Waste Package Emplacement Research Articles

Emplacement optimization for the first-stage disposal facility of LILW complex disposal facility by employing simulated annealing algorithm

• POWER is an emplacement optimization program for a complex LILW disposal facility in Korea. • POWER was developed using Simulated Annealing algorithm. • POWER was applied to an emplacement optimization problem consisting of 69 waste drums. • A series of calculations was performed to verify the effectiveness of POWER. • It was shown that POWER is capable of finding the near-optimal emplacement schemes for the disposal facility. In order to optimize the emplacement of LILW (Low- and Intermediate-Level radioactive Waste) complex disposal facility, an emplacement optimization program POWER (Program for Optimization of Waste package Emplacement in Repository) based on Simulated Annealing (SA) algorithm is developed. The POWER program derives near-optimum emplacement schemes that minimize the objective function consisting of total dose, radiotoxicity flux, and radiotoxicity concentration obtained from GoldSim calculations for the post-closure safety assessment for LILW rock cavern disposal facility. An optimization problem of sixty nine drums with different degradation mechanism, source term, and drum type to be disposed of into the first-stage disposal facility is defined first and a series of optimization calculations is performed. It is verified that the POWER can generate the near-global optimized emplacement scheme from SA process.

Understanding cementitious backfill interactions with groundwater components

Abstract Many concepts for the disposal of intermediate-level radioactive waste specify the emplacement of waste packages in underground vaults backfilled with cementitious material. The cementitious backfill performs a number of functions including maintaining alkaline conditions and providing a high sorption capacity for radionuclides. In the UK, the Nirex Reference Vault Backfill (NRVB) has been developed for such a role. To improve our understanding of its long-term performance, a detailed programme was undertaken to investigate interactions of the NRVB with selected groundwater components using sodium sulphate and magnesium chloride solutions. Experiments using the selected solutions were performed under conditions of constant flow, over timescales of up to 860 days. Elemental composition and pH of the egressing eluent were determined throughout the experiments. At the end of the experiments, detailed solid analyses were performed on samples from various locations in the NRVB using a wide range of techniques. In the sodium sulphate experiment, sulphate attack resulted in the formation of ettringite, thaumasite and gypsum. For the magnesium chloride case, the results were dependent on the concentration of the magnesium in the ingressing solution. At higher concentration, precipitation of brucite at the upstream surface of the NRVB led to clogging and termination of the experiment after about one month. At lower concentration, no such clogging occurred, despite almost all the ingressing magnesium being removed by the NRVB over timescales of more than two years. For both systems, high pH values of the eluent were maintained over the full periods studied; for the sodium sulphate experiment the final pH was 12.7 (after 180 sample volumes) and for the low concentration magnesium chloride case pH 11.5 (after 749 sample volumes). A “top-down” modelling approach was applied to interpret the results, involving solution equilibration with the NRVB represented by a simple set of mineral phases. Further processes were modelled only where they improved the overall fits. The resulting models were considered to be simple while maintaining consistency with the major features of the experimental data sets. This has helped to provide mechanistic interpretations of the evolution of NRVB, based on relatively simple sets of secondary mineral phases, early kinetic control of mineral dissolution and diffusion-restricted release of portlandite at longer times.

Thermal-Hydrologic Model Study for an Alternative Waste Package Design for Yucca Mountain Repository

The long-term thermal, hydrologic, and psychrometric storage environment of nuclear waste is analyzed within an emplacement drift at Yucca Mountain Repository in Nevada. Pertinent issues regarding temperature, relative humidity, and liquid water in contact with the waste packages are studied for a modified design currently considered by the U.S. Department of Energy (DOE). For cost reduction and improved repository performance, the proposed design implements a slight modification in the waste package emplacement sequence and thermal load. The main change is an increase from 44 boiling water reactor (BWR) to 52 BWR fuel assemblies to reduce the number of waste packages for the same storage capacity. The results of the analysis show that acceptable temperature, moderate relative humidity, and no liquid water are expected on the hot waste package including the new BWR containers of the proposed design for the 5000-yr study period. The cold DOE high-level waste and the colder defense spent nuclear fuel containers in the alternative design will experience about the same amount of condensates as those in the DOE baseline design.

Investigations concerning the semihydrostatic emplacement of waste packages in vertical boreholes

The semihydrostatic emplacement of waste packages in unlined vertical boreholes in salt domes involves the uncontrolled deposition of the packages in the borehole and their complete enclosure in crushed salt. In order to calculate the pressure distribution in the backfilled borehole section, the borehole filling composed of packages and crushed salt is treated by approximation as a homogeneous medium in the theoretical model. The eigenvalues of the pressure tensor depend on the radial distance r from the borehole centre and on the depth z in the borehole and are thus directionally dependent. Emplacement-typical crushed-salt parameters such as lateral pressure coefficient and friction coefficients were experimentally determined. The calculations show that, according to the semihydrostatic model, the maximum pressure in the borehole does not exceed 107 kPa for a 4001 waste package, that this maximum pressure is already reached at a depth of 9.26 m on the borehole wall and that there is an optimum borehole radius for which stressing of the emplaced packages is minimized.

Safety assessment of the thermal influence upon the host rock of the planned Konrad repository / Sicherheitsanalysen zur thermischen Beeinflussung des Wirtsgesteins beim geplanten Endlager Konrad

The safety assessment of the thermal influence upon the host rock is part of the comprehensive safety assessment for a repository. The effects of the decay heat generated by the emplaced radioactive waste must be thoroughly analysed. Based on the plannings for the Konrad repository, a facility for the disposal of radioactive waste with negligible heat generation, calculations of the thermal influence upon the host rock have been carried out, and resulting requirements on the emplacement of waste packages have been derived. The procedure and the results are described in this paper

Near-Field Diffusion Releases Through One and Two Finite Planar Zones from a Nuclear Waste Package

For a nuclear waste package emplacement in a potential repository in partially saturated rock, a rock rubble or backfill zone may act more as a barrier than as a pathway for diffusive release of radionuclides. The authors approximate the diffusive transport process using one-dimensional, one- and two-barrier geometries. The one-barrier model suffices when the effective diffusion coefficient in the first zone, the rubble, is substantially lower than that in the second zone, the host rock. For more generally, such as two zones of comparable diffusivities, or for an additional barrier zone, the authors model two barrier zones both of finite extent. They present solutions for three types of radionuclide mobilization at the source: a pulse transient input, a steady input rate, and a constant concentration. The algebraic series form of the solutions aids analysis of sensitivity of break-through times and peak release rates. For the one-zone case, dimensionless parameters allow plotting of the family of transient solutions on a single graph. Comparisons between results of one- and two-zone models and with published results for different geometries and solution methods support verification of the solutions in this study.

Final disposal in Germany / Endlagerung in Deutschland

The disposal of radioactive waste is planned and has been carried out in Germany in deep geological formations only. From 1965 until 1978, procedures and techniques for the emplacement of waste packages were developed and tested in the abandoned Asse salt mine. At present, three sites are considered for disposal. The Gorleben salt dome in the north-east of Niedersachsen is being investigated for its suitability for disposal of all kinds of radioactive waste. The abandoned Konrad iron ore mine in Niedersachsen is envisaged for the disposal of radioactive waste with negligible heat generation. The former Morsleben salt mine in Sachsen-Anhalt, which was used as a repository for short-lived low and intermediate level waste from 1981 until 1991, is intended to resume its operation within a short period of time

Analytic Models for Assessing the Performance of Engineered Barriers in a Basalt Repository

ABSTRACTAnalytic models have been developed and used to study the performance of engineered barriers in a basalt repository with waste packages vertically emplaced below the storage room floors. Models have been developed for the groundwater flow field and for waste transport through the waste package backfill and up to the top of the storage room. The lowest release rates at the top of the storage room occur for a room backfill with high hydraulic conductivity (greater than the surrounding rock), high porosity and good sorption. Such a room backfill is much more effective than waste package backfill at delaying releases from the engineered system, especially for short-lived nuclides. A repository design with horizontal emplacement of waste packages has also been studied and found to have higher release rates, owing to a larger cross-section to groundwater flow and the absence of an overlying storage room.