Zircon Promises to be a Host Phase for the Immobilization of Excess Weapon Plutonium

One of the new and daunting challenges in nuclear waste management is the disposition of plutonium recovered from dismantled nuclear weapons. Under the first and second Strategic Arms Reduction Treaties, as well as unilateral pledges made by the United States and Russia, several thousand nuclear weapons will be dismantled. Dismantlement will result in an estimated 100 metric tons of excess weapon plutonium that will require long-term disposition. The disposal strategy should not only protect the public and the environment, but should also ensure that, in the interests of nonproliferation and denying plutonium to terrorists, the plutonium is not readily recoverable.

Figure 1. A ball-and-stick representation of the atomic structure of zircon. Ruled tetrahedra are SiO4 groups, and shaded polyhedra are ZrO8 groups.

Two recent reports by the National Academy of Sciences presented "promising alternatives": 1) partial consumption of Pu in a mixed oxide fuel in existing or modified reactors and final disposal as spent nuclear fuel without reprocessing; 2) vitrification as a borosilicate glass mixed with radioactive high-level waste and disposal as "glass logs." In both cases the waste will be so highly radioactive that it cannot be handled by terrorists or potential proliferators. A third, less studied, option involves deep burial (4 to 6 kilometers).

A key consideration for any disposition strategy is the form of the immobilized waste. The deep burial option will require a waste form durable at elevated temperatures (>150°C) that result from the normal geothermal gradient and with a waste loading sufficiently high so as to reduce the volume of material. Additionally, the higher the 239Pu content, the more important are concerns for criticality. A highly durable waste form can retain not only the Pu but also other nuclides to serve as neutron "poisons."

With such considerations in mind, researchers at the University of New Mexico (UNM) are initiating a collaboration with the Nuclear Materials Technology Division (NMT) of Los Alamos National Laboratory to investigate zircon (ZrSiO4), a highly durable, naturally occurring mineral, as a host for the immobilization of Pu. This project is also part of a collaboration with Battelle Pacific Northwest Laboratory (PNL) and the Khlopin Radium Institute in St. Petersburg, Russia.

Zircon occurs in nature with uranium and thorium in its structure in concentrations up to 5,000 ppm. It is an extremely durable accessory mineral in igneous and metamorphic rocks and is often found as a heavy mineral in stream sediments. After transport over great distances, it shows limited chemical alteration or physical abrasion. Some of the oldest zircons occur in sedimentary rocks that have been through numerous cycles of erosion, transport, and deposition, all the while retaining the original U/Pb systematics of the rock in which the zircon crystallized. Because of zircon's importance in geochronological and geochemical studies of the earth's crust, there are already numerous studies that have analyzed the variations in the U/Pb systematics as a result of radiation damage, thermal events, and alteration. Most recently, zircon with high actinide concentrations (10 wt % UO2) have been identified in the Chernobyl "lavas," molten materials resulting from the Chernobyl reactor meltdown accident.

The zircon structure is well known, and compositions of ASiO4, for which A4+ = Zr, Hf, Th, Pa, U, Np, Pu and Am, were synthesized in the early 1960s. Four of these compositions, hafnon (HfSiO4), zircon, coffinite (USiO4) and thorite (ThSiO4), occur naturally. There is complete miscibility between ZrSiO4 and HfSiO4. Zircon with 10 wt % Pu has already been synthesized. The fact that a pure, end-member composition, PuSiO4, has been synthesized suggests that extensive substitution of Pu for Zr is possible.

Because of the large number of previous studies of zircon, a number of the issues that are inevitably raised in the evaluation of a nuclear waste form can already be addressed. Radiation damage studies of zircon began in the early part of this century. Zircon undergoes a radiation-induced transformation from the periodic to the aperiodic state at doses over the range of 1018 to 19 a-decay events/g ( = 0.2 to 0.6 displacements per atom) with a density decrease and a corresponding volume expansion. Previous studies provide a firm basis for predicting the microstructure of the radiation-damaged zircon as a function of dose and temperature. Based on these data, for a waste loading of 10 wt % of 239Pu, the zircon will reach the saturation value of damage (1.2 x 1019 a-decay events/g or 0.8 dpa) in approximately 1,700 years.

Alteration of natural zircons has import-ant implications for the use of U/Th/Pb techniques in geochronology; thus, there is an extensive literature that describes alteration. These studies show that even under extreme geologic conditions, the alteration is usually minor.

Leaching studies of natural zircons in laboratory experiments have confirmed the loss of U, Th and Pb under hydrothermal conditions. However, at lower temperatures (< 80°C) and near neutral pH values, i.e., conditions more pertinent to nuclear waste disposal, zircon is extremely insoluble.

There are much less data than necessary for a full evaluation of zircon as a waste form, which will be the focus of the UNM-LANL collaboration. Crystalline zircon is stable to such an extent that the equilibrium concen-trations of Zr and Si are in the order of 10-9 moles/L (0.1 ppb) at 25°C. The dissolution rate of amorphous zircon is still considerably lower than that of glass in stagnant, silica-saturated solutions. In an open system (e.g., moving ground water), the leach rate for zircon does not increase; however, the leach rate of borosilicate glass may increase by three orders of magnitude. Thus, one of the main advantages of zircon may be its high durability in an open system in which ground waters are present, as this attribute allows considerably greater flexibility in repository design.

Criticality is a concern for any Pu-waste form during both processing and final waste disposal. The possibility of criticality can be mitigated by adjusting the waste loading in the zircon or by including neutron "poisons," such as hafnium and gadolinium, in the formula-tion. Natural zircons contain up to several thousand parts per million rare earths, including up to 5,000 ppm of Gd, and hafnium is a bothersome, major impurity in natural zircons. Thus, neutron-absorbing nuclides may be incorporated into the zircon structure as dilute solid solutions.

Processing and production technologies are always an important consideration in the adoption of a waste form. As already discussed, Pu-doped zircons have been synthesized with 10 wt % Pu. Zircon can be synthesized by sintering (1200°C to 1300°C), and yields of nearly 90% are obtained in less than one day. Although larger-scale pilot demonstrations are required, much is already known about the requirements for synthesis. In the case of weapon plutonium, one may use to advantage the essentially reagent grade of the Pu in engineering the process and the waste form. The final processing technology can draw directly on the extensive experience gained in producing mixed-oxide fuel from sintered powders.

This article was contributed by Regents Professor Rodney C. Ewing, who is in the Department of Earth and Planetary Sciences at UNM. Collaborators on this research effort are Ewing and W. Lutze at UNM, W. J. Weber at PNL, personnel in the NMT Division at Los Alamos, and E. Alexandrov and B. Burakov at the Khlopin Radium Institute, Russia.

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