME

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