Hull and end-piece wastes generated from reprocessing plant operations are expected to be disposed of in a deep underground repository as Group 2 TRU wastes under the Japanese classification system. The activated metals that compose the spent fuel assemblies such as Zircaloy claddings and stainless steel nozzles are mixed and compressed after fuel dissolution, and then stuffed into stainless steel canisters. Carbon 14 is a typical activated product in the hulls and end-pieces and is mainly generated by the 14 N(n,p) 14 C reaction. In the previous safety assessment of the TRU waste in Japan, the radionuclides inventory was calculated by ORIGEN-2 code. Some conservative assumptions and preliminary estimates were used in this calculation. For example, total radionuclides generated from a single type of fuel assembly (45 GWd/tU for a PWR unit), and the thickness of the Zircaloy oxide film on the hulls (80 μm) were both overestimated. The second assumption in particular has a large effect on exposure dose evaluation. Therefore, it is essential to have a realistic source term evaluation regarding such items as the C-14 inventory and its distribution to waste parts. In the present study, a C-14 inventory of the hull and end-piece wastes from the operation of a commercial reprocessing plant in Japan corresponding to 32,000 tU (16,000 tU in each BWR and PWR) was calculated. Analysis using individual irradiation conditions and fuel characteristics was conducted on 6 types of fuel assemblies for BWRs and 12 types for PWRs (4 pile types × 3 burnup limits). The oxide film thickness data for each fuel type cladding were obtained from the published literature. Activation calculations were performed by using ORIGEN-2 code. For the amount of spent assembly and other waste characteristics, representative values were assumed based on the published literature. As a preliminary experiment, C-14 in irradiated BWR claddings was measured and found to be consistent with the calculated activation. The total C-14 inventory was estimated as 4.46×10 14 Bq, consisting of 2.58×10 14 Bq for BWRs and 1.87×10 14 Bq for PWRs, and is consistent with the safety assessment of 4.4×10 14 Bq. However, the distribution of the C-14 inventory to hull oxide, which was estimated under the assumption of instantaneous radionuclide release in the safety assessment, decreased from 5.72×10 13 Bq (13% of the total) in the previous assessment to 1.30×10 13 Bq (2.9% of the total; consisting of 1.48×10 12 for BWRs and 1.15×10 13 for PWRs). In other words, the exposure dose peak is reduced to approximate 25% of its previous value due to the use of detailed oxide film data that the BWR cladding has a thin oxide film. Other instantaneous release components for C-14 such as the fuel residual were negligible.

A program of decommissioning of MR research reactor in the Kurchatov institute started in 2008. The decommissioning work presumed a preliminary stage, which included: removal of spent fuel from near reactor storage; removal of spent fuel assemble of metal liquid loop channel from a core; identification, sorting and disposal of radioactive objects from gateway of the reactor; identification, sorting and disposal of radioactive objects from cells of HLRW storage of the Kurchatov institute for radwaste creating form the decommissioning of MR. All these works were performed by a remote controlled means with use of a remote identification methods of high radioactive objects. A distribution of activity along high radiated objects was measured by a collimated radiometer installed on the robot Brokk-90, a gamma image of the object was registered by gammavisor. Spectrum of gamma radiation was measured by a gamma locator and semiconductor detector system. For identification of a presence of uranium isotopes in the HLRW a technique, based on the registration of characteristic radiation of U, was developed. For fragmentation of high radiated objects was used a cold cutting technique and dust suppression system was applied for reduction of volume activity of aerosols in air. The management of HLRW was performed by remote controlled robots Brokk-180 and Brokk-330. They executed sorting, cutting and parking of high radiated part of contaminated equipment. The use of these techniques allowed to reduce individual and collective doses of personal performed the decommissioning. The average individual dose of the personnel was 1,9 mSv/year in 2011, and the collective dose is estimated by 0,0605 man×Sv/year. Use of the remote control machines enables reducing the number of working personal (20 men) and doses. X-ray spectrometric methods enable determination of a presence of the U in high radiated objects and special cans and separation of them for further spent fuel inspection. The sorting of radwaste enabled shipping of the LLRW and ILRW to special repositories and keeping of the HLRW for decay in the Kurchatov institute repository.

Organic compounds of various kinds have been used in the nuclear industry for numerous duties in uranium chemical, metal and ceramic processing plants. In the course of the various operations undertaken, these organic compounds have become contaminated with uranic material, either accidentally or as an inevitable part of the process. Typically, the chemical/physical form and/or concentration of the uranic content of the organics has prevented disposal. In order to address the issue of contaminated liquid organic wastes, the National Nuclear Laboratory (NNL) has developed a suite of treatments designed to recover uranium and to render the waste suitable for disposal. The developed processes are operated at industrial scale via the NNL Preston Laboratory Residue Processing Plant. The Oil Waste Leaching (OWL) Process is a fully industrialised process used for the treatment of contaminated oils with approximately 200 tonnes of uranium contaminated oil being treated to date. The process was originally developed for the treatment of contaminated tributyl phosphate and odourless kerosene which had been adsorbed onto sawdust. However, over the years, the OWL process has been refined for a range of oils including “water emulsifiable” cutting oils, lubricating oils, hydraulic oils/fluids and “Fomblin” (fully fluorinated) oils. Chemically, the OWL process has proved capable of treating solvents as well as oils but the highly volatile/flammable nature of many solvents has required additional precautions compared with those required for oil treatment. These additional precautions led to the development of the Solvent Treatment Advanced Rig (STAR), an installation operated under an inert atmosphere. STAR is a small “module” (100 dm 3 volume) which allows the treatment of both water miscible and immiscible solvents. This paper discusses the challenges associated with the treatment of liquid organic wastes and the process developments which have allowed a wide range of materials to be successfully treated.

Basic characterization of the waste management process during decommissioning of nuclear installations is described in the presented paper. A brief description is given of conditional and unconditional release of materials into the environment. The paper deals also with metal melting as prospective decontamination technique which can significantly reduce metallic radioactive waste. The material and radioactivity flow in the decommissioning process should be followed using the integrated material flow tool that is implemented into the standardized analytical decommissioning parameters calculation code OMEGA. Applying the integrated material flow tool, it is possible to monitor radiological and physical properties of individual material items listed in the nuclear installation input database, from dismantling up to their release into the environment or disposal in repository. Using the OMEGA code and two model databases, several scenarios related to metal melting are evaluated. The impact of applying different input decommissioning parameters on the metal distribution is the main result discussed in the paper.

NNL and ANSTO on behalf of Sellafield Ltd have developed a process for the immobilisation of a range of Pu containing wastes and residues. Following the inactive demonstration of the technology the project is now focusing on the design of an active pilot plant capable of validating the technology and ultimately immobilising a waste inventory containing around 100kg plutonium. The diverse wastes from which it is uneconomic to recover Pu, require a flexible process with a wide product envelope capable of producing a wasteform suitable for disposal in a UK repository. Ceramics, glass ceramics and metal encapsulated wasteforms can be delivered by the process line which incorporates size reduction and heat treatment techniques with the aim of feeding a hot isostatic pressing process designed to deliver the highly durable wasteforms. Following a demonstration of feasibility, flowsheet development is progressing to support the design which has the aim of a fully flexible facility based in NNL’s Central Laboratory on the Sellafield site. Optimisation of the size reduction, mixing and blending operations is being carried out using UO 2 as a surrogate for PuO 2 . This work is supporting the potential of using an enhanced glass ceramic formulation in place of the full ceramic with the aim of simplifying glove box operations. Heat treatment and subsequent HIPing strategies are being explored in order to eliminate any carbon from the feeds without increasing the valence state of the uranium present in some of the inventory which can result in an unwanted increase in wasteform volumes. The HIP and ancillary systems are being specifically designed to meet the requirements of the Sellafield site and within the constraints of the NNL Central Laboratory. The HIP is being configured to produce consolidated product cans consistent with the requirements of ongoing storage and disposal. With the aim of one cycle per day, the facility will deliver its mission of immobilising the identified waste and residues inventory within 3 years. During that period it will also be used to demonstrate the potential of this technology to deliver the immobilisation of a proportion of the UK plutonium stockpile that may not be suitable for use as MOx fuel should that decision be taken.

Carbon-14 is a key radionuclide in the assessment of the safety of a geological disposal facility for radioactive waste because of the calculated assessment of the radiological consequences of gaseous carbon-14 bearing species [i]. It may be that such calculations are based on overly conservative assumptions and that better understanding could lead to considerably reduced assessment of the radiological consequences from these wastes. Alternatively, it may be possible to mitigate the impact of these wastes through alternative treatment, packaging or design options. The Radioactive Waste Management Directorate of the UK’s Nuclear Decommissioning Authority (NDA RWMD) has established an integrated project team in which the partners are working together to develop a holistic approach to carbon-14 management in the disposal system [ii]. For a waste stream containing carbon-14 to be an issue: • There must be a significant inventory of carbon-14 in the waste stream; AND • That waste stream has to generate carbon-14 bearing gas; AND • A bulk gas phase has to entrain the carbon-14 bearing gas: AND • These gases must migrate through the engineered barriers in significant quantities; AND • These gases must migrate through the overlying geological environment (either as a distinct gas phase or as dissolved gas); AND • These gases must interact with materials in the biosphere (i.e. plants) in a manner that leads to significant doses and risks to exposed groups or potentially exposed groups. The project team has developed and used this “AND” approach to structure and prioritise the technical work and break the problem down in a manageable way. We have also used it to develop our approach to considering alternative treatment, packaging and design options. For example, it may be possible to pre-treat some wastes to remove some of the inventory or to segregate other wastes so that they are removed from any bulk gas phase which might facilitate migration through the geosphere. Initially, the project team has undertaken a six month programme of work to examine the current understanding of these aspects and has captured this in the Phase 1 report [ii], in a modelling basis spreadsheet and in scoping assessments, which help us better understand the potential significance of carbon-14. Using the current modelling basis, but ignoring any potential benefits from the geosphere in retarding or preventing gas from reaching the surface, the calculated release of carbon-14 is dominated by: corrosion of irradiated reactive metals (in the operational and early post-closure time frame); corrosion of irradiated stainless steel and leaching of irradiated graphite (in the longer term). The Phase 1 work has shown that there is considerable scope for reducing the calculated radiological consequence for these wastes and a roadmap has been developed for a second Phase of work.

In a green house pot experiment, the effects of three amendments, sulphur (S), ammonium sulphate ((NH 4 ) 2 SO 4 ) and ethylenediaminetetracetic acid (EDTA) were tested for phytoextraction of Cd and Zn by rapeseed ( Brassica napus L.). Elemental sulphur was applied as 20.00, 60.00, and 120.00 mg.kg −1 soil. EDTA was tested at a dose of 585.00 mg.kg −1 soil, and (NH 4 ) 2 SO 4 ) at a rate of 0.23 mg.kg −1 soil. All treatments received a base fertilization (Hogland) before sowing. Plants were harvested after 51 days of growth and shoot dry matter and soil samples were analysed for metal contents. All amendments caused a significant increase in Cd and Zn contents in plant shoots of all treatments than control treatment. Further, EDTA was most effective for extraction metals concentrations in shoot biomass but the plants showed significant signs of toxicity and yield were severely depressed. The addition of sulfur favorably influenced plant biomass production. The fertilized ammonium sulfate treatment resulted in the highest phytoextraction of Cd and Zn and the amounts of these metals accumulated in plant shoot exceeded by a factor of 4 and 3 respectively. Finally, Brassica napus could be used for soil remediation keeping its other uses which will make the contaminated site income generating source for the farmers.

A number of proven technologies applicable to the chemical and physical decontamination of radioactive and non-radioactive contaminants within the environmental remediation and radiological waste management sectors exist. Previous work generally acknowledges that these methods have limitations such as production of large volumes of waste, destruction of the substrate, complex safety considerations [1a] and application of special precautions to meet disposal acceptance criteria [2] . A method that removes a variety of contaminants from the surface and subsurface of porous materials, with minimal contaminated waste arisings, is highly desirable. TechXtract ® is a patented, sequential chemical extraction process developed to remove radionuclides, PCBs, and other hazardous organic and inorganic substances from solid materials such as concrete, brick, steel, and exotic metals [3] . The technology uses multifarious task-specific chemical formulations and engineered applications to achieve surface penetration and removal of the contaminants from the atomic voids of metals and other substrates, or the capillaries and gel pores of concretes. TechXtract ® is proven to remove a variety of contaminants from various substrates, allowing free release of the substrate as waste for disposal, or re-use, whilst producing minimal secondary waste. Data from testing of TechXtract’s capabilities and evidencing the technology’s efficacy during site based applied research and development is presented here.

Nuclear operations with reactors and particle accelerators lead to the generation of neutrons, with a consequent activation of surrounding shielding materials. In general these materials consist of a concrete bioshield with associated metal work from steel reinforcement and support structures. This paper considers the interaction mechanisms of the neutrons in the concrete and the distribution and activity profile of activated materials. The most common source of high thermal neutron fluxes is of course from fission reactor systems. However, there are also other systems such as fusion reactors, which generate a far higher energy neutron flux, as the requirement for electrical confinement of the plasma means that the available bioshielding close to the power source is more limited. In addition, there are also neutron generation mechanisms from particle accelerator systems such synchrotrons and Van De Graaf devices. This study examines the interaction mechanisms of the neutrons in the bioshields, the radioelement abundances of the activated products generated, in-situ and ex-situ measurement of their concentration and profile of the radionuclides and finally methods of assaying and sentencing such materials. Examples of decommissioning operations of reactor and accelerator systems will be given as practical examples of the use of the assaying methodologies developed.

The U.S. Department of Energy (DOE) and CH2M Hill Plateau Remediation Company (CHPRC) are seeking a waste treatment solution for sludge stored in the K-Basin spent fuel pond at the Hanford Nuclear Site, in Washington State, USA. This sludge is a Remote Handled Transuranic (RHTRU) waste destined for final disposal at the Waste Isolation Pilot Pant (WIPP) in New Mexico. Removal of the sludge from the K-Basin and transport for interim storage at the Hanford Site is referred to as Phase 1 in this process. Phase 2 is defined as the treatment and packaging of the sludge such that it can be transported and disposed at the WIPP. This paper discusses work in support of Phase 2. ISI’s GeoMelt ICV process is ideally suited to treating a heterogeneous sludge that is rich in uranium metal and which contains a mixture of other fuel derived products, earthen materials, and miscellaneous items (operational debris, resin, etc). GeoMelt can quickly and efficiently treat small drum load batches and will fully destroy organics, oxidize reactive metals, and permanently immobilize radioactive constituents within a high-integrity vitrified product that will meet or exceed all WIPP acceptance criteria. The GeoMelt Technology has an extensive experience base, having treated more waste by vitrification than any other company in the world (25,500 metric tons). The equipment tested for this Project phase constituted the front end, or Sludge Pretreatment and Transport steps, of the proposed GeoMelt process. These components first focused on an engineering scale (22-liter), followed by testing with a full-scale (130-liter), horizontal rotary plow dryer/mixer. The dryer removes water from the sludge, via external heating and under reduced pressure, and mixes it with glass forming minerals (GFM) prior to treatment in the GeoMelt ICV system. Testing was first performed in July and September 2010 using a 22-liter drying system, which demonstrated a baseline drying technique and allowed an assessment of the resulting physical properties before, during, and after drying/mixing. Full-scale testing using a 130-liter dryer and condensate system was then performed in October 2010. An Operational Acceptance Test (OAT) of the equipment, followed by four “Dryer Holdup” tests and three “GFM Cleanout” tests were performed. Each of the Project Test Objectives was successfully met. Both the 22-liter engineering-scale and the 130-liter full-scale steam jacketed, horizontal plow, dryer are shown to dry and mix 5-vol% solids K-Basin sludge and GFM without difficulty. These test results, combined with previous treatability testing in 2004 wherein successful GeoMelt vitrification of a K-Basin sludge simulant was demonstrated, confirm the efficacy of the overall treatment process towards providing an immediate solution to the final disposition of K Basin Sludge.

Hydrogen generation mitigation for K Basin sludge was examined by encapsulation of uranium metal in BoroBond®, pre-oxidation of uranium metal with Fenton’s reagent and grinding of Densalloy SD170, an irradiated uranium metal surrogate. Encapsulation in BoroBond ® resulted in pressure increase rates at 60 °C ranging from 0.116 torr/h to 0.186 torr/h compared to 0.240 torr/h for a uranium metal in water standard. Samples cast with higher water content led to increased rates. A Fenton’s reagent system consisting of a simple reagent mix of FeSO 4 ·7H 2 O, H 2 O 2 and HCl effectively oxidized 1/4 ″ cubes of uranium metal in under four days at room temperature. Increased peroxide addition rate, increased FeSO 4 ·7H 2 O concentration and low pH all increase the corrosion rate. Densalloy SD170 with an average particle size of 581 μm with 7.63% of particles less than 90 μm was milled so that over 90% of the Densalloy mass measured less than 90 μm in 6 hours of milling. Acceptable wear rates were seen on wear components that were from standard materials (Nitronic SS and 440SS).

Building 220 at Harwell was built by the Ministry of Works as a Radiochemical Research and Development facility in the latter part of the 1940’s. The facility has been operational since 1949 and has been extended several times, most notably the Plutonium Glove Box Wing in the 1950’s and the Remote Handling Wing in the 1980’s. Only the Remote Handling wing remains operational, processing Historic Waste which is being recovered from storage holes elsewhere on site. The remainder of the facility is undergoing progressive strip out and decommissioning. In the Plutonium Wing and associated areas the waste ‘fingerprint’ (nuclide vector) consists predominately of alpha emitting radionuclides. Decommissioning and Decontamination (D&D) operations often result in the production of large volumes of scrap metal waste with little or no radioactive contamination. Proving that the waste is clean can be costly and time consuming, as the shape and size of the metallic waste items often means that it is difficult or impossible to monitor all surfaces using conventional hand-held survey meters. This is a particular problem for alpha contamination measurement. Traditional radiological surveying techniques are very labour intensive and involve surveyors checking every surface using hand held instruments and smear sampling the hard to access areas. Even then 100% monitoring cannot be guaranteed. An alternative to traditional methods is the Long Range Alpha Detection (LRAD) technique which remotely detects and measures secondary ionization created in air by alpha particle interactions, allowing extremely low levels of alpha contamination to be measured. A survey system, IonSens ® , using the LRAD technique, was developed by BNFL Instruments Ltd (now Babcock Nuclear) which allows rapid surveying of scrap metal for alpha contamination at very low levels. Two versions of this system exist but both essentially comprise a measurement chamber into which scrap metal is placed and sealed while a measurement is carried out. Research Sites Restoration Ltd (RSRL) has purchased an IonSens ® system with a single 2m long measurement chamber. This instrument is optimised for clearance monitoring of pipework of up to 150mm diameter. Testing and use of support baskets has enabled the instrument to be used as a clearance monitor for a wide range of metallic waste including hand tools. This paper illustrates how the instrument, and a robust consignment procedure, has been used to justify the clearance of several tonnes of metal wastes for recycling rather than paying for disposal of this material as low level waste.

Since 1979, Uranium enrichment technology has been researched through the gas centrifuge method, at Ningyo-toge Environmental Engineering Center of Japan Atomic Energy Agency (JAEA). In addition, the Demonstration Plant, that is final stage test facilities, was operating continuously from 1988 to 2001. As a result, a lot of residues accumulated in the plant. Most of this accumulation was found be uranium intermediate fluoride. The basic decommission policy of JAEA is that equipments of gas centrifuge will be decontaminated by sulfuric acid immersion method for clearance and reuse. In our plan, approximately 90% of metals will be cleared and reused, and then the remaining 10% will be disposed of radioactive waste. We propose a combination of sulfuric acid immersion method and the systematic chemical decontamination as an efficient method for decontamination of uranium enrichment facilities. This paper focuses on the method and performance of systematic chemical decontamination using IF 7 gas. The following (Figure 1) shows our decommission policy and position of systematic chemical decontamination by IF 7 gas for uranium enrichment plant. The IF 7 treatment technique belongs to the systematic decontamination technology. It has the high performance decontamination technique for the plant that accumulates the uranium intermediate fluoride, such as UF 4 , UF 5 , U 2 F 9 , and U 4 F 17 , which exist in the uranium enrichment plant through the Gas Centrifuge, called GCF. The one of characteristics of the IF 7 treatment, the secondary waste is just an IF 5 and little residues. In addition, this IF 5 can be reused as materials for making new IF 7 gas. The IF 7 treatment can also be performed in the room temperature and very low pressure like a 10–45hPa. Furthermore, the IF 7 treatment is a simple method using chemical reaction. For this reason, we hardly need to care about secondary reaction with the exception of the reaction with IF 7 gas and the uranium intermediate fluoride. This is a very important feature when applying to a large-scale plant. In order to carry out the IF 7 treatment, we only set up a few equipments in GCF uranium enrichment plant, which were IF 7 feeding equipment and two circulating pumps. IF 7 gas cylinders are seated in IF 7 feeding equipment. This is the only equipment. Figure 2 shows the IF 7 treatment system. We carried out the IF 7 treatment for the four cascades in the uranium enrichment Demonstration Plant. The weights of uranium residue in the cascades were approximately 700kgU per cascade prior to the IF 7 treatment. In the IF 7 treatment, we were able to find the near-optimal processing condition. As a result, we could confirm the IF 7 treatment period for one cascade which was 60 days. The main factor to determine the IF 7 treatment period is the pressure and the flow rate of reaction product gas (UF 6 and IF 5 mixture gas) exhausted from the cascade. Although we carried out the IF 7 treatment with the maximum value of the flow rate, which our facility has, it is possible to further shorten the IF 7 treatment period by setting a higher gas flow rate. Moreover, after the IF 7 treatment, we evaluated the uranium recovery rate for cascades and the residues’ uranium weight in the main equipment of GCF. In addition, in the evaluation of the uranium recovery rate, we enable to confirm the uranium recovery rate of all cascades achieved more tan 98%. Furthermore, the average of uranium recovery rate more than 99% in the cascade that has been processed at the end. As a result, radioactive concentration of uranium in the main parts of the GCF fell to 1.0B q/g and below.

With more than 25 years of experience in the development of reactor vessel internals segmentation and packaging technology, Westinghouse has accumulated significant know-how in the reactor dismantling market. Building on tooling concepts and cutting methodologies developed decades ago for the successful removal of nuclear fuel from the damaged Three Mile Island Unit 2 reactor (TMI-2), Westinghouse has continuously improved its approach to internals segmentation and packaging by incorporating lessons learned and best practices into each successive project. Westinghouse has developed several concepts to dismantle reactor internals based on safe and reliable techniques, including plasma arc cutting (PAC), abrasive waterjet cutting (AWJC), metal disintegration machining (MDM), or mechanical cutting. Westinghouse has applied its technology to all types of reactors covering Pressurized Water Reactors (PWR’s), Boiling Water Reactors (BWR’s), Gas Cooled Reactors (GCR’s) and sodium reactors. The primary challenges of a segmentation and packaging project are to separate the highly activated materials from the less-activated materials and package them into appropriate containers for disposal. Since space is almost always a limiting factor it is therefore important to plan and optimize the available room in the segmentation areas. The choice of the optimum cutting technology is important for a successful project implementation and depends on some specific constraints like disposal costs, project schedule, available areas or safety. Detailed 3-D modeling is the basis for tooling design and provides invaluable support in determining the optimum strategy for component cutting and disposal in waste containers, taking account of the radiological and packaging constraints. Westinghouse has also developed a variety of special handling tools, support fixtures, service bridges, water filtration systems, video-monitoring systems and customized rigging, all of which are required for a successful reactor vessel internals segmentation and packaging project. The purpose of this paper is to provide an overview of the Westinghouse reactor internals segmentation experience by illustrating projects related to various types of reactors and providing feedback from project execution.

Since more than 20 years the company Siempelkamp is deeply involved in the field of melting and recycling of radioactively contaminated metals from operation and decommissioning of nuclear installations across Europe. The experience of this long period shows clearly that only a combination of recycling inside the nuclear industry and release for reuse outside the nuclear market will generate the optimum results for the minimisation of radioactive waste volume. Final disposal volume is becoming more and more the status of an own resource within our nuclear business and should be handled very carefully in the future. The paper gives a compact overview about the impressive results of melting treatment, the current potential of the melting plant CARLA and about further developments.

Metal scrap is a major waste generated from the decommissioning of nuclear facilities. Through a decontamination process, most of the metal scraps can be cleaned to meet the clearance levels, which can then be reused or released according to the regulatory procedure. Usually, chemical processes will be used in the cleaning step. Phosphoric acid and fluoroboric acid are the typical chemicals used for decontamination. Although the decontaminant could be reused multiple times after regeneration, its decontamination efficiency would decrease after 3 to 5 cycles. In addition, the radioactive nuclides such as Cs-137 are not easily removed during the regeneration process; it tends to accumulate slowly in the decontaminant. According to the ALARA principle, decontaminant must be replaced if its radioactive activity exceeds the regulatory levels. As a result, a significant amount of spent strong acid solution would be generated. The conventional way of treatment is to neutralize the acid solution with an alkaline solution. However, such method will produce a large amount of sludge that requires further stabilization, which offsets the advantages of metal decontamination by use of the decontaminant. A high-efficiency solidification method has been developed and used to treat the spent phosphoric acid and fluoroboric acid solution in Institute of Nuclear Energy Research (INER). The self-polymerization nature of highly concentrated phosphoric acid is adopted to immobilize the radioactive nuclides. The volume of solidified form is almost equal to that of the treated acid solution. The waste form demonstrates its quality by compression test and leaching test. This cementation process is an excellent method to minimize the secondary waste, which is generated from chemical decontamination for treating metal waste.

The Sellafield Magnox Swarf Storage Silo (MSSS) was constructed to provide an underwater storage facility for irradiated magnox cladding metal swarf, as well as miscellaneous beta-gamma waste from several sources. Liquid effluent arisings from hazard reduction activities at this facility represent the toughest effluent treatment challenge within the company’s Legacy Ponds & Silos portfolio. The key requirement for hazard reduction has generated many substantial challenges as the facility is readied for decommissioning. This has demanded the production of carefully thought out strategies for managing, and overcoming, the key difficulties to be encountered as hazard reduction progresses. The complexity associated with preparing for waste retrievals from the Magnox Swarf Storage Silo, has also generated the demand for a mix of creativity and perseverance to meet the challenges and make progress.

This study is part of a generic investigation for the assessment of the required minimum distance between a Spent Fuel/High-Level Waste/ Intermediate-Level Waste (SF/HWL/ILW) repository and a Low/ Intermediate-Level Waste (L/ILW) repository. For this, a large-scale numerical model was constructed to investigate the two-phase flow behavior for such a repository configuration in a low-permeability claystone formation. The modeling focused on the pressurization mechanisms associated with (a) resaturation of backfilled underground facilities, (b) thermal effects caused by heat generation from the SF/HLW canisters, and (c) gas generation from corrosion and degradation of different wastes in the L/ILW and ILW caverns and in the SF/HLW emplacement tunnels. The model accounts for gas generation from corrosion and degradation of both L/ILW and ILW wastes indicating decreasing rates with time, and from corrosion of the SF/HLW canisters characterized by a constant rate. Heat generation from radioactive decay of radionuclides of MOX/UO2 wastes is described by an exponential decay with time. The preceding operational phases of the different repository components were simulated representing the transient initial conditions for the post-closure phase. The simulated pressure buildup in the L/ILW repository shows a near linear increase between 10 and 4,000 years when the peak pressure of 6.5 MPa is reached for a repository at about 370 m bg. This is followed by a similar decline, recovering to near hydrostatic pressures after 1 million years. The SF/HLW repository (repository level 600 m bg) indicates a pressure rise between 100 and 1,000 years affected by the early thermal effects, followed by a steep increase between 3,000 and 100,000 years when the pressures level off to a maximum of 6.5 MPa after 160,000 years (corresponding to a steel corrosion rate of 1 μm/year). This is the time when all the metal is corroded and the gas generation stops resulting in a sudden decline, and the pressures level off to about 4.5 MPa in the SF/HLW emplacement tunnel after 1 million years. The numerical modeling demonstrates that the main pressurization mechanism is from gas generation in the different repository components. The pressure histories show a distinct separation of the pressure peaks between the L/ILW repository and the SF/HLW/ILW repository. Moreover, the thermal phenomena affect the pressures in the SF/HLW repository at early time only (prior to about 2,000 years). The thermal expansion of the pore water in the nearfield around the SF/HLW tunnels does produce a relatively steep pressure buildup after 100 years, but it dissipates rapidly prior to the main pressure buildup caused by the gas generation and gas accumulation in the SF/HLW repository. The thermally induced pressure buildup is restricted to the vicinity of the SF/HLW emplacement tunnels (decameter range) and thus, significant interference of the thermally induced pressure perturbation around the SF/HLW/ILW repository with the early gas pressure buildup in the L/ILW repository can be excluded.

In low/intermediate-level waste (L/ILW) repositories, anaerobic corrosion of metals and degradation of organic materials produce hydrogen, methane, and carbon dioxide. Gas migration in a L/ILW repository is one of the processes evaluated in the safety assessment of deep geological disposal in low-permeability formations, in particular with respect to the development of gas pressures in the repository caverns which could negatively affect the host rock or the engineered barrier system (EBS). In order to restrict build-up of gas overpressures in the emplacement caverns, Nagra (National Cooperative for the Disposal of Radioactive Waste, Switzerland) has proposed design options aimed at increasing the gas transport capacity of the backfilled underground structures, compromising neither the low hydraulic conductivity nor the radionuclide retention capacity of the EBS (Nagra, 2008). They involve specially designed backfill and sealing materials such as high porosity mortars as backfill materials for the emplacement caverns and sand/bentonite (S/B) mixtures with a bentonite content of 20% to 30% for the seals themselves and for backfilling other underground structures. These increased gas permeability materials can supplement the gas flow that is expected to occur through the excavation damaged zone (EDZ) and avoid the creation of overpressures. Preliminary experimental studies have confirmed the gas transport capacity of the S/B mixtures and demonstrated the ability to design mixtures with specific target permeabilities for water and gas flow (Nagra, 2008). Two-phase flow modelling studies have shown that the gas transport capacity of seals is largely dependent on their permeability and length. More detailed models of sealing elements show a rather complex history of seal saturation during the early saturation phase and the later gas escape phase (Gaus et al., 2010). Note, however, that current modelling approaches are based on parameters and conceptual understanding of small-scale laboratory experiments. Two large(r) scale experiments which aim at validating and, if necessary, improving current conceptual models for the resaturation and gas invasion processes into S/B seals and the determination of up-scaled gas / water permeabilities of S/B seals (i.e. two-phase flow parameters for large-scale models) have been initiated and will be highlighted in the paper. The first one, a mock-up experiment, was set up in 2010 as part of the EU 7 th FP project FORGE, aiming at demonstrating seal performance on an intermediate (decimetre scale). The second one is a large-scale experiment (metre-scale), the Gas-Permeable Seal Test (GAST), which was also initiated in 2010 at the Grimsel Test Site (GTS). For GAST, a seal will be emplaced at the GTS to demonstrate the effective functioning of gas-permeable seals on a realistic scale and with realistic boundary conditions (‘proof of concept’).

Deep vadose zone environments can be a primary source and pathway for contaminant migration to groundwater. These environments present unique characterization and remediation challenges that necessitate scrutiny and research. The thickness, depth, and intricacies of the deep vadose zone, combined with a lack of understanding of the key subsurface processes (e.g., biogeochemical and hydrologic) affecting contaminant migration, make it difficult to create validated conceptual and predictive models of subsurface flow dynamics and contaminant behavior across multiple scales. These factors also make it difficult to design and deploy sustainable remedial approaches and monitor long-term contaminant behavior after remedial actions. Functionally, methods for addressing contamination must remove and/or reduce transport of contaminants. This problem is particularly challenging in the arid western United States where the vadose zone is hundreds of feet thick, rendering transitional excavation methods exceedingly costly and ineffective. Delivery of remedial amendments is one of the most challenging and critical aspects for all remedy-based approaches. The conventional approach for delivery is through heterogeneous deep vadose zone environments present hydrologic and geochemical challenges that limit the effectiveness. Because the flow of solution infiltration is dominantly controlled by gravity and suction, injected liquid preferentially percolates through highly permeable pathways, by-passing low-permeability zones that frequently contain the majority of contamination. Moreover, the wetting front can readily mobilize and enhance contaminant transport to the underlying aquifer prior to stabilization. Development of innovative in-situ technologies may be the only means to meet remedial action objectives and long-term stewardship goals. Surfactants can be used to lower the liquid surface tension and create stabile foams, which readily penetrate low permeability zones. Although surfactant foams have been used for subsurface mobilization efforts in the oil and gas industry, thus far the concept of using foams as a delivery mechanism for transporting remedial amendments into deep vadose zone environments to stabilize metal and long-lived radionuclide contaminants has not been explored. Foam flow can be directed by pressure gradients, rather than being dominated by gravity; and foam delivery mechanisms limit the volume of water (< 5% vol.) required for remedy delivery and emplacement, thus mitigating contaminant mobilization. We will present the results of a numerical modeling and integrated laboratory-/intermediate-scale investigation to simulate, develop, demonstrate, and monitor (i.e., advanced geophysical techniques and advanced predictive biomarkers) foam-based delivery of remedial amendments to remediate metals and radionuclides in vadose zone environments.