The WIPP is a transuranic waste disposal project of the U.S. Department of Energy and administered by the Carlsbad Area Office. The WIPP facility is a mined repository located near Carlsbad, New Mexico. It was developed and constructed 658 meters below the surface to demonstrate the safe, permanent isolation of transuranic radioactive wastes in a bedded salt formation. After the site selection phase and construction phase were completed, DOE embarked upon a pre-disposal experimental program phase. There are four major WIPP experimental program areas: shaft seals and rock mechanics; disposal room interactions; fluid flow and transport; and direct releases due to human intrusion.

The Pre-Disposal Phase Experimental Program provided the necessary information to support the WIPP performance assessment process. The results of the program have been instrumental in demonstrating that the WIPP meets the regulatory requirements.

The role of performance assessment was discussed by a group of panelists and the participants of the 20th MRS symposium on the “Scientific Basis for Nuclear Waste Management.” Panel members were Professor Thomas Pigford, Dr. Alan Cooper, and Dr. Patrik Sellin; Dr. Michael J. Apted served as moderator. For discussion purposes, “performance assessment” (PA) was defined as the analysis of the release of radionuclides from a repository system of barriers to the accessible environment.

The Hyrkkölä U-Cu mineralization is located in south-western Finland, near the Palmottu analog site, in crystalline, metamorphic bedrock. The age of the mineralization is estimated to be between 1.8 and 1.7 Ga. The existence of native copper and copper sulfides in open fractures in the near-surface zone allows us to study the native copper corrosion process in conditions analogous to a nuclear fuel waste repository.

From the study of mineral assemblages or paragenesis, it appears that the formation of copper sulfide (djurleite, Cu1.934S) after native copper (Cu°) under anoxic (reducing) conditions is enhanced by the availability of dissolved hydrogen sulfide (HS) in the groundwater circulating in open fractures in the near-surface zone. The minimum concentration of HS in the groundwater is estimated to be of the order of 10-5 M (∼ 10-4 g/1) and the minimum pH value not lower than about 7.8 as indicated by the presence of calcite crystals in the same fracture.

The present study is the first one performed on occurrences of native copper in reducing, neutral to slightly alkaline groundwaters. Thus, the data obtained is of most relevance in improving models of anoxic corrosion of copper canisters.

Perched water zones have been identified in the fractured, welded tuff in the semi-arid to arid environments of Yucca Mountain, Nevada and near Superior, Arizona. An understanding of the formation of such zones is necessary in order to predict where future perched water might form at Yucca Mountain, the proposed site of a high-level nuclear waste repository. The formation or growth of a perched zone near a repository is one of the factors to be considered in the risk assessment of the Yucca Mountain site.

The Apache Leap Research Site near Superior, Arizona is a natural analog to the Yucca Mountain site in terms of geology, hydrology, and climate. Information used to study possible mechanisms for the formation of the perched zone included data regarding isotopie and geochemical properties of the waters in and above the perched water zone; measured hydrologie parameters of the perched zone; geophysical and measured parameters of the tuff; megascopic and microscopic observations of the tuff, including mineralogical, alteration, and structural features; and the lateral and vertical extent of perched water in the region.

Aquifer test, geophysical, geochemical, and radioisotopic data show that fractures are the means by which water is recharging the perched zone. The reduced hydraulic conductivity of the formation in the perched zone appears to result from both a severe reduction in matrix porosity and permeability caused by welding, devitrification, and vapor phase crystallization; and by an increase in fracture filling which restricts the pathways for flow.

One of the major problems in testing radionuclide transport models with a natural analogue is the potential discrepancy between experimental and modelling concepts. In the matrix diffusion studies at Palmottu the measured U-series concentration profiles and mathematical simulations showed disagreement, suggesting discrepancy in respective concepts of radionuclide attachment on the rock pores. Here, the discrepancy was approached by studying uranium fixation in more detail within the most loosely-bound fraction.

The accuracy of uranium-series dating was studied quantitatively using the natural long-lived decay chains 4n+2 and 4n+3. An unusually well-bounded matrix diffusion case was used as the reference for USD simulations. The simulations applied the traditional closed and open system models, as well as detailed multistage models that incorporated the known accumulation history of the decay chain members. The results show clearly the improved accuracy and consistency in dating with detailed mass flow information.

Gibbs free energies of formation (ΔG°ƒ) for several structurally related U(VI) minerals are estimated by summing the Gibbs energy contributions from component oxides. The estimated ΔG°f values are used to construct activity-activity (stability) diagrams, and the predicted stability fields are compared with observed mineral occurrences and reaction pathways. With some exceptions, natural occurrences agree well with the mineral stability fields estimated for the systems Sio2-Cao-Uo3-UOH2O and Co2-caO-UO3-H2O providing confidence in the estimated thermodynamic values. Activity-activity diagrams are sensitive to small differences in ΔG°f values, and mineral compositions must be known accurately, including structurally bound H2O. The estimated ΔG°f values are not considered reliable for a few minerals for two major reasons: (1) the structures of the minerals in question are not closely similar to those used to estimate the ΔG°f* values of the component oxides, and/or (2) the minerals in question are exceptionally fine grained, leading to large surface energies that increase the effective mineral solubilities.

Ianthinite, [U4+2(UO2)4O6(OH)4(H2O)4](H2O)5, is the only known uranyl oxide hydrate mineral that contains U4+, and it has been proposed that ianthinite may be an important Pu4+ -bearing phase during the oxidative dissolution of spent nuclear fuel. The crystal structure of ianthinite, orthorhombic, a 7.178(2), b 11.473(2), c. 30.39(1) Å, V 2502.7 Å3, Z = 4, space group P21cn, has been solved by direct methods and refined by least-squares methods to an R index of 9.7 % and a wR index of 12.6 % using 888 unique observed [ | F | ≥ 5σ | F | ] reflections. The structure contains both U6+ and U4+. The U6+ cations are present as roughly linear (U6+O2)2+ uranyl ions (Ur) that are in turn coordinated by five O2-and OH located at the equatorial positions of pentagonal bipyramids. The U4+ cations are coordinated by O2-, OH and H2O in a distorted octahedral arrangement. The Urφ5 and U4+φ6 (φ: O2-, OH, H2O) polyhedra link by sharing edges to form two symmetrically distinct sheets at z z ≈ 0.0 and z ≈ 0.25 that are parallel to (001). The sheets have the β-U3O8 sheet anion-topology. There are five symmetrically distinct H2O groups located at z ≈ 0.125 between the sheets of Uφn polyhedra, and the sheets of Uφn polyhedra are linked together only by hydrogen bonding to the intersheet H2O groups. The crystal-chemical requirements of U4+ and Pu4+ are very similar, indicating that extensive Pu4+ ↔ U4+ substitution can occur within the sheets of Uφn polyhedra in the structure of ianthinite.

Exceptionally low δ 18O values of primary uraninite and pitchblende (i.e. -32 per mil to -19.5 per mil) from Proterozoic unconformity-type uranium deposits in Saskatchewan, Canada, in conjunction with theoretical uraninite-water oxygen isotope fractionation factors suggest that primary uranium mineralization is not in oxygen isotopie equilibrium with clays and silicates. The low δ 18O values have been interpreted to have resulted from the recrystallization of primary uranium mineralization in the presence of modern meteoric fluids having low δ 18O values of ca. -18 per mil. The absence of apparent alteration in many of the uraninite and pitchblende samples requires that the uranium minerals exchange oxygen isotopes with fluids, with only minor disturbances to their original chemical compositions and textures. However, experiments on the interaction between water and natural uraninites, from these deposits, and detailed electron micro-probe analyses of natural uraninite and pitchblende indicate that, in the presence of water, old uraninite rapidly alters to curite (Pb2U5O174H2O). The hydration of uraninite to curite releases uranium and calcium into solution and becquerelite (Ca(UO2)6O4(OH)6H2O) is precipitated. In the presence of Si-saturated waters, uranium silicate minerals, soddyite ((UO2)2(SiO4)2H2O) and kasolite (Pb(UO2)SiO4H2O are precipitated in addition to, curite and schoepite ((UO2)8O2(OH)12(H2O)12). The mineral paragenesis observed in these experiments is similar to sequences observed in oxidized zones in uranium deposits and UO2-water experiments. Therefore, it is unlikely that natural uraninite and pitchblende can simply exchange oxygen with an oxidizing fluid without a concomitant change in phase chemistry or structure, nor will oxidation of uraninite lead to the formation of U3O7. as predicted by theoretical calculations used in natural analogue studies for the disposal of high level nuclear waste.

Uraninites from the Bangombé natural fission reactor (RZB) and “normal” uranium-ore occur as fine veins in the sandstone host-rock as well as altered, broken, and slightly displaced grains in an illitic matrix, and in nodules and veins of solid bitumen. Inclusions of galena, (Y,Gd)-rich phosphates, a Pb-oxide and a Ti-oxide? were observed. Uraninites just below RZB were partially altered to a uranyl-sulfate. Three generations of uraninite were identified based on their PbO-contents of 8–11.06 wt%, 6 wt% (the largest population), and a younger generation with 3 wt%. The high Pb-uraninites may be the precursor to the low Pb-uraninites. Diffusional loss of Pb is indicated by the presence of a Pb-oxide at the interface to the uraninites. The behaviour of the metallic fission products, incompatible with the uraninite structure, may mimic the behaviour of Pb in these uraninites. The averaged impurity-content ranges from 4.29 to 6.89 wt%, and consists mainly of SiO2, TiO2, ZrO2, FeO, CaO, Al2O3 and P2O5. The averaged content of Y2O3 and the Ln's is less than 0.78 wt% and there is a scattered positive correlation with P2O5. The content of Y + Ln's is generally highest in the uraninites from RZB. Uraninite hydration and the formation of “uranopelite/zippeite” have caused complete loss of Y and the Ln's. These elements seems also to be partially lost by weak phosphatian coffinitization. The analytical results indicate that Y and the Ln's, which are high yield fission products, may be released from uraninite during alteration in the presence of P.

“Gel-zircon”, an unusual Zr-silicate phase from the Manibay uranium mine, northern Kazakhstan, was studied using X-ray diffraction (XRD), electron microprobe energy dispersive X-ray spectroscopy (EDS) and high resolution transmission electron microscopy (HRTEM). XRD results indicate that gel-zircon is mostly amorphous and occurs with numerous impurity phases. Microprobe EDS results indicate a UO2 content up to 9.14 wt. %. HRTEM images revealed that the microtexture of gel-zircon consists of nanocry stal lites of zircon, 2–10 nm in size, in a dominantly amorphous matrix. Despite the U-Pb age of 420±25 my and the lack of significant crystallinity, the gel-zircon is an apparently chemically durable phase. Leaching of uranium ores which contain gel-zircon as the major U-bearing phase is impossible using existing uranium plant technologies. The alpha-decay dose, 2.64 displacements per atom (dpa), corresponding to the age of gel-zircon is much higher than that (0.5 dpa) required to cause metamictization of crystalline zircon. However, the morphology of gel-zircon which occurs as veins up to 5 mm thick and tens of mm long does not indicate initial crystallinity. Initially crystalline natural zircons often preserve their crystal morphology after metamictization. This amorphous phase is analogous to the highly damaged state characteristic of zircon proposed as a waste form for the disposition of excess weapons plutonium.

Two alkali-tin-silicate (ATS) glasses have been prepared at Argonne National Laboratory (ANL) as part of our ongoing research in radioactive waste glass development. These glasses dissolved 5% and approximately 7% Pu. Early corrosion test results indicate that Pu-bearing ATS glass is extremely durable. The initial goal in this project concerned equally both the solubility of Pu and the durability of the ATS glasses; however, our primary emphasis has changed recently to maximizing the loading of Pu in the glass. ATS-based glasses, using Th(VI) and Ce(III) as surrogates for Pu(IV), are now being investigated to increase the solubility of Pu without substantially sacrificing the durability of the current ATS formulations. The solution data from various corrosion tests on the original Pu-containing ATS glasses are also presented.

To examine the proliferation resistance of borosilicate glass, a process to extract and recover a plutonium analog (thorium) from borosilicate glass was developed and examined. The glass matrix examined was a modified standard frit consisting of the ARM-1 frit (with simulated fission products) loaded with 2 wt. % thorium (as an analog for plutonium) and 2 wt. % each of three rare earth elements (Gd, Sm, Eu), which were added for criticality control and to possibly increase the proliferation resistance of the glass matrix. The plutonium analog was extracted from the crushed glass with a nitric acid dissolution process, and subsequently decontaminated using a solvent extraction process. The acid dissolution process was able to extract 88.4 ± 6.8 % of the plutonium surrogate from the glass host form. The bench top solvent extraction process was 30.2 ± 10.9 % efficient in recovering the plutonium analog as a purified product. Overall, this process was able to extract 26.7 ± 9.9 % of the plutonium analog from the glass as a purified product. To quantify the proliferation resistance of borosilicate glass as a host form for weapons-grade plutonium, MCNP was used to determine the compressed critical mass of a plutonium alloy with the same composition as the product of the extraction process. For the average product composition, the compressed critical mass was 4.7 kg of material. On average, one compressed critical mass could be recovered from 613 kg of borosilicate glass (2 wt. % Pu loading).

Vitrification is currently being explored as an option for plutonium (Pu) scrap and residue stabilization because of its ability to convert plutonium-bearing materials into a safe, durable, and accountable waste form with reduced need for safeguarding. To develop this glass, the effect of composition on key glass properties (viscosity and chemical durability) were measured. A one-component-at-a-time change test matrix with 36 glasses was designed. The viscosity is reported as the temperature at 5 Pas (T5), and the durability as the normalized release of boron and sodium (Гb and ГNa) from MCC-1 and PCT static leach tests and the negative logarithm of fractional aluminum and lithium release (-ln[fA1] and -ln[fLi]) from TCLP. The effects of Al2O3, B2O3, CaO, CeO2, Gd2O3, K2O, Li2O, MgO, Na2O, PbO, SiO2, SnO, TiO2, and ZrO2 were found to be linear for T5 and non-linear for chemical durability. T5 is increased most by the addition of SiO2 followed by Al2O3, and decreased most by Li2O followed by Na2O. MCC-1 ln[rB] and ln [rNa] are increased most by the addition of Li2O followed by MgO and decreased most by Al2O3 followed by ZrO2. PCT In[rB] and In[rNa] are increased most by the addition of MgO followed by B2O3 and decreased most by Al2O3 followed by SiO2. The relative effect of components on rB and rNa are similar for both PCT and MCC-1. TCLP -ln[fAl] and ln[fLi are increased most by the addition of SiO2 followed by SnO and decreased most by Li2O.

Results are reported on several new glass and glass-ceramic waste formulations for plutonium disposition. The approach proposed involves employing existing calcined high level waste (HLW) present at the Idaho Chemical Processing Plant (ICPP) as an additive to: 1) aid in the formation of a durable waste form and 2) decrease the attractiveness level of the plutonium from a proliferation viewpoint. The plutonium, PuO2, loadings employed were 15 wt% (glass) and 17 wt% (glass-ceramic). Results in the form of x-ray diffraction patterns, microstructure and durability tests are presented on cerium surrogate and plutonium loaded waste forms using simulated calcined HLW and demonstrate that durable phases, zirconia and zirconolite, contain essentially all the plutonium.

A titanate-based ceramic waste form, rich in phases structurally related to zirconolite (CaZrTi2O7), is being developed as a possible method for immobilizing excess plutonium from dismantled nuclear weapons. As part of this program, Lawrence Livermore National Laboratory (LLNL) produced several ceramics that were then characterized at Argonne National Laboratory (ANL). The plutonium-loaded ceramic was found to contain a Pu-Gd zirconolite phase but also contained plutonium titanates, Gd-polymignyte, and a series of other phases. In addition, much of the Pu was remained as PuO2-x. The Pu oxidation state in the zirconolite was determined to be mainly Pu4+, although some Pu3+ was believed to be present.

Synroc-C containing 10wt% simulated PW-4b-D HLW including 0.62 wt% 239Pu was subjected to MCC-1 type leach tests at 70°C in deionised water, silicate and carbonate leachates for 53 d and deionised water for 2472 d. The normalised total (i.e. unfiltered leachate + vessel wall) Pu leach rates in deionised water, silicate and carbonate leachates for periods up to 53 d were found to be of the order of 10-5, 10-4 and 10-4 g m-2 d-1 respectively. After 2472 d, the differential, normalised, Pu leach rate in deionised water dropped to ∼5 × 10-6 (total) and ∼5 × 10-8 (solution - after filtration through a 1000NMW filter) g m-2 d-1. SEM and AEM were used to characterise our starting material and investigate the secondary phases on the surfaces of leached Synroc-C discs. Calculated and measured normalised Pu leach rates are compared and the partitioning of Pu between zirconolite and perovskite is discussed.

Disposition of weapons fissile material as once-through MOx fuel in reactors or immobilization in waste glass would result in end products requiring geologic disposal. Similar considerations apply to some spent fuels in a geologic repository. The criticality potential of these end products in a geologic setting is an important consideration for the final disposal of these materials. In this document we present calculations that help define parameters and configurations for criticality, including those that, if the criticality is achieved, could lead to autocatalytic power transients. It is not the purpose of this paper to define scenarios that could reasonably lead to criticality or to establish the probabilities that critical configurations could occur. We do suggest some parameters that should be considered by waste package designers and repository analysts to ensure that criticality cannot occur.

Immobilization by vitrification is one potential disposition option for a portion of the United States’ excess plutonium inventory. Research has been performed to determine the glass forming region of a frit, plutonium and rare earth system. The frit contains mainly oxides of aluminum, silicon and boron; small amounts of ZrO2 and SrO are also included. The rare earth elements provide a flux to the glass during processing. The rare earths are also added as neutron absorbers (to prohibit criticality) in the final vitreous product.

This report will show the compositional region, using Th as a Pu surrogate, that should be targeted in future studies for maximum Pu solubility in the lanthanide borosilicate glass system. Durability data and process variables will also be provided.

The following problems related to the destroyed 4th Unit of the Chernobyl Nuclear Power Plant (NPP) are considered.

Monitoring of the unit's current state, including:

1. 1) Monitoring of the atmospheric release of radioactive elements in the form of aerosols. In this case the problem is the unqualified area of cracks and holes that have formed in the “Sarcophagus” (the “Shelter”) since 1986. In addition, the method used to evaluate the release, in which an uncontrolled escape of radioactive contamination is possible, raises objections;

2. 2) Monitoring of the aquatic medium both inside the “Shelter” and on the industrial site. At present, the quantity of water in the 4th Unit is about 3000 m, which in fact is low- and intermediate-level radioactive waste. This quantity is not constant and essentially varies depending on the season, amount of atmospheric precipitation, the intensity of dust suppression, etc. In this case the problem is that the paths of water migration inside the unit are not known, so that an uncontrolled penetration of water into the ground, via the damaged foundation slab or the unit's walls, is quite possible;

3. 3) Monitoring of the state of lava-like fuel-containing masses (LFCM) and the fuel. The problem is that the exact location of the fuel in the 4th Unit, the amount of fuel in major aggregates of LFCM, and the nuclear safety of these aggregates are all unknown to date.

The arrangements for the stabilization of the 4th Unit first of all include the reinforcement of the structural construction.

The transformation of the “Shelter” into an ecologically safe system is subject to many questions still unresolved. In particular:

• Whether to create a common “Shelter-2” for the 3r and 4th Units or to merely encase the destroyed 4th Unit in concrete (the project “Cube”);

• Whether to peace a plant for conditioning and reprocessing of radioactive waste under the roof of the “Shelter-2”, or to build this plant separately;

• Where to create a temporary repository for containers of radioactive waste from the ChNPP 4th Unit;

• How to solve the problem of radioactive waste permanent storage in the territory of the Ukraine with proper observation of all nuclear and radiation safety requirements