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Determination of the hydrogen heat of transport in Zircaloy-4

Journal Article · · Journal of Nuclear Materials
 [1];  [2];  [3];  [3];  [2];  [4];  [5]
  1. Pennsylvania State Univ., University Park, PA (United States); OSTI
  2. Univ. of Michigan, Ann Arbor, MI (United States)
  3. Univ. of Michigan, Ann Arbor, MI (United States); Eidgenoessische Technische Hochschule (ETH), Zurich (Switzerland); Paul Scherrer Inst. (PSI), Villigen (Switzerland)
  4. Bettis Atomic Power Laboratory (BAPL), West Mifflin, PA (United States)
  5. Pennsylvania State Univ., University Park, PA (United States)
+ Show Author Affiliations
During operation in a nuclear reactor, Zr-based nuclear fuel cladding is subject to waterside corrosion which can lead to hydrogen ingress. Here, the hydrogen that enters the material will migrate to colder spots and precipitate as zirconium hydrides if the hydrogen content exceeds the hydrogen terminal solid solubility in the material. Since a temperature gradient is established in the radial direction of the cladding during operation, the hydrides can preferentially precipitate at the colder outer surface of the cladding. Other gradients can also occur in the longitudinal and azimuthal directions of the cladding tube. As a consequence, hydrogen redistributes itself in response to the concentration and temperature gradients present in the sample. The response of the hydrogen in solid solution to temperature gradients is governed by the heat of transport Q* as a function of temperature, so it can be used in the BISON code. A set of experiments was set up to determine the heat of transport (Q*), in which a uniformly hydrided Zircaloy-4 sample is annealed under a fixed temperature gradient at a range of temperatures, and the resulting hydrogen distribution is analyzed to determine Q*. The results are discussed in terms of existing literature.
Research Organization:
Pennsylvania State Univ., University Park, PA (United States)
Sponsoring Organization:
USDOE; USDOE Office of Nuclear Energy (NE), Nuclear Energy University Program (NEUP)
Grant/Contract Number:
NE0008717
OSTI ID:
2417711
Alternate ID(s):
OSTI ID: 1898445
Journal Information:
Journal of Nuclear Materials, Journal Name: Journal of Nuclear Materials Journal Issue: C Vol. 573; ISSN 0022-3115
Publisher:
ElsevierCopyright Statement
Country of Publication:
United States
Language:
English

References (5)

Zirconium Alloys in Nuclear Applications book September 2006
Hydrogen in zircaloy-2: Its distribution and heat of transport journal December 1960
Modeling and simulation of hydrogen behavior in Zircaloy-4 fuel cladding journal September 2014
Hydrogen in zirconium alloys: A review journal May 2019
Predicting the hydride rim by improving the solubility limits in the Hydride Nucleation-Growth-Dissolution (HNGD) model journal January 2022

Figures / Tables (13)

Table 1(p. 5)table Table 1
Figure 3(p. 6)figure Figure 3
Figure 6(p. 10)figure Figure 6
Table 2(p. 13)table Table 2
Figure 2(p. 4)figure Figure 2
Figure 5(p. 8)figure Figure 5
Table 3(p. 15)table Table 3
Figure 1(p. 4)figure Figure 1
Figure 9(p. 16)figure Figure 9
Figure 8(p. 15)figure Figure 8
Figure 4(p. 7)figure Figure 4
Figure 7(p. 11)figure Figure 7
Figure 10(p. 17)figure Figure 10

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Related Subjects

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS
Bison
Heat of transport
Hydrogen distribution
Hydrogen migration
Materials Science
Nuclear Science & Technology
Zircaloy