The corrosion behavior of high-level waste container materials Ti and Ti-Pd alloy under long-term gamma irradiation in Beishan groundwater

AbstractLocalized corrosion in aqueous environments forms an important bounding condition for the performance assessment of high-level waste (HLW) container materials. A predictive methodology using repassivation potential is examined in this paper. It is shown, based on long-term (continuing for over 11 months) testing of alloy 825, that repassivation potential of deep pits or crevices is a conservative and robust parameter for the prediction of localized corrosion. In contrast, initiation potentials measured by short-term tests are non-conservative and highly sensitive to several surface and environmental factors. Corrosion data from various field tests and plant equipment performance are analyzed in terms of the applicability of repassivation potential. The applicability of repassivation potential for predicting the occurrence of stress corrosion cracking (SCC) and intergranular corrosion in chloride containing environments is also examined.

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Effect of Fe2O3 on the Immobilization of High-Level Waste with Magnesium Potassium Phosphate Ceramic

Science and Technology of Nuclear Installations ◽

10.1155/2019/4936379 ◽

2019 ◽

Vol 2019 ◽

pp. 1-10

Author(s):

Hailin Yang ◽

Mingjiao Fu ◽

Bobo Wu ◽

Ying Zhang ◽

Ruhua Ma ◽

...

Keyword(s):

Sintering Temperature ◽

Potassium Phosphate ◽

High Level Waste ◽

Radioactive Nuclide ◽

Hydration Reaction ◽

Sintering Process ◽

Positive Role ◽

Ceramic Structure ◽

High Level

For the proposed novel procedure of immobilizing HLW with magnesium potassium phosphate cement (MKPC), Fe2O3 was added as a modifying agent to verify its effect on the solidification form and the immobilization of the radioactive nuclide. The results show that Fe2O3 is inert during the hydration reaction. It slows down the hydration reaction and lowers the heat release rate of the MKPC system, leading to a 3°C-5°C drop in the mixture temperature during hydration. Early comprehensive strength of Fe2O3 containing samples decreased slightly while the long-term strength remained unchanged. For the sintering process, Fe2O3 played a positive role, lowering the melting point and aiding the formation of ceramic structure. CsFe(PO4)2, or CsFePO4, was generated by sintering at 900°C. These products together with the ceramic structure and absorption benefit the immobilization of Cs+. The optimal sintering temperature for heat treatment is 900°C; it makes the solidification form a fired ceramic-like structure.

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Flow Barrier System for Long-Term High-Level-Waste Isolation: Experimental Results

Nuclear Technology ◽

10.13182/nt98-a2911 ◽

1998 ◽

Vol 124 (1) ◽

pp. 88-100 ◽

Cited By ~ 12

Author(s):

James L. Conca ◽

Michael J. Apted ◽

Wei Zhou ◽

Randolph C. Arthur ◽

John H. Kessler

Keyword(s):

Experimental Results ◽

High Level Waste ◽

High Level

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Corrosion behavior of Alloy 690 and Alloy 693 in simulated nuclear high level waste medium

Journal of Nuclear Materials ◽

10.1016/j.jnucmat.2011.07.005 ◽

2011 ◽

Vol 418 (1-3) ◽

pp. 27-37 ◽

Cited By ~ 11

Author(s):

Pradeep Kumar Samantaroy ◽

Girija Suresh ◽

Ranita Paul ◽

U. Kamachi Mudali ◽

Baldev Raj

Keyword(s):

Corrosion Behavior ◽

High Level Waste ◽

Alloy 690 ◽

High Level

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Dose-Dependence of Pb-Ion Implantation Damage in Zirconolite, Hollandite, and Zircon

MRS Proceedings ◽

10.1557/proc-11-379 ◽

1981 ◽

Vol 11 ◽

Cited By ~ 2

Author(s):

T. J. Headley ◽

G. W. Arnold ◽

C. J. M. Northrup

Keyword(s):

Radiation Damage ◽

Structural Damage ◽

Accelerated Aging ◽

Dose Dependence ◽

High Level Waste ◽

Long Term Stability ◽

Waste Forms ◽

Recent Approach ◽

High Level

The long-term stability of nuclear waste forms is an important consideration in their selection for safe disposal of radioactive waste. Stability against long-term radiation damage is particularly difficult to assess by short-term laboratory experiments. Much of the displacement damage in high-level waste forms will be generated by heavy recoil nuclei emitted during the α-decay process of long-lived actinide elements. Hence, an accelerated aging test which reliably simulates the α-recoil damage accumulated during thousands of years of storage is desirable. One recent approach to this simulation is to implant the waste form with heavy Pb-ions.I- 6 If the validity of this approach is to be fully assessed, two important questions which have not yet been investigated must be answered.(1) Is the structural damage, including cumulative effects, similar for irradiation by Pb-ions and a-recoil nuclei in a given material? (2) Is the dose-dependence of the accumulated damage similar? The purpose of this investigation was to assess the extent of these similarities in selected materials. We utilized transmission electron microscopy (TEM) to characterize the radiation damage and measure its dose-dependence.

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An Approach to Analysis of High Level waste Container Performance in an Unsaturated Repository Environment

MRS Proceedings ◽

10.1557/proc-333-863 ◽

1993 ◽

Vol 333 ◽

Author(s):

John C. Walton ◽

Narasi Sridhar ◽

Gustavo Cragnolino ◽

Tony Torng ◽

Prasad Nair

Keyword(s):

Dominant Role ◽

Near Field ◽

Temperature Fields ◽

Localized Corrosion ◽

High Level Waste ◽

Quantitative Methodology ◽

Waste Container ◽

Wide Range ◽

Engineered Barrier ◽

High Level

ABSTRACTOne of the requirements for the performance of waste packages prescribed in 10CFR 60.113 is that the high level waste must be “substantially completely” contained for a minimum period of 300 to 1000 years. During this period, the radiation and thermal conditions in the engineered barrier system and the near-field environment are dominated by fission product decay. In the present U.S design of the engineered barrier system, the outer container plays a dominant role in maintaining radionuclide containment. A quantitative methodology for analyzing the performance of the container is described in this paper. This methodology enables prediction of the evolution of the waste package environment in terms of temperature fields, stability of liquid water on the container surface, and concentration of aggressive ions such as chloride. The initiation and propagation of localized corrosion is determined by the corrosion potential of the container material and critical potentials for localized corrosion. The coiTOsion potential is estimated from the kinetics of the anodic and cathodic reactions including oxygen diffusion through scale layers formed on the container surface. The methodology described is applicable to a wide range of metals, alloys and environmental conditions.

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