Low And Intermediate Level Waste Research Articles

Deformations of bentonite-sand mixture without lateral confining pressure subject to high suctions changing

Generation of electricity in nuclear power plants have supplied further energy to society in the world, and remain arising of radioactive waste as results of performance industry with activity. In fact, all of radioactive waste encourage some categories types as following; two most important points are high-level waste (HLW) and low- and intermediate-level waste (LLW). To end disposal completely, every country operated Nuclear Energy plant foresees further deep geological repository. This study conducted out measurement of deformations due to changing suction, which were volume change, axial strains and radial strains. Used triaxial compression apparatus was improved to be possible controlling suction for air flow circulation, and measured both axial and radial strains and variations of mass under drying-wetting cyclic in suction. Required suction rage is from 2.8 MPa to 296 MPa, and it corresponds to from 98.0 % to 11.0 % in relative humidity. Axial strains and radial strains for all of specimens have described obviously the shrinkage phenomena due to drying and wetting cyclic process regard to suction potential, and bentonite-sand mixture specimens are statically compacted. Axial strains compare that is larger than radial strains regard to all applied suctions. Subsequently, changing of void ratio was classified into two categories, which one was elastic components and other one was plastic component, and elastic component in void ratio was remaining larger than plastic component for high suction ranges.

Sorption and mobility of radioactive Ni in Wolsong waste repository site, South Korea

Radioactive nickels are significant environmental concerns because of their long half-lives (59Ni = 76,000 years and 63Ni = 100.1 years) and high proportions in nuclear wastes in the repositories. Therefore, the investigation of sorption and mobility of Ni at the disposal facility sites is important for the prediction and evaluation of radiological risks to the public. Herein, the batch sorption and column experiments were performed under various geochemical conditions using site-specific rock and groundwater samples collected at the Wolsong low and intermediate level waste (LILW) disposal facility site in South Korea. The batch sorption results revealed that the sorption of Ni was strongly affected by the pH. Approximately 20% Ni adsorbed at pH 5, which enhanced over 90% at pH > 8. The batch sorption results were used to develop the surface complexation model (SCM) to predict the Ni sorption and its mobility behavior. The non-electrostatic generalized composite (GC) SCM approach simulated the Ni sorption data well. We believe, our study can interpret and extrapolate the sorption and transport behavior of Ni in the underground repository conditions.

Permeability and diffusion of tritium in bentonite

Abstract. As a part of the Croatian RD&D (research, development, and demonstration) program within the Program for LILW (low- and intermediate-level waste) and SNF (spent nuclear fuel) disposal, research of permeability and the diffusion of tritium in bentonite was organized, as part of the research of the future engineering barriers for repositories of LILW and SNF. Two types of materials were prepared, namely pure sodium bentonite and a mixture of sodium bentonite (12 %) with quartz sand (88 %). Compaction of materials was conducted with a modified Proctor test, and the apparatus was designed and made at the Faculty of Mining, Geology and Petroleum Engineering at the University of Zagreb, utilizing fused deposition modeling 3D printer and PETG (polyethylene terephthalate glycol) material. Water enriched with 3H has been prepared by dilution of the National Institute of Standards and Technology (NIST) 3H standard (4926e) to 6 L. The activity of the prepared enriched water was 49 700±180 TU. One part of the device was filled with 3H-free water and the other with 3H-spiked water (0.6 L volume for each part). In total, five devices with bentonite and five devices with a sand–bentonite mixture were prepared this way. The devices were thoroughly sealed to avoid loss of sample and isotopic exchange with the atmosphere and put horizontally in holders. The water for 3H activity measurements was sampled monthly or bimonthly. A total of 8 mL of the sample was taken by syringe and transferred to a measurement vial. Each vial was additionally filled with 12 mL of a scintillator (Ultima Gold LLT, PerkinElmer) and thoroughly mixed. 3H activity was determined by a liquid scintillation counter (LSC; Quantulus 1220) at the Laboratory for Low-level Radioactivities, Ruđer Bošković Institute, Zagreb. Due to the penetration of bentonite from the barrier into the water samples, tritium-free water has relatively high activity after 2 months of the experiment (average value was 5560±240 TU). In the final sampling, after a 2-month time span, the activity reached 9670±2490 TU. Later, sampling and measurement were not possible due to the high bentonite amount within the samples. Tritium-free samples from the sand–bentonite mixture device were stable, and the first analyses (1 month after the experiment setup) showed an average activity of 533±31 TU, while the final (5-month time span from the first sampling) average activity was 12 800±3430 TU. The increase in the concentration from the beginning to the end of the experiment was almost exponential. Spiked samples from the bentonite device had an average activity of 36 950±2550 TU at the first sampling and 17 580±13 320 TU after a 1-month time leg. Later, sampling was not possible. Spiked samples from sand–bentonite mixture showed a decrease in tritium activity from the first sampling (43 390±1310 TU) to 27 210±285 TU (5 months after the first sampling). Considering the penetration of bentonite material into the water samples, only the sand–bentonite mixture can be considered for further analyses through the calculation of the diffusion.

Long-Term Experiments for Demonstrating Durability of a Concrete Barrier and Gas Generation in a Low-and Intermediate-Level Waste Disposal Facility

Long-Term Experiments for Demonstrating Durability of a Concrete Barrier and Gas Generation in a Low-and Intermediate-Level Waste Disposal Facility

Dynamic cost analysis for disposal of low and intermediate level nuclear waste in China

The management and financial problems of nuclear waste have raised the concern worldwide due to lack of experience in decommissioning and disposal. In this paper, a primary cost estimation model is proposed based on Activity-based Life Cycle Costing (AB-LCC) approach in order to influence the strategy of Low and Intermediate Level Waste (LILW) disposal in China. AB-LCC model provides a dynamic cost estimation of LILW disposal and its application is not restricted by the region and time. A dynamic analysis of resources due to time and regional differences of LILW management during its life cycle time has been performed on the basis of Activity-based Costing (ABC) matrix. The static and dynamic costs of the primary and secondary activities at each stage of the disposal repositories are obtained. The model is verified by comparison with statistics from the cases of Aube repositories and difference is approximately less than 5%. Therefore AB-LCC model is used to estimate the LILW disposal costs in China with different disposal modes. Finally, LILW management strategy was suggested to the Chinese policy-makers that the Northwest Centralized Disposal Mode (NCDM) is more cost-saving than the Coastal Independent Disposal Mode (CIDM) and Gansu province is the most recommended disposal site for economic considerations.

ANALYSIS OF DOSE RATES AROUND THE SLOVENIAN SILO-TYPE LILW REPOSITORY USING ADVANTG

The ADVANTG code was used to analyze dose rates from the proposed Slovenian silo-type low and intermediate level waste (LILW) repository. Detailed calculations of dose rates are challenging as gamma-sources are located in thick concrete containers and the effect of backscattering in the air has to be considered. Compared to analog MCNP simulations maximum relative speed-up of up to 9500 was achieved by using ADVANTG. We have analyzed the influence of different repository configurations on the final annual γ-dose rates. For different measurement positions, which were located from 10 m to 150 m from the outer edge of the repository, γ-dose rates were calculated. The results show that for all configurations of the open repository where the waste from different sources is deposited, except using the waste from Krško NPP, the calculated annual γ-dose rates for all measurement positions were lower than the prescribed limit values for professionals of 20 mSv/year. To ensure that the γ-dose rates are lower than the prescribed limit values even for the case with the waste from Krško NPP an additional analysis was made where the concrete casks, in which the waste is stored, were modelled in more detail and the quantity of additional grout on top of the waste was varied.

Gasification-based thermal treatment of Low and Intermediate Level Waste containing organic matter

VTT has developed a gasification-based technology for processing low- and intermediate-level waste (LILW) rich in organic matter. VTT’s process is based on well-controlled thermal fluidised-bed gasification followed by efficient gas cleaning, gas conditioning/oxidation, wet scrubbing of the oxidised gas (flue gas), and finally, filtering through a high efficiency particulate air (HEPA) filter, which acts as a backup cleaning system. In the case of an organic ion exchange resin (IXR) the primary product from thermal gasification is fine filter dust. In addition, the process produces some bottom ash, which consists mainly of the bed material. The wet scrubber liquid may also contain some activity. Demonstration trials were carried out using unspent organic IXR containing a small amount of added stable Cs in order to simulate Cs content in spent IXR. Gasification tests confirmed the capability of the process to remove organic matter from the IXR and clean the resulting off-gas as required. Simulated waste IXR was reduced to approximately 1 wt-% of its original mass before immobilisation. Filter dust and bottom ash have to be immobilised before final disposal. VTT has previously developed an advanced immobilisation technology based on geopolymerisation and this has been applied to samples from the THERAMIN test trials. VTT’s technology has been designed as a compact process, which can be operated at the nuclear power plant site. Until now, all development and verification test trials have used simulated waste materials. The next step is the demonstration of the process using real radioactive waste.

Temporal changes of geochemistry and microbial community in low and intermediate level waste (LILW) repository, South Korea

Microbes play a major role in affecting the radioactive wastes disposed of in underground wastes repositories. In particular, low- and intermediate-level waste (LILW) contains biodegradable constituents that can affect stability and mobility of radionuclides by the changes of geochemical conditions due to microbial activity. Despite these potential impacts, the bacterial communities in underground repository remain largely unexplored and previous studies have focused on the traditional culture-dependent methods. We examined the bacterial community in a large-scale in situ container packed with simulated waste in LILW underground repository of South Korea using 454 pyrosequencing. Taxonomic analysis showed the presence of Curvibacter, Azonexus, Pseudomonas, Novosphingobium, Acidovorax, and Victivallis as the dominant bacterial genera. Based on the thermodynamic and geochemical results, the precipitation of calcite was estimated to occur in the presence of genus Curvibacter. Although the flow rate of groundwater input had an impact on the bacterial communities, the community structure was resilient to the change of groundwater’s velocity. In addition, total Fe concentration, [Fetotal] was positively correlated to [Cl−] in this system. The temporal changes of geochemical parameters and bacterial communities provide insight for understanding of the microbial activity inside the large scale container and additional biogeochemical information for long-term risk assessment of disposal facility.

Modeling of the Cs137 and Sr90 contamination transportation process performed for the vicinity of National Radioactive Wastes Disposal in Różan (NE Poland)

This paper contains the results of radionuclides transportation modeling under National Radioactive Waste Disposal (NRWD) grounds in Różan (northeast Poland). The disposal is of the low- and intermediate-level waste (LILW) type. We simulated the radionuclides transportation process through sandy soils. The simulation was performed in a self-written simulator in Scilab using the finite difference method. The model included diffusion, advection and radioactive decay. The model was tested according to convergence and stability. Assuming the hydrological gradient being 0.008, the contamination transportation time was 30–46 years depending on the modeled problem. The modeled distance of 600 m was from underneath the disposal to the exudation in the Narew ravine. Radioactive decay for both cesium (Cs137) and strontium (Sr90) had a significant impact on the results. The model proved to be a useful tool for performing simple scientific simulations. This survey was part of a PhD thesis.

Demonstration of structural performance of IP-2 package by simulation and full-scale horizontal free drop test

Packaging systems for the transportation of radioactive wastes have to be designed according to rigorous acceptance criteria and requirements in order to protect people and environment against radiation exposure and contamination risk. The IAEA requirements for type IP2 (Industrial Package Type 2) packages include to carry out free drop tests that represent normal conditions of transport. In such conditions, obviously, the required containment capability of the package has to be ensured.In this study the mechanical performances of a new Italian packaging system for the transportation of low and intermediate level wastes (LILW) undergoing horizontal free drop test are investigated. Especially, deformations caused in the sealing area of the package, which can affect the capability of the containment system, are evaluated.The carried out numerical analyses and experimental tests, at the lab. Scalbatraio of the DICI- University of Pisa, are presented and discussed.Numerical analyses (by qualified MARC® code) have been performed to investigate the stress histories in the bolts, lid, and package body as well as the deformations in the sealing area and the compression conditions of the gasket.Localised stress appeared at the flange and at the bottom of packaging system. The maximum stresses resulted lower than the stress limits, so the structural integrity of the package was maintained and confirming its tightness. As a consequence of the primary impact a local deformation appeared at the primary lid, no cliff edge or loss of the safety features resulted.

중·저준위 방사성폐기물 처분시설 폐쇄후 기체이동

본 연구에서는 중·저준위방사성폐기물 처분시설(이하 처분시설)에서 발생하는 기체의 이동현상을 예측하기 위한 2차원 수 치 모델링을 수행하였다. 또한, 기체 이동 모델링에서 주요 입력변수로 적용되는 사일로 콘크리트의 기체침투압(gas entry pressure)와 기체 투과도(gas permeability)를 실측하여, 모델링 입력변수로 적용하였다. 사일로 콘크리트의 기체침투압(gas entry pressure)와 기체 투과도(gas permeability)는 각각 0.97±0.15 bar 및 2.44×10-17 m2로 측정되었다. 기체 이동 모델링 결과, 사일로 내부에서 발생하는 수소 기체는 기상으로 이동하지 않고 지하수에 용해되어 지하수와 함께 생태계로 이동하는 것을 알 수 있다. 또한, 폐쇄 후 약 1,000 년 후 부터 사일로 상부부터 수소기체 밀도가 증가하기 시작하는 것으로 예측되었 다. 따라서, 사일로 내부에서 발생된 기체는 기상으로 사일로 내부에 축적되지 않으며, 이로 인해 사일로 콘크리트의 내구성 에 영향을 미치지 않을 것으로 판단된다.

Dynamic assessments on high-level waste and low- and intermediate-level waste generation from open and closed nuclear fuel cycles in Republic of Korea

We estimated the generation of low- and intermediate-level waste (LILW) and high-level waste (HLW) from open and closed nuclear fuel cycles. The closed fuel cycle reflects the development and deployment of fast reactors and pyroprocessing from 2013 to 2100, while the open fuel cycle only considers pressurized water reactors. The closed fuel cycle hardly affects short-term spent fuel management but can save nearly 60% space of interim storage compared with the open fuel cycle. Compared with the open fuel cycle, the accumulated volume of HLW can be significantly reduced in the closed fuel cycle up to over 95% in 2100. For this volume reduction, high heat generating fission products should be separated from transuranic waste in pyroprocessing and stored in decay storages for a few hundred years. Mining and milling waste in the closed fuel cycle decreases by about 31%. In contrast, the closed fuel cycle generates 3.0%–4.5% more LILW than the open fuel cycle because fast reactors and pyroprocessing produce more LILW and conversion, enrichment, and fabrication produce less LILW. In the closed fuel cycle, operation and decommissioning wastes from reactor and pyroprocessing, respectively, contribute to 74% and 8% of LILW excluding mining and milling waste.

Clays in Natural and Engineered Barriers for Radioactive Waste Confinement: an introduction

The concept of engineered geological disposal has been developed for the safe long-term management of long-lived radioactive waste. This involves emplacement of radioactive waste in deep geological repositories that contain and isolate the waste and, consequently, protect humans and the environment. The ‘multiple barrier concept’ is the cornerstone of all proposed schemes for the geological disposal of radioactive waste. Based on the principle that uncertainties in performance can be minimized by conservatism in design, the concept invokes a series of complementary barriers, both engineered and natural (geological), between the waste and the surface environment. Each successive barrier represents an additional impediment to the movement of radionuclides. Regarding engineered barriers, for lowand intermediate-level waste the waste may be incorporated in a relatively stable and inert matrix such as cement, bitumen, lead-alloy or polymer resin (the choice varying depending on the waste management organization); glass may be used in the case of certain high-level reprocessing wastes. Owing to the very low leach-rate of glass in groundwater, vitrification is widely accepted to be one of the best methods of immobilizing the aqueous products from the reprocessing of spent fuel. Many waste containers will provide some form of physical barrier to groundwater. However, because of the relatively small volumes of waste involved, spent fuel, vitrified waste and other highly active wastes will be totally encapsulated in corrosion-resistant metal canisters that are designed to prevent groundwater entry for extended time periods in excess of 100 000 years. Depending on the disposal concept, engineered barriers may comprise the buffer/backfill medium enclosing the waste containers, the tunnel/borehole liner, and the backfill and high-integrity seals placed in the repository access ways or emplacement boreholes. The buffer/backfill medium enclosing the waste will often also provide both a physical and a chemical barrier to radionuclide migration. The functions of the engineered/chemical barriers are: † to reduce the rate of corrosion of the waste containers and thus extend their life; † to limit the rate of hydraulic transport; † to limit the release of radionuclides from the waste-form to the far-field (geosphere) after container failure; † to limit the migration of radionuclides along the pathway provided by the access tunnels and shafts of a repository or the boreholes in the case of a deep borehole emplacement.

Tracer transport modeling with the Alliances platform in the presence of evapotranspiration

AbstractThe knowledge and understanding of water flow and solute transport in the unsaturated zone is becoming increasingly important especially in mitigation of groundwater pollution. Fate of radionuclide in the geological environment is a topic to address in performance and safety assessment studies for nuclear waste disposal and may be modeled considering flow and transport in porous media. However, often, due to the heterogeneity and anisotropy of the real systems, the computer simulations may be difficult to render the real behavior. This paper addresses the simulation of a tracer transport in the unsaturated zone of the Saligny site, the potential location for the Romanian low and intermediate level waste (LILW) disposal. Computation was based on experimental data and was performed with the Alliances platform, a numerical tool developed by French organizations CEA, ANDRA and EDF. In order to obtain information regarding the solute migration in depth and the solute lateral dispersion, the dispersivity coefficients of iodine were investigated in order to match the experimental concentration determined on samples from different locations of the site. A close fit of the simulation over experimental data for the water saturation profile at a depth of 0.5 m in transient state was targeted by taking into account evapotranspiration in order to obtain a realistic estimation of the water infiltration in the porous media. Dispersivity coefficients obtained from the simulation of the tracer transport are in good order of magnitude for the unsaturated area and allow to have a good preview of the tracer plume. However, further investigations are recommended on new samples in order to validate the migration of the tracer plume as expected.

Unsaturated water flow and tracer transport modeling with Alliances

Understanding water flow and solute transport in porous media is of central importance in predicting the radionuclide fate in the geological environment, a topic of interest for the performance and safety assessment studies for nuclear waste disposal. However, it is not easy to predict transport properties in real systems because they are geologically heterogeneous from the pore scale upwards.This paper addresses the simulation of water flow and solute transport in the unsaturated zone of the Saligny site, the potential location for the Romanian low and intermediate level waste (LILW) disposal. Computation was based on the current available experimental data for this zone and was performed within Alliances, a software platform initially jointly developed by French organizations CEA, ANDRA and EDF.The output of the model developed was compared with the measured values in terms of saturation profile of the soil for water movement, in both steady and transient state. Very good results were obtained for the saturation profile in steady state and a close fit of the simulation over experimental data for the water saturation profile at a depth of 3m in transient state.In order to obtain information regarding the solute migration in depth and the solute lateral dispersion, a tracer test was launched on site and dispersivity coefficients of the solute were fitted in order to match the experimental concentration determined on samples from different locations of the site. Results much close to the experiment were obtained for a longitudinal dispersivity coefficient of 4 and a transversal dispersivity coefficient of 4, indicating a high tendency of radial dispersion of the solute rather than migration in depth.

Korean Status and Prospects for Radioactive Waste Management

The safe management of radioactive waste is a national task required for sustainable generation of nuclear power and for energy self-reliance in Korea. Since the initial introduction of nuclear power to Korea in 1978, rapid growth in nuclear power has been achieved. This large nuclear power generation program has produced a significant amount of radioactive waste, both low- and intermediate-level waste (LILW) and spent nuclear fuel (SNF); and the amount of waste is steadily growing. For the management of LILW, the Wolsong LILW Disposal Center, which has a final waste disposal capacity of 800,000 drums, is under construction, and is expected to be completed by June 2014. Korean policy about how to manage the SNF has not yet been decided. In 2004, the Atomic Energy Commission decided that a national policy for SNF management should be established considering both technological development and public consensus. Currently, SNF is being stored at reactor sites under the responsibility of plant operator. The at-reactor SNF storage capacity will run out starting in 2024. In this paper, the fundamental principles and steps for implementation of a Korean policy for national radioactive waste management are introduced. Korean practices and prospects regarding radioactive waste management are also summarized, with a focus on strategy for policy-making on SNF management.

Gas threshold pressure and gas permeability of silo concrete specimens for a low- and intermediate-level waste disposal facility in Korea

In order to simulate gas migration at an underground disposal facility, the characteristics of a medium such as the gas threshold pressure and gas permeability need to be measured in advance. In this study, the gas threshold pressure and gas permeability of silo concrete specimens for a Korean LILW (Low- and Intermediate-Level Waste) disposal facility were measured. The concrete specimens had the same composition as the concrete used in the construction of the silo. The gas threshold pressure was measured by injecting a constant gas flow into cross sections of the specimens. To measure the gas permeability, selected pressures were applied to the specimens and the apparent permeability was calculated using the Hagen-Poiseuille equation. The intrinsic permeability was calculated with the Klinkenberg empirical equation. The gas threshold pressure and gas permeability ranged from 30 to 40bar and from 10−17 to 10−18m2, respectively.

Characterization of Radionuclide Retaining Properties of Backfill Materials for Near Surface Repositories for Low and Intermediate Level Radioactive Wastes

ABSTRACTThe sand of glaciolacustrine origin is offered as a major component of the backfill in repositories for solid or solidified low and intermediate level waste (LILW). Clinoptilolite, hematite, and magnesium oxide are offered as additives for increasing of sorption. In this work was carried investigation of sorption properties of sand, clinoptilolite, hematite and magnesium oxide and mixtures of sand with these mineral additives.

방사성물질 운반용기의 적층시험조건에 대한 안전성 평가

IAEA 및 국내의 방사성물질 운반 관련 규정에 따라 중 저준위 방사성폐기물 드럼 8개를 운반할 수 있는 IP-2형 운반용기를 개발하였다. IP-2형 운반용기는 낙하시험 및 적층시험을 거친 후 내용물의 유실 또는 분산과 운반용기 외부표면에서의 방사선량률이 20 % 이상 증가할 수 있는 차폐능력의 상실이 없어야 한다. 본 연구의 목적은 적층시험조건에 대한 시험방법 및 절차를 수립하고 IP-2형 운반용기의 적층조건에 대한 구조적 건전성을 평가하는데 있다. 운반용기의 원형시험모델을 이용하여 운반용기 중량의 5배 하중으로 24시간 동안 압축하는 적층조건에 대한 시험 및 전산해석을 수행하였다. 적층시험 시 운반용기의 모서리기둥에서의 변형률 및 변위를 측정하였으며, 측정된 변형률 및 변위는 해석결과와 서로 일치하였다. 컨테이너 바닥부의 처짐량은 측정이 어렵기 때문에 전산해석 방법으로 구하였다. 모서리기둥의 최대 변위와 컨테이너 바닥의 최대 처짐은 법규에서 규정하는 허용치에 비하여 낮게 나타났다. 적층시험 전 후에는 운반용기의 외형치수, 차폐체 두께, 볼트토크 등을 측정하였으며, 그 값들을 비교분석한 결과 운반용기는 내용물의 유실 및 분산, 차폐체 두께의 감소가 나타나지 않았다. 따라서 적층시험조건에서 IP-2형 운반용기의 구조적 건전성이 입증되었다. Radioactive waste transport package was developed to transport eight drums of low and intermediate level waste(LILW) in accordance with the IAEA and domestic related regulations. The package is classified with industrial package IP-2. IP-2 package is required to undergo a free drop test and a stacking test. After free drop and stacking tests, it should prevent the loss or dispersal of radioactive contents, and loss of shielding integrity which would result in more than 20 % increase in the radiation level at any external surface of the package. The objective of this study is to establish the safety test method and procedure for stacking test and to prove the structural integrities of the IP-2 package. Stacking test and analysis were performed with a compressive load equal to five times the weight of the package for a period of 24 hours using a full scale model. Strains and displacements were measured at the corner fitting of the package during the stacking test. The measured strains and displacements were compared with the analysis results, and there were good agreements. It is very difficult to measure the deflection at the container base, so the maximum deflection of the container base was calculated by the analysis method. The maximum displacement at the corner fitting and deflection at the container base were less than their allowable values. Dimensions of the test model, thickness of shielding material and bolt torque were measured before and after the stacking test. Throughout the stacking test, it was found that there were no loss or dispersal of radioactive contents and no loss of shielding integrity. Thus, the package was shown to comply with the requirements to maintain structural integrity under the stacking condition.

Environmental assessment of advanced partitioning, transmutation, and disposal based on long-term risk-informed regulation: PyroGreen

The safety of spent nuclear fuel (SNF) storage has become one of the major issues of nuclear power plant operation. Direct disposal and recycling have both been criticized by the general public due to uncertainty of the long-term safety of SNF and high level waste (HLW) repositories. To meet the goal of sustainable nuclear energy, an innovative recycling approach using pyrochemical partitioning and transmutation termed PyroGreen, which is a conceptually advanced pyrochemical partitioning flowsheet, has been proposed to eliminate the need for HLW repositories. From the previous partitioning process, PyroGreen uses a combination of hull electrorefining, reductive extraction, and selective oxidation to further decontaminate SNF and HLW into low- and intermediate level waste (LILW). This paper examines the long-term environmental performance of a geological repository that houses all of the final PyroGreen-produced wastes while describing the feasibility of PyroGreen partitioning and transmutation. The final PyroGreen wastes are evaluated based on long-term risk-informed criteria: alpha-emitting isotope concentration, heat generation, and radiation dose in the surrounding biosphere. All final wastes arising from PyroGreen are to be disposed of in a geological repository at an intermediate depth, in compliance with new International Atomic Energy Agency Safety Guide for LILW. Migration assessment found that several long-lived fission products including C-14, Cl-36, Se-79, Sn-125, and I-129 dominate the long-term dose rate, whereas transuranic elements govern the risk of an inadvertent human intrusion scenario. In order to turn entire batches of SNF from 24 light water reactors with 1000 MWe with a design life of 40 years into LILW, we determined and compared the required decontamination factors of the key radionuclides for between intermediate-depth and near-surface disposal.