Joonhong Ahn

Geological repository systems for safe disposal of spent nuclear fuels and radioactive waste

Part 1 Introduction: Multiple-barrier geological repository design and operation strategies Spent nuclear fuel recycling practices and impacts Near-surface, intermediate depth and borehole disposal Underground research facilities and rock laboratories. Part 2 Geological repository systems: Crystalline systems Clay systems Assessing long-term stability Far-field process analysis and radionuclide transport modelling. Part 3 Engineered barrier systems: Immobilisation of spent nuclear fuel and radioactive waste Low-pH concretes Smectitic buffer and backfill Near-field processes, evolution and assessment Nuclear waste canisters, corrosion and long-term performance Post-containment performance. Part 4 Performance assessment: Safety assessment for deep geological disposal Safety assessment for near-surface disposal Treatment of uncertainty in the performance assessment Assessment of expert judgments Knowledge management systems. Part 5 Radiation protection, regulation and social engagement: Radiation protection principles and standards Risk-informed, performance-based regulations Environmental monitoring programs and public engagement Methods for social dialogue.

Relationship Among Performance of Geologic Repositories, Canister-Array Configuration, and Radionuclide Mass in Waste

Abstract The relationship among the repository performance, the canister-array configuration, and the radionuclide mass in waste has been investigated by developing a radionuclide-transport model, where multiple waste canisters and their spatial configuration are taken into account. A mathematical analysis and numerical results show that the radionuclide concentration in the groundwater leaving the canister array increases with the number of canisters included in a water stream parallel to the array axis, but not necessarily in a linear manner. The dependency on the number of canisters is determined mainly by canister-array configuration to the water flow and by model assumptions for transport between multiple canisters. Reduction in the initial mass loading in the waste can potentially have significant effects on the repository performance. The way the mass-reduction effects on the repository performance appear is related to the canister-array configuration. Thus, designs for a repository and a partitioning-transmutation system should be done in a coupled manner.

Considerations of Autocatalytic Criticality of Fissile Materials in Geologic Respositories

Potential routes to autocatalytic criticality in geologic repositories are systematically assessed. If highly enriched uranium (HEU) or {sup 239}Pu are transported and deposited in concentrations similar to natural uranium ore, in principle, criticality can occur. For some hypothesized critical configurations, removal of a small fraction of pore water provides a positive feedback mechanism that can lead to supercriticality. Rock heating and homogenization for these configurations can also significantly increase reactivity. At Yucca Mountain, it is highly unlikely that these configurations can occur; plutonium transport would occur primarily as colloids and deposit over short distances. HEU solute can move large distances in the Yucca Mountain setting; its ability to precipitate into critical configurations is unlikely because of a lack of active reducing agents. Appropriate engineering of the waste form and the repository can reduce any remaining probability of criticality.

Integrated Radionuclide Transport Model for a High-Level Waste Repository in Water-Saturated Geologic Formations

Presented are results of a mathematical analysis on radionuclide transport in parallel planar fractures in water-saturated geologic formations integrated with the source term model, where precipitation of hardly soluble species at the waste-form alteration location and subsequent radionuclide transport in the engineered barriers are considered. Radioactive decay chains of an arbitrary length are considered. A computer code has been developed based on the analytical solutions. With a transport distance of 100 m through the natural barrier, a four-orders-of-magnitude reduction in the total hazard is observed. Thus, the importance of the region in the vicinity of the engineered barriers in the context of the safety assessment can be pointed out. Because the region is disturbed by repository construction, further analysis must be performed by taking into account differing geochemical, hydrological, and mechanical properties from those in the undisturbed host rock. Because the major contributors in the host rock are the decay daughters of minor actinides, recovery of minor actinides reduces the total hazard evaluated at the exit of the geosphere. However, the radiological hazard would be reduced much more effectively by the 100-m-thick geologic formation around the repository than by even a 99% recovery of the actinides.

Diffusion of the 241Am → 237Np Decay Chain Limited by Their Elemental Solubilities in the Artificial Barriers of High-Level Radioactive Waste Repositories

A mathematical analysis of the diffusion of the [sup 241]Am [r arrow] [sup 237]Np decay chain in the artificial barrier of a high-level radioactive waste repository is presented. First, analytical solutions obtained are for the space- and time-dependent concentration of [sup 241]Am in the artificial barrier and the time-dependent amount of americium precipitated at the surface of the waste glass, based on the assumption of the congruency of the radionuclides with solubility-limited dissolution of the glass matrix. The effects of solubility sharing with coexisting [sup 243]Am are considered. Transport and precipitation of [sup 237]Np in the artificial barrier are analyzed by dividing the time domain into a small time domain, where the [sup 241]Am concentration is so large that [sup 237]Np precipitation is dominant, and a large time domain, where the [sup 241]Am becomes negligible and the precipitation region shrinks by diffusion from the precipitation front. The equation for the movement of the precipitation front is obtained. As the overpack lifetime increases, the effect of neptunium precipitation becomes less significant. With a lifetime longer than [approximately]6000 yr, an earlier model, where neptunium is treated as a mother nuclide and the precipitation occurs only at the glass surface, can be used.

Congruent Release of Long-Lived Radionuclides from Multiple Canister Arrays

Abstract Results are presented of an analytical study of mass release of a long-lived radionuclide from multiple waste canisters placed in a water-saturated repository in a two-dimensional array configuration. The radionuclide is assumed to be released congruently with the dissolution of the waste matrix. The concentration and release rate of the radionuclide from the downstream side of the repository region are numerically calculated to observe the effects of canister multiplicity and the leach time of the waste form. Peak values of the concentration and the release rate have been analytically formulated. For numerical illustration, the case of a Japanese repository concept is considered, where canisters containing vitrified wastes are placed in a water-saturated granitic rock. For the illustration, the nuclide 135Cs is chosen, which is characterized by a long half-life and high mobility in the assumed geologic media. The peak exit concentration becomes independent of the number of waste canisters in the flow direction if the number is sufficiently great. This peak value is a theoretical upper bound of the exit concentration, regardless of the number of canisters or the waste matrix leach time. The model is suitable for assisting in the design of a repository since the effects of the canister array configuration are reflected by the peak exit concentration and the peak release rate.

Criticality safety assessment for a conceptual high-level-waste repository in water-saturated geologic media

An assessment for the criticality safety of a conceptual repository for vitrified high-level radioactive waste from reprocessed fuel of commercial light water reactors in a water-saturated granitic rock has been performed by quantitatively estimating the mass of fissile {sup 235}U existing in the entire far field as the performance measure. The uncertainties associated with the performance measure have been obtained by a statistical analysis with the Latin hypercube sampling method. With the assumed probability distribution functions for the model parameters, the mass of {sup 235}U released from the repository and existing in the far field at 100 million years is estimated to be < 40 kg with a 90% confidence level. This implies that all {sup 235}U existing in the entire far field at that time must accumulate in a single location for an overmoderated criticality event to occur in granitic rock.

An Environmental Impact Measure for Nuclear Fuel Cycle Evaluation

Review of the models and measures for repository performance assessment has revealed that dedicated measures for environmental impacts need to be developed for the purpose of nuclear-fuel-cycle evaluation from the viewpoint of environmental impact minimization. The present study proposes the total toxicity index of released radionuclides that have accumulated in the region exterior to the repository as an environmental impact measure. The measure is quantitatively evaluated by a radionuclide transport model that incorporates the effects of canister-array configuration and the initial mass loading in the waste canister. With the measure, it is demonstrated that the environmental impact of the repository can be effectively reduced by reduction of the initial mass loading and change in the canister-array configuration in the repository. Environmental impacts of the mill tailings and the depleted uranium are as important as those from the high-level radioactive wastes repository. For a fair comparison of various fuel cycles, the sum of these impacts should be compared.

Sorption behavior of Np (IV), Np (V) and Am (III) in the disturbed zone between engineered and natural barriers

In order to be more confident of the performance assessment of high-level radioactive waste disposal, radionuclide transport must be investigated in more detail in the disturbed host rock region adjacent to the engineered barriers where disturbance has been introduced during the construction and waste-emplacement period. Geochemical, hydrological, and rock-mechanical properties should be quite different from those of undisturbed host rock. We have to elucidate the effect of bentonite intrusion into intersecting fractures from the standpoint of radionuclide confinement. In the present work, sorption distribution ratios (Kds) of Np and Am are measured experimentally for various values or redox potential (Eh) in a simulated rock fracture filled with bentonite. The Kd of Am is approximately 6.5×103 ml/g and found to be insensitive to the redox potential. Under anaerobic conditions, the Kd of Np is approximately 6×104 ml/g. Under aerobic conditions, Kd is as small as 30 to 100 ml/g. This is the first report to measure the sorption behavior of Np and Am in a simulated rock fracture filled with bentonite (namely, in a disturbed zone) under pH, Eh and ionic strength control. We aan make use of these Kd data for numerically evaluating the mass transfer from bentonite filled fractures into the water-flowing fracture network1.

A SYSTEMS ASSESSMENT FOR THE KOREAN ADVANCED NUCLEAR FUEL CYCLE CONCEPT FROM THE PERSPECTIVE OF RADIOLOGICAL IMPACT

In this study, we compare the mass release rates of radionuclides (1) from waste forms arising from the KIEP-21 pyroprocessing system with (2) those from the directly-disposed pressurized-water reactor spent fuel, to investigate the potential radiological and environmental impacts. In both cases, most actinides and their daughters have been observed to remain in the vicinity of waste packages as precipitates because of their low solubility. The effects of the waste-form alteration rate on the release of radionuclides from the engineered-barrier boundary have been found to be significant, especially for congruently released radionuclides. The total mass release rate of radionuclides from direct disposal concept is similar to those from the pyroprocessing disposal concept. While the mass release rates for most radionuclides would decrease to negligible levels due to radioactive decay while in the engineered barriers and the surrounding host rock in both cases even without assuming any dilution or dispersal mechanisms during their transport, significant mass release rates for three fissionproduct radionuclides, 129 I, 79 Se, and 36 Cl, are observed at the 1,000-m location in the host rock. For these three radionuclides, we need to account for dilution/dispersal in the geosphere and the biosphere to confirm finally that the repository would achieve sufficient level of radiological safety. This can be done only after we have known where the repository site would be sited. The footprint of repository for the KIEP-21 system is about one tenth of those for the direct disposal.