Release of fission tritium through Zircaloy-4 fuel cladding tubes

Andrieu, Christophe; Ravel, S.; Ducros, G.; Lemaignan, C.

doi:10.1016/j.jnucmat.2005.06.008

Cited by 14 publications

(10 citation statements)

References 27 publications

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“…The collective behavior of interstitial oxygen atoms in hcp-Zr was recently described by Ruban et al [73]. Their calculations indicated that although interstitial oxygen atom produces just a minimal shift of the surrounding Zr atoms, it is energetically unfavorable (by about 0.5 eV) when two O atoms occupy any neighboring interstitial positions.…”

Section: Oxidationmentioning

confidence: 81%

“…Their calculations indicated that although interstitial oxygen atom produces just a minimal shift of the surrounding Zr atoms, it is energetically unfavorable (by about 0.5 eV) when two O atoms occupy any neighboring interstitial positions. The authors of [73] later used these atomic-scale energies in their Monte Carlo simulations of thermodynamically preferred interstitial configuration(s). We repeated these calculations and confirmed their validity.…”

Section: Oxidationmentioning

confidence: 99%

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Modeling of Some Physical Properties of Zirconium Alloys for Nuclear Applications in Support of UFD Campaign

Glazoff¹

2013

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Zirconium-based alloys Zircaloy-2 and Zircaloy-4 are widely used in the nuclear industry as cladding materials for light water reactor (LWR) fuels. These materials display a very good combination of properties such as low neutron absorption, creep behavior, stress-corrosion cracking resistance, reduced hydrogen uptake, corrosion and/or oxidation, especially in the case of Zircaloy-4. However, over the last couple of years, in the post-Fukushima Daiichi world, energetic efforts have been undertaken to improve fuel clad oxidation resistance during off-normal temperature excursions. Efforts have also been made to improve upon the already achieved levels of mechanical behavior and reduce hydrogen uptake. In order to facilitate the development of such novel materials, it is very important to achieve not only engineering control, but also a scientific understanding of the underlying material degradation mechanisms, both in working conditions and in storage of used nuclear fuel.This report strives to contribute to these efforts by constructing the thermodynamic models of both alloys; constructing of the respective phase diagrams, and oxidation mechanisms. A special emphasis was placed upon the role of zirconium suboxides in hydrogen uptake reduction and the atomic mechanisms of oxidation. To that end, computational thermodynamics calculations were conducted concurrently with first-principles atomistic modeling.

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Section: Oxidationmentioning

confidence: 81%

Section: Oxidationmentioning

confidence: 99%

Modeling of Some Physical Properties of Zirconium Alloys for Nuclear Applications in Support of UFD Campaign

Glazoff¹

2013

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“…Zirconium alloys oxidize and the surface ZrO 2 may be an effective permeation barrier, preventing both hydrogen release and formation of detrimental hydrides. Andrieu, et al [137] demonstrated that the rate of tritium release of zircaloy-4 (Zry4) decreased substantially upon oxide formation in tritiated water.…”

Section: Zirconium Alloysmentioning

confidence: 99%

“…The diffusivities reported in studies they cite and in others is plotted in Figure 17. [137,141,145]. Austin,, et al [141] were able to measure the diffusivity in both alpha and beta phases by measuring the activity, due to tritium, in tomographic slices of samples.…”

Section: < Figure 17 Here>mentioning

confidence: 99%

Tritium Barriers and Tritium Diffusion in Fusion Reactors

Causey

Karnesky

Marchi

2012

Comprehensive Nuclear Materials

122

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Tritium and deuterium induce radiation in reactor materials and some radioactive tritium gas may be released. Most of the materials used in fusion reactors are metals that have relatively high permeabilities for tritium. The fusion community has been working on barriers to minimize tritium release. Unfortunately, most barrier materials work very well during laboratory experiments, but fail to meet requirements when placed in radiation environments.This chapter presents tritium permeation characteristics of various materials used in fusion reactors, including plasma-facing, structural, and barrier materials. The necessary parameters for tritium release calculations for various regions of a fusion reactor are given.

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“…However, more current (and mostly non-U.S.) literature indicates, through experimentation and modeling, that the release rate from Zircaloy clad fuel is actually ~1·10 -4 to 1·10 -2 %, making fuel a negligible tritium source and leaving hundreds of curies in the RCS unaccounted for (Liu et al 2003;UK-EPR 2006;Andrieu et al 2005). Documentation submitted to the British regulatory agency states that the unaccounted for tritium source in the coolant could be attributed to secondary source rods (UK-EPR 2006).…”

Section: History Of Accounting For Tritium Sources In the Rcsmentioning

confidence: 99%

Secondary Startup Neutron Sources as a Source of Tritium in a Pressurized Water Reactor (PWR) Reactor Coolant System (RCS)

Shaver¹,

Lanning²

2010

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SummaryCurrently there are large uncertainties in the attribution of tritium in a Pressurized Water Reactor (PWR) Reactor Coolant System (RCS). The measured amount of tritium in the coolant cannot be separated out empirically into its individual sources. Therefore, to quantify individual contributors, all sources of tritium in the RCS of a PWR must be understood theoretically and verified by the sum of the individual components equaling the measured values. The traditional accounting for tritium in the RCS assumes there are two main contributors. The first is soluble species in the RCS, and the second is release from fuel. This stems from the historical knowledge that when fuel had stainless steel cladding it was a major contributor. When the fuel cladding was changed to zirconium alloys in the 1970s, it was assumed that fuel was still the remainder of the source not attributable to the soluble species. Current investigation indicates that this may not be the case. The hypothesis of this paper is that the Zircaloy clad fuel source is minimal and that secondary startup neutron sources are the significant contributors of the tritium in the RCS that was previously assigned to release from fuel.Secondary startup neutrons sources are made of cold-pressed antimony-beryllium pellets in stainless steel cladding. Beryllium is a source of tritium when irradiated with neutrons. Because the secondary source pellets have high porosity and stainless steel has a high permeability for tritium, the tritium that is made in the secondary sources permeates to the coolant. This secondary source contribution of tritium in the coolant is calculated computationally and statistically shown to account for all tritium in the coolant previously assigned to the fuel. Nine reactor cycles from three reactors were analyzed, and the results support the hypothesis of the paper.v

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Release of fission tritium through Zircaloy-4 fuel cladding tubes

Cited by 14 publications

References 27 publications

Modeling of Some Physical Properties of Zirconium Alloys for Nuclear Applications in Support of UFD Campaign

Modeling of Some Physical Properties of Zirconium Alloys for Nuclear Applications in Support of UFD Campaign

Tritium Barriers and Tritium Diffusion in Fusion Reactors

Secondary Startup Neutron Sources as a Source of Tritium in a Pressurized Water Reactor (PWR) Reactor Coolant System (RCS)

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