Abstract: An improved radiation shielding material and storage systems for radioactive materials incorporating the same. The PYRolytic Uranium Compound (“PYRUC”) shielding material is preferably formed by heat and/or pressure treatment of a precursor material comprising microspheres of a uranium compound, such as uranium dioxide or uranium carbide, and a suitable binder. The PYRUC shielding material provides improved radiation shielding, thermal characteristic, cost and ease of use in comparison with other shielding materials. The shielding material can be used to form containment systems, container vessels, shielding structures, and containment storage areas, all of which can be used to house radioactive waste. The preferred shielding system is in the form of a container for storage, transportation, and disposal of radioactive waste. In addition, improved methods for preparing uranium dioxide and uranium carbide microspheres for use in the radiation shielding materials are also provided.

Abstract: An improved radiation shielding material and storage systems for radioactive materials incorporating the same. The PYRolytic Uranium Compound (“PYRUC”) shielding material is preferably formed by heat and/or pressure treatment of a precursor material comprising microspheres of a uranium compound, such as uranium dioxide or uranium carbide, and a suitable binder. The PYRUC shielding material provides improved radiation shielding, thermal characteristic, cost and ease of use in comparison with other shielding materials. The shielding material can be used to form containment systems, container vessels, shielding structures, and containment storage areas, all of which can be used to house radioactive waste. The preferred shielding system is in the form of a container for storage, transportation, and disposal of radioactive waste. In addition, improved methods for preparing uranium dioxide and uranium carbide microspheres for use in the radiation shielding materials are also provided.

Abstract: An improved radiation shielding material and storage systems for radioactive materials incorporating the same. The PYRolytic Uranium Compound (“PYRUC”) shielding material is preferably formed by heat and/or pressure treatment of a precursor material comprising microspheres of a uranium compound, such as uranium dioxide or uranium carbide, and a suitable binder. The PYRUC shielding material provides improved radiation shielding, thermal characteristic, cost and ease of use in comparison with other shielding materials. The shielding material can be used to form containment systems, container vessels, shielding structures, and containment storage areas, all of which can be used to house radioactive waste. The preferred shielding system is in the form of a container for storage, transportation, and disposal of radioactive waste. In addition, improved methods for preparing uranium dioxide and uranium carbide microspheres for use in the radiation shielding materials are also provided.

Abstract: Metal surfaces having an oxide coating containing radioactive substances, such as the primary system of a pressurized water reactor, are decontaminated by passage thereover of a decontamination solution containing a weak chelating agent, such as nitrilotriacetic acid, and a ferrous salt, such as ferrous glutonate. The weak chelating agent is present in an aqueous solution in an amount of 0.1 to 2.0 percent by weight and the ferrous salt in an amount to provide 50 to 500 parts per million iron based on the weight of the solution. The solution, after contact with the metal surfaces is regenerated by an ion exchange resin or, preferably, by electrolysis.

Abstract: Disclosed is a method of decontaminating the metal surfaces in the cooling system of a nuclear reactor by contacting the metal surfaces with an aqueous solution containing about 0.5 to about 3% of a ceric acid which can be tetrasulfato ceric acid, hexasulfamato ceric acid, hexaperchlorato ceric acid, or mixtures thereof, and about 1 to about 5% of an inorganic acid that forms a complex with the ceric acid.The cerium III in the aqueous solution can be oxidized to cerium IV to increase the life and effectiveness of the solution. After oxidation, the aqueous solution can be passed through a hydrogen form cation exchange column to remove metal ions. If the aqueous solution contains uranyl or plutonyl ions these can be recovered by extraction for use in making fuel.Also disclosed is a decontaminating solution of water containing about 0.

Abstract: Both a method and apparatus for electrolytically removing radioactive metal ions from a decontamination solution to regenerate the solution and prepare the ions for disposal as disclosed herein. At least the cathodic portion of the electrode used in the electrolysis is formed from a combustible material, such as a semi-fluidized bed of graphite particles. In the method of the invention, the decontamination solution is passed in intimate contact with the graphite particles forming the cathodic portion of the electrode as an electric potential is applied to the electrode. As a result of the electric potential, the metal ions are detached from the chelate in the decontamination solution and deposited onto the graphite particles of the cathodic portion of the electrode. After the electrode becomes spent, it is incinerated in order to reduce the volume of the resulting radioactive ash. The gases produced from the incineration are scrubbed with a scrubbing liquid to remove radioactive particles therefrom.

Abstract: Disclosed is a method of decontaminating metal surfaces contaminated with a radioactive deposit by passing over the coating an aqueous solution of a water soluble condensation reaction product of (1) a hydrazine compound having the general formula ##STR1## where each R group is independently selected from hydrogen and alkyl to C.sub.4 and (2) a water soluble aliphatic polycarboxylic acid. The method includes the additional last steps of passing an aqueous oxidizing solution over the deposit followed by passing a decontamination solution over the deposit a second time. The solution is circulated through a cationic exchange column which can be preloaded with hydrazine ion. The decontamination solution can be decomposed by the addition of an oxidant to form nitrogen, carbon dioxide, and water.

Abstract: Disclosed is an improved process for oxidizing chromium in deposits in the cooling system of a nuclear reactor using a solution of ozone. The improvement consists of adding to the solution about 0.01 to about 0.5% of a water-soluble cerium IV compound, about 0.1 to about 0.5% of a water-soluble aromatic compound having at least one ketone group on an aromatic ring, or adding both. Also disclosed is the composition of water, ozone, the cerium IV compound and the aromatic compound. Also disclosed is a process for decontaminating the cooling system of nuclear reactors by adding a decontamination composition to the coolant, circulating the coolant between the cooling system and a cation exchange resin, removing the decontamination composition by passing it through an anion exchange resin, adjusting the temperature to 40.degree. to about 100.degree. C., adding the ozone oxidation composition, circulating the coolant through the cooling system, raising the temperature to at least about 100.degree. C.

Abstract: Disclosed is a method of decontaminating the metal surfaces in the cooling system of a nuclear reactor by contacting the metal surfaces with an aqueous solution containing about 0.5 to about 3% of a ceric acid which can be tetrasulfato ceric acid, hexasulfamato ceric acid, hexaperchlorato ceric acid, or mixtures thereof, and about 1 to about 5% of an inorganic acid that forms a complex with the ceric acid.The cerium III in the aqueous solution can be oxidized to cerium IV to increase the life and effectiveness of the solution. After oxidation, the aqueous solution can be passed through a hydrogen form cation exchange column to remove metal ions. If the aqueous solution contains uranyl or plutonyl ions these can be recovered by extraction for use in making fuel.Also disclosed is a decontaminating solution of water containing about 0.

Abstract: Disclosed is an oxidizing composition of water, about 0.1 to saturation of an alkali metal hypohalite, and sufficient alkali metal hydroxide to raise the pH of the solution to at least about 12. A method of decontaminating metal surfaces having a coating thereon which contains radioactive substances is also disclosed. The composition is passed over the coating at a temperature of about 50.degree. to about 120.degree. C. followed by passing a decontamination solution over the coating.

Abstract: A method for decontaminating metal surfaces having radioactive oxide deposits thereon is disclosed which comprises preparing a decontamination solution having a pH of about 1.5 to about 4 comprising water, about 0.02 to about 0.5% of a water soluble organic acid which has an equilibrium constant in a complex with ferric ion of at least about 10.sup.9 and which is capable of producing a pH of about 2 to about 3 in water, and about 0.01 to about 0.4% of a chelate in free acid form which has an equilibrium constant in a complex with ferric ion of about 10.sup.15 to about 10.sup.19 and which is at least 0.4% soluble at 40.degree. C. in water having a pH of about 2 to about 3. An apparatus for performing the above method is disclosed.

Abstract: Disclosed is a method of decontaminating metal surfaces having a radioactive coating thereon where the coating contains metal ions. An aqueous decontamination solution containing at least one chelate is passed over the coating to solubilize the metals. The decontamination solution is then passed through a porous direct current electrode to reduce the metal ions in the solution and the decontamination solution is again passed over the coating. Also disclosed is an apparatus for decontaiminating metal surfaces coated with radioactive substances. The apparatus consists of a container means for holding an aqueous decontamination solution where the solution contains a chelate, a porous cathode through which the aqueous decontamination solution is passed, and means for circulating the aqueous decontamination solution between the container means, the porous cathode, and the metal surfaces.

Abstract: Slab-shaped high efficiency catalytic reformer configurations particularly useful for generation of fuels to be used in fuel cell based generation systems. A plurality of structures forming a generally rectangular peripheral envelope are spaced about one another to form annular regions, an interior annular region containing a catalytic bed and being regeneratively heated on one side by a hot combustion gas and on the other side by the gaseous products of the reformation. An integrally mounted combustor is cooled by impingement of incoming oxidant.

Abstract: Slab-shaped high efficiency catalytic reformer configurations particularly useful for generation of fuels to be used in fuel cell based generation systems. A plurality of structures forming a generally rectangular peripheral envelope are spaced about one another to form annular regions, an interior annular region containing a catalytic bed and being regeneratively heated on one side by a hot comubstion gas and on the other side by the gaseous products of the reformation. An integrally mounted combustor is cooled by impingement of incoming oxidant.