²³⁸Pu Processing Line Will Provide New NMT Capability

A ²³⁸Pu aqueous scrap recovery glove box line is being built at the TA-55 Plutonium Facility with an annual throughput capacity of several kilograms ²³⁸Pu. This new capability within NMT Division is anticipated to be in operation by fiscal year 2001 and further supports NMT's role as a lead DOE facility for plutonium processing.

The aqueous line is designed to purify ²³⁸Pu-oxide (²³⁸PuO₂) fuel, used in the fabrication of general-purpose heat sources (GPHS) or light-weight radioisotope heater units (LWRHUs). The heat sources supply the thermal energy used in thermoelectric generators to power spacecraft for deep space missions and to heat critical components in the cold environs of space. The Power Source Technologies Group, NMT-9, has manufactured LWRHUs for use in the NASA space program for approximately 20 years (see Fig. 1). More recently, Los Alamos manufactured the GPHSs to power the spacecraft in the Cassini mission to Saturn (See Actinide Research Quarterly, Fall 1996 and Summer 1997).

Figure 1. The 238Pu aqueous recovery glove box line is being built at the TA-55 Plutonium Facility. It is designed to purify 238PuO2 used in the fabrication of heat sources that supply thermal energy to power spacecraft for deep space missions and to heat critical components in the cold environs of space. Three such heat sources were used on this Mars Pathfinder Rover to keep electronic equipment within normal operating temperatures.

The purification of ²³⁸PuO₂ is necessary because of unacceptable levels of ²³⁴U and other impurities in scrap fuel. Impurities at levels above GPHS and LWRHU specifications may impair the performance of the heat sources. The purification involves a nitric acid/hydrofluoric acid reflux of the oxide powder, followed by oxalate precipitation and filtration. In cases where the ²³⁸Pu material contains gross levels of impurities, it is necessary to treat it through the nitrate anion exchange process before the oxalate precipitation step. Plutonium-238 recovered from other material and various waste forms will eventually be processed through the aqueous line. With the expected high levels of impurities in this material, nitrate anion exchange becomes an important unit operation in the ²³⁸Pu aqueous scrap recovery line. The bench-scale experimental efforts in performing nitrate anion exchange for ²³⁸Pu purification are summarized in this article.

Previous research at Los Alamos in collaboration with Reilley Industries focused on producing an anion exchange resin with increased safety and high-loading-capacity characteristics. This work led to the formulation of the polyvinylpyridine-based Reillex-HPQ resin. The studies showed the resin to be resistant to radiolytic and thermal degradation and to display comparable or superior sorption kinetics in comparison to several other polystyrene-based resins used for actinide purification. Based on these studies, the Reillex-HPQ anion exchange resin was chosen for ²³⁸Pu aqueous processing.

Bench-scale experiments are being conducted to demonstrate that high levels of impurities are separated from ²³⁸Pu solutions using Reillex-HPQ resin, and to determine if chemical pretreatment is necessary to maintain the ²³⁸Pu in the (IV) oxidation state. The results of the bench-scale experiments also determine the baseline operation method to be used for the full-scale process. Other work in collaboration with Westinghouse Savannah River Technology Center (WSRTC) involves heat transfer calculations to determine the thermal gradients expected during ion exchange processing.

It is the Pu(IV)-hexanitrato complex in 7 molar (M) nitric acid that sorbs onto the Reillex-HPQ resin during ion exchange. Maintaining all of the Pu in the (IV) tetravalent state is difficult because of the high oxidizing environment that develops in 7 M nitric acid containing 238Pu. Most of the plutonium in 7 M nitric acid is expected to exist in the Pu(IV) tetravalent state. However, the high alpha activity of 17 Ci/gm ²³⁸Pu results in an increased oxidizing environment (more oxidizing radiolysis products), which results in the formation of Pu(VI). The hexavalent state weakly sorbs to the resin leading to Pu breakthrough in the effluent waste solutions and low ion exchange efficiencies.

Our bench-scale studies show that chemical pretreatment is important for maintaining ²³⁸Pu in the (IV) valence state. Greater than 99% efficiencies using 3 to 5 grams of ²³⁸Pu in nitric acid solutions have been achieved with chemical pretreatment. Chemical pretreatment was accomplished using urea, ferrous ammonium sulfate, and sodium nitrite. Similar results were obtained using ferrous sulfamate instead of ferrous ammonium sulfate. Without chemical pretreatment a large percentage of ²³⁸Pu is lost to the effluent and wash streams, most likely as the Pu(VI) species. The results of experiments in which the ²³⁸Pu solution had been spiked with high levels of GPHS impurities showed that decontamination factors as high as 103 are achievable.

The heat transfer calculations determined the maximum bed temperature during loading, washing, or elution for normal column (continual flow) operation and the equilibrium temperature for no flow through the column. The constants in the calculations were 75 grams of ²³⁸Pu loaded onto 1.6 liters of resin in a 3-inch-diameter Pyrex column. The parameters varied were ²³⁸Pu concentration, flow rate, and temperature of the Pyrex column outside wall. The results of the heat transfer calculations indicate that under full-scale operating conditions, the maximum resin bed temperature does not exceed 60 degrees C. The operation of ion exchange columns below this temperature is deemed safe in safety review studies of ion exchange columns in nuclear processing. The results of the heat transfer calculations are available, as necessary, for process hazards analysis of the full-scale ion exchange operation.

The heat transfer calculations were based on a previous study at WSRTC in support of the ²³⁸Pu production campaign to provide material to Los Alamos National Laboratory for heat source fabrication. The heat transfer calculations were performed using a computer code that incorporated models for absorption and elution of ²³⁸Pu, and for forced and natural convection within the resin bed.

Other efforts include current work to develop the separation of fractional levels of thorium from ²³⁸Pu solutions utilizing a mixture of, for example, 0.007 M hyrofluoric acid and 0.45 M nitric acid as the eluant. The method of using 0.007 M hydrofluoric acid and 0.45 M nitric acid was developed by F. Marsh at Los Alamos. Some impure ²³⁸PuO₂ fuel sources are expected to have high levels of thorium, which must be decreased to below the GPHS specification of 0.5% during the purification process. In the past this has proven to be difficult because some thorium sorption occurs during ion exchange.

A major effort is also underway to qualify all of the experimental methodology developed during bench-scale work for the full-scale operation. This includes the comminution, dissolution and filtration, ion exchange, and oxalate precipitation processes.

The ²³⁸Pu aqueous scrap recovery line provides a unique capability for the aqueous purification of ²³⁸Pu heat sources as well as aqueous processing of ²³⁸Pu recovered from other material and various waste forms.

This article was contributed by M. E. Pansoy-Hjelvik (NMT-9). Others involved in this work are J. Laurinat(WSRTC) and J. Nixon, J. Brock, G. Silver, M. A. Reimus, and K. B. Ramsey (NMT-9)

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